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abstract | Systems for enhancing preignition conditions of a fusion reaction are disclosed. A first system includes a target chamber for receiving a fusion fuel, and energy driving means oriented to direct plasma confinement structure onto to the fusion fuel to facilitate ignition of a controlled fusion reaction of said fusion fuel. A plurality of electron sources provides electron beams of a predetermined energy and one of fluence and quantity, directed onto and illuminating, a fusion fuel-derived plasma for controlling the ratio of ion temperature and electron temperature of the plasma. A second system comprises a central target chamber for receiving a spherical pellet of fusion target material and at least first and second pluralities of energy drivers oriented to supply temporally-staged X-ray pulses to the fusion target material in a 3-dimensionally symmetric manner about said pellet. A third system combines aspects of the first and second systems. |
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description | This application claims priority to and the benefit of U.S. Provisional Application Ser. Nos. 60/781,977 which was filed Mar. 14, 2006, entitled ARC QUENCHING CIRCUIT TO MITIGATE ION BEAM DISRUPTION, and 60/784,852 which was filed Mar. 22, 2006, entitled A METHOD OF ION BEAM CONTROL FOR GLITCH RECOVERY the entirety of which is hereby incorporated by reference as if fully set forth herein. The present invention relates generally to ion implantation systems, and more particularly to an arc quenching circuit for extinguishing an arc that may form between high voltage electrodes within an ion implantation system, and to a method of repainting the ion beam to recover any dose losses during such arcing to attain more uniform ion implantations and duty factor reduction. Ion implantation systems are used to impart impurities, known as dopant elements, into semiconductor substrates or wafers, commonly referred to as workpieces. In such systems, an ion source ionizes a desired dopant element, and the ionized impurity is extracted from the ion source as a beam of ions. The ion beam is directed (e.g., swept) across respective workpieces to implant ionized dopants within the workpieces. The dopant ions alter the composition of the workpieces causing them to possess desired electrical characteristics, such a may be useful for fashioning particular semiconductor devices, such as transistors, upon the substrates. The continuing trend toward smaller electronic devices has presented an incentive to “pack” a greater number of smaller, more powerful and more energy efficient semiconductor devices onto individual wafers. This necessitates careful control over semiconductor fabrication processes, including ion implantation and more particularly the uniformity of ions implanted into the wafers. Moreover, semiconductor devices are being fabricated upon larger workpieces to increase product yield. For example, wafers having a diameter of 300 mm or more are being utilized so that more devices can be produced on a single wafer. Such wafers are expensive and, thus, make it very desirable to mitigate waste, such as having to scrap an entire wafer due to non-uniform ion implantation. Larger wafers and high density features make uniform ion implantation challenging, however, since ion beams have to be scanned across larger angles and distances to reach the perimeters of the wafers, yet not miss implanting any region therebetween. In addition, the high voltage necessary to supply the ion source of such an ion beam is subject to occasional arcing between the various extraction and suppression electrodes and other nearby parts. This tendency for arcing often fully discharges one or more affected HV supplies until the arc naturally self-extinguishes at a much lower supply voltage. While arcing, the beam current may become serious erratic or may be interrupted until the supply voltage is restored, during which time ion implantation may experience intermittent or non-uniform ion implantation dose levels. Accordingly, there is a need for mitigating the effects of HV arcing associated with an ion source or the electrodes of an ion implanter to provide uniform implantation over such larger implantation angles and distances of the ion beam. The following presents a simplified summary in order to provide a basic understanding of one or more aspects of the invention. This summary is not an extensive overview of the invention, and is neither intended to identify key or critical elements of the invention, nor to delineate the scope thereof. Rather, the primary purpose of the summary is to present some concepts of the invention in a simplified form as a prelude to the more detailed description that is presented later. The present invention is directed to a circuit for quenching an arc that may form between high voltage (HV) electrodes associated with the ion source of an ion implantation system to shorten the duration of the arc, to mitigate erratic ion beam current, and to mitigate non-uniform ion implantations, for example. Several high voltage high speed (HVHS) switching circuit arrangements are disclosed that each incorporate a HVHS switch added in series between each high voltage supply and its respective electrode (e.g., a suppression or extraction electrode) associated with the ion source for quickly extinguishing the harmful arcs. The arcs that otherwise form in these areas have a tendency to substantially discharge the high voltage capacitors within such HV power supplies, for example, for the ion source or extraction electrode supply voltage (Vext), or the suppression electrode supply voltage (Vsup). Consequently, the ion beam current is dramatically affected by these “glitches” in the ion beam current (Ibeam), and accordingly takes considerable time thereafter for the supply voltages and beam current Ibeam to recover. Thus, the arc quench circuit of the present invention mitigates ion beam disruption and speeds beam current recovery. Further, the circuit and method also facilitates repainting the ion beam over those areas where an arc was detected to recover any dose loss during such arcing. The circuit also comprises a motion control system operable to control horizontal and vertical scan motions of a wafer implanted by the ion implanter, to monitor horizontal and vertical scan positions associated with the detection of the arc, and to initiate a return to an initial position along a scan associated with the detection of the arc. The trigger control circuit of the present invention may also be further operable to receive a repaint command from the motion control system, and to force the HV switch on or off in response to the repaint command in order to repaint the ion beam between the initial position along the scan associated with the detection of the arc and a final position associated with the detection of the arc, thereby recovering any dose loss during such arcing. According to one or more aspects of the present invention, an arc quenching circuit for an ion source of an ion implantation system suitable for use in implanting ions into one or more workpieces is disclosed. In one aspect of the invention, the system includes one or more high voltage high speed (HS) switches connected in series with a HV power supply (HVPS) for the ion source (or one of several HV extraction or suppression electrodes), the HVHS switches operable to interrupt the HV power supply current to the ion source or electrodes to quench the arc, and further operable to reestablish the power supply current. The quantities of ions that can be extracted from the ion source are in the form of an ion beam having a beam current. The system also includes a trigger control circuit operable to detect a current or voltage change associated with the ion source or HV electrodes and to control the one or more HVHS switches to open or close based on the current or voltage change detection. One or more protection circuits are also included to protect the respective HVHS switch, and are operable to absorb energy from reactive elements external to the respective HV switch, and to clamp an over-voltage that may occur across the switch. In another aspect of the invention the system further comprises a synchronization circuit operable to synchronize and time the trigger control circuits of two or more arc quenching circuits for the opening and closing of two or more high voltage switches. In still another aspect, the current or voltage change detection associated with the ion source comprises detecting one of a current surge in the HV power supply, a decrease in an ion beam current, a drop in a suppression electrode voltage, and a drop in an extraction electrode voltage. In yet another aspect, one of the protection circuits is connected in series with the HV switch it protects. In one aspect, one of the protection circuits is connected in parallel with the HV switch it protects. In another aspect of the present invention the system further comprises an extraction suppression electrode located close to the ion source. In still another aspect, the current or voltage detection is accomplished during the ion implantation process to facilitate feedback or closed-loop adjustments to the ion source current or voltage. In yet another aspect, the current or voltage detection is accomplished prior to the ion implantation process to facilitate open loop adjustments to the ion source current or voltage. In another aspect, the current or voltage detection is accomplished during the ion implantation process to facilitate feedback or closed-loop adjustments to the ion beam current. In accordance with one or more other aspects of the present invention, an arc quenching circuit for a high voltage power supply of an ion implantation system is also disclosed comprising a high voltage switch connected in series with a high voltage power supply for an electrode associated with the implanter, operable to interrupt and reestablish a current to the electrode, to quench an arc produced within the ion implantation system. The system also includes a trigger control circuit operable to detect a current or voltage change associated with the electrode and to control the one or more HV switches to open or close based on the detection. Finally, the system comprises one or more protection circuits, each protection circuit associated with one of the high voltage switches, operable to absorb energy from reactive elements external to the respective HV switch, and to limit an over-voltage across the switch. In accordance with another aspect of the present invention, a method of quenching an arc in an ion implantation system and repainting the ion beam to recover any dose loss during such arcing using an arc quenching circuit associated with a high voltage supply for an electrode of the ion implantation system comprises horizontally scanning a wafer in front of the ion beam, vertically scanning the wafer in front of the ion beam, detecting a current or voltage change associated with an arc at the electrode, monitoring the horizontal and vertical scan positions associated with the detection of an arc, and monitoring the time duration associated with the detection of an arc. The method further comprises controlling a HV switch connected between the high voltage supply and the electrode to open when an arc is detected in order to interrupt an arc current to the electrode and to quench the arc, controlling the HV switch to close when the arc is not detected in order to connect the high voltage supply to the electrode and to establish the ion beam, and repainting the ion beam after an arc. In another aspect of the invention the repainting process comprises moving the wafer to a first horizontal and vertical scan position associated with an initial detection of the arc, and closing the HV switch to enable the ion beam, and scanning the wafer horizontally and vertically in front of the ion beam until a second horizontal and vertical scan position associated with a final detection of the arc is encountered, and opening the HV switch to disable the ion beam. In yet another aspect of the invention the method further comprises synchronizing two or more arc quenching circuits having two or more high voltage switches used to quench an arc between the electrodes of two or more respective high voltage power supplies for the ion implanter, and repainting the ion beam subsequent to detecting an arc from the electrodes. In another aspect of the present invention the repainting process is only accomplished if the time duration of the arc detected is longer than a predetermined interval. In still another aspect, the repainting process is delayed until the ion beam scan returns to a wafer exchange position of the wafer. In yet another aspect, the repainting process is delayed until the ion beam scan completes a current horizontal scan movement. In one aspect, the horizontal and vertical scanning continue after the detection of an arc. In another aspect, the repainting process is delayed until the end of the ion beam scans, wherein one or more arc detections may be repainted collectively during one or more continuous scan movements. To the accomplishment of the foregoing and related ends, the following description and annexed drawings set forth in detail certain illustrative aspects and implementations of the invention. These are indicative of but a few of the various ways in which the principles of the invention may be employed. Other aspects, advantages and novel features of the invention will become apparent from the following detailed description of the invention when considered in conjunction with the drawings. The present invention will now be described with reference to the drawings wherein like reference numerals are used to refer to like elements throughout. The illustrations and following descriptions are exemplary in nature, and not limiting. Thus, it will be appreciated that variants of the illustrated systems and methods and other such implementations apart from those illustrated herein are deemed as falling within the scope of the present invention and the appended claims. The present invention relates to quenching an arc that may form between high voltage extraction or suppression electrodes, for example, associated with an ion source of an ion implantation system. An arc quenching circuit is discussed that shortens the duration of the arc, thereby mitigating the duration of erratic ion beam current, and minimizing the non-uniformity of ion implantations, for example. Further, the circuit and method also facilitates repainting the ion beam over those areas where an arc was detected in order to recover any dose loss during such arcing. In one aspect, the circuit may also comprise or communicate with a motion control system operable to control horizontal and vertical scan motions of a wafer implanted by the ion implanter, to monitor horizontal and vertical scan positions associated with the detection of the arc, and to initiate a return to an initial position along a scan associated with the detection of the arc. In accordance with the present invention, high voltage high speed (HVHS) switching circuits comprising HVHS switches (e.g., 65 KV @ 2 MHz MOSFET switches) are added in series with the high voltage supplies to the suppression and/or extraction electrodes, or ground electrodes, for example, to extinguish the harmful arcs. When such HV arcs occur, the high voltage capacitors of such HV power supplies may be substantially discharged. This deep discharge dramatically affects the ion beam current and requires considerable time thereafter for the power supply voltages and the ion beam current Ibeam to recover. Such high voltage high speed switches have just recently become available as a manufactured item, and thus find immediate use in such applications incorporating the arc quenching circuit(s) of the present invention. Advantageously, these HVHS switches also provide the ion implanter with the ability to simply turn the ion beam ON or OFF at will, either manually with a switch or via command from one of the implanters control systems, its computer, or by an external input. As ion implanters may take a considerable time to sequence through a power up and warm up to a stable ion beam level that is useful for implantation, it is a tremendous advantage, after such a warm-up, to be able to turn the beam ON/OFF, for example, when loading or unloading a new wafer, at the start/end of each wafer scan, and if desired, even in portions of the over-travel regions of each row scan of a wafer. Thus, the system of the present invention facilitates this beneficial feature, known as “beam duty factor”, which is the ratio of ON to OFF time of the ion beam. By having this ability to reduce this beam duty factor, the inventors of the present system also anticipate reducing the particle count on a wafer, because the beam will be used to a greater percentage usefully on the wafer and less on peripheral surfaces adjacent to the wafer. The high voltage switches are controlled by trigger circuits which detect current or voltages changes in the HV supplies to the electrodes, such changes as are associated with the formation of an arc at one of the electrodes. The arc quenching circuit also comprises one or more protection circuits for the HV switches to absorb excess energy from reactive components surrounding the HVHS switches and clamp any overvoltages from the HVHS switches. The protection circuits may be connected in parallel with and/or in series with a respective HVHS switch. The arc quenching circuits of the present invention may further comprise a synchronization circuit to sequence and synchronize the reestablishment of the current and voltage to each of three electrodes and high voltage supply circuits associated with an ion implantation system. To facilitate repainting the ion beam over those areas where the arc has occurred, the circuitry of the present invention also communicates with the motion control system of the ion implanter. In particular, during a typical implantation scan, horizontal (e.g., row) and vertical motion of the wafer in front of the ion beam (or the beam relative to the wafer) is monitored, typically by the motion control system. When an arc occurs, the initial and final horizontal and vertical positions associated with the arc detection, for example, are stored for a subsequent repaint process. Then, at the end of a particular row scan, or at the end of the wafer scan, for example, around the load/unload or wafer exchange position, the motion control system initiates the repaint process. In the repaint process, the ion beam is first disabled by opening the HVHS switch, and the wafer is moved to a first horizontal and vertical scan position associated with the initial detection of the arc. Optionally, the beam may be scanned over the wafer if the implanter facilitates this type of scanning. Alternately, the repaint process may return the wafer motion to the beginning of the row (horizontal) scan wherein the arc was initially detected, where the scan motion can begin a row as usual prior to the position of the initial arc detection. This variation may be preferable, as the scan motion would then be fully accelerated up to the same speed as that which was present when the arc was initially detected. Thereafter, when the ion beam is at the position of initial arc detection, the beam is enabled by closing the HVHS switch, while the wafer is horizontally and vertically scanned until a second horizontal and vertical scan position is encountered associated with a final detection of the arc. When this final detection point is reached, the HVHS switch is opened to disable the ion beam. Although the HVHS arc quenching circuit of the present invention is illustrated and described in the context of ion sources and ion implanters, those skilled in the art may appreciate that such high voltage high speed arc quenching circuits may also be utilized in other applications requiring HV and high speed arc quenching, such as x-ray equipment, accelerators, other ion source applications, for example. In this manner, unwanted arc shorting of high voltage supplies may be quenched before the high voltage power supply has been significantly discharged and has had a chance to affect the output of related systems (e.g., the ion beam of an ion implanter). Referring initially to FIG. 1, an exemplary arc quenching circuit 100 for a high voltage supply of an ion source suitable for implementing one or more aspects of the present invention is depicted in block diagram form. The circuit 100 includes a high voltage power supply 102, a high voltage high speed HVHS switch 104, a current transformer (CT) 106 for detecting a change of current in the supply 102 to an ion source 120 for producing a quantity of ions that can be extracted in the form of an ion beam 130. The change of supply current to the ion source 120 is detected by the CT 106 and a trigger control circuit 108 which opens HVHS switch 104 when a current surge is detected. The HVHS switch 104 is protected by parallel and series protection circuits 110 and 115, respectively, to absorb energy from reactive components surrounding the switch 104 and protect the switch from over-voltage damage. The protection circuits 110 and 115 also protect the switch 104 and other components of the ion implanter, by dampening any ringing induced by switching transients and the reactive components external to the HVHS switch 104. The arc quenching circuit 100 may be used in any ion implanter, or other such applications as may use a high voltage supply subject to arc discharges at the output of the supply. For example, arc quenching circuit 100 operates by detecting a current surge in CT 106 when an arc occurs within the ion source 120, at the extraction electrodes, or at the output of the ion source, for example, as in the ion beam current. The trigger control circuit 108 receives the current surge detection from the CT 106 and in turn controls the HVHS switch 104 to open. When HVHS switch 104 opens, the arc current through CT 106 drops to near zero and the arc extinguishes or “quenches”. The inventors of the present invention have further found that the arc must remain extinguished for a finite period of time before the HVHS switch is closed again, or the more conductive gaseous byproducts from the arc which remain in the region, will permit the arc to reoccur. Thus, a delay time within the trigger control circuit or within a synchronization circuit, (e.g., 740 of FIG. 7), may provide such a delay, and will be discussed further infra. Alternately, the switch may be allowed to repeatedly open and close until the arc no longer reoccurs. FIG. 2 illustrates an exemplary ion implantation system 200 such as may utilize the arc quenching circuit similar to that of 100 of FIG. 1, of the present invention. For example, ion implantation system 200 comprises an ion source 120 having several extraction electrodes 208, for providing a source of ions as an ion beam 130 for implantation system 200. The ions within ion beam 130 are initially analyzed in a first region 210 by a mass analyzing magnet 212 by way of magnetic deflection to filter ions of unwanted mass or energy. The mass analyzing magnet 212 operates to provide a field across the beam path 130 so as to deflect ions from the ion beam 130 at varying trajectories according to mass (e.g., charge to mass ratio). Ions traveling through the magnetic field experience a force that directs individual ions of a desired mass along the beam path 130 and deflects ions of undesired mass away from the beam path. Those ions of ion beam 130 having the desired mass and energy are then accelerated or decelerated in a second region 220, focused by resolving aperture and deceleration plates 232, measured by setup faraday cup 234, and in region 230, the beam is conditioned by a plasma shower 236 providing for space charge neutralization. Finally, the ion beam 130 enters an end station 240 for implantation in a wafer 242 the dose level of which is measured by a disk faraday cup 244. During ion implantation, an arc 205 may occur between the high voltage extraction, suppression, or ground electrodes, for example, associated with the ion source. In conventional implantation systems, this arc has a tendency to completely discharge the high voltage supply before the arc self-extinguishes. The arc quenching circuit 100 of FIG. 1, for example, is designed to avoid this problem. FIG. 3, for example, illustrates a plot 300 of the change in the beam current which results when an arc occurs in the high voltage extraction and suppression voltages of an ion implanter similar to the ion implantation system of FIG. 2. Plot 300 of FIG. 3, for example, illustrates that an arc discharges extraction voltage 310 from about 2.2 KV to near 0V at a time 315 at about 0 ms. At about the same time, the suppression voltage 320 drops from about −9.3 KV to near 0V while the beam current Ibeam 330 drops to near 0 mA. As the extraction and suppression voltages 310, and 320, respectively, fall to near 0 volts, the arc self extinguishes, thereby allowing these voltages to recharge toward their original voltage levels. As shown at 340, the extraction voltage 310 overshoots this original voltage, and detrimentally delays the recovery of beam current Ibeam 330 until time 345 at about 67 ms wherein extraction voltage 310 has generally recovered. It may be observed from plot 300 that extraction voltage changes have a relatively large and lasting impact on beam current. Thus, FIG. 3 suggests that it may be very beneficial to quickly open the high voltage current paths between the electrodes for the ion beam and the high voltage supplies for the electrodes before the HV supplies have had a chance to significantly discharge. The HVHS switch of the present invention accomplishes this goal. FIG. 4 illustrates a portion of an exemplary ion implantation system 400 having high positive voltage extraction supply 403 which feeds extraction slits 404, and a high negative voltage suppression supply 406 which feeds suppression electrodes 408 neighboring ground electrodes 409. The HV suppression supply 406 has a conventional arc suppression or protection circuit 410, which may use a current limiting resistor 412 to limit the arc current to the suppression electrodes 408, a capacitor 414 to filter and stabilize the voltage of the supply, and a fly-back diode 416 to limit any reverse voltages generated from reactive elements of the circuit during arc on-off cycling. In the context of the present invention, the arc protection board 410 may also be used in association with the HVHS switch (e.g., 104 of FIG. 1) of the invention to protect the HVHS switch from damage. FIG. 5 illustrates an exemplary arc quenching circuit 500 utilized in association with a high voltage supply of an ion source such as may be used in an ion implantation system in accordance with the present invention. For example, arc quenching circuit 500 comprises a high voltage supply (Vb) 503 (e.g., a high voltage positive supply) connected in series with a HVHS switch 504 (e.g., a series stack of MOSFET transistors) and a series switch protection circuit 515, which drives a load (e.g., an ion source 120). The HVHS switch 504 is also connected in parallel with a parallel protection circuit 510 which protects the switch 504 from reactive overvoltages, for example. Arc quenching circuit 500, further comprises a current transformer CT 506 that detects a change of current in the supply 503 to the ion source 120, used for example, for producing a quantity of ions that can be extracted in the form of an ion beam (e.g., ion beam 130 of FIG. 1). Circuit 500 also includes a trigger control unit 508 for detecting a change of current in the supply current (Iext) 509 to the ion source 120. If a current surge indicative of an arc, is detected in supply current (Iext) 509 by the CT 506, then the trigger control circuit 508 controls HVHS switch 504 to open and quench the arc. A capacitance C1 518 within the load (e.g., an ion source 120), and the voltage at the load (Va) is therefore isolated by HVHS switch 504 from the voltage Vb of the high voltage supply 503. Thus, Va at C1 514 of the load may discharge due to the occurrence of an arc, but the positive supply voltage Vb will remain generally charged at voltage due to isolation by the HVHS switch 504. Again, the HVHS switch 504 is protected by parallel and series protection circuits 510 and 515, respectively, to absorb energy from reactive components external to the switch 504 and therefore protect the switch from over-voltage damage. The arc quenching circuit 500 of the present invention may be used in any ion implanter, or other such applications as may use a high voltage supply subject to arc discharges at the output of the supply. FIG. 6 illustrates the arc quenching effects of opening and closing a HVHS switch of the arc quenching circuit of the present invention tested in a vacuum (e.g, 650 of FIG. 6), during arcing of an extraction electrode associated with an ion source. FIG. 6, illustrates a plot 650 of the relative amplitude level of signals provided by an arc quenching circuit (e.g., 500 of FIG. 5), in accordance with the present invention during arcing of an extraction electrode (e.g., 208 of FIG. 2) associated with an ion source (e.g., 120 of FIGS. 1 and 5), as tested in the actual vacuum environment, for example, of an ion implanter. FIG. 6 further illustrates the faraday current detected 660, during the opening and closing of a HVHS switch (e.g., 504 of FIG. 5) as measured at the extraction electrode voltage Vext 670, which is fed by a high positive supply voltage, and as triggered by a Vext trigger control signal 680 derived by the current in the Vext power supply (e.g., from CT 506), and having a suppression voltage Vsup 690, which is fed by a high negative supply voltage. FIG. 6 further illustrates a voltage 670 across a HVHS switch 504 when the switch is closed producing a high Vext level 670a and when the switch is open producing a low Vext level 670b, the high voltage supply Vb 630 at the supply 503, and the high voltage Va 620 as seen at the load (e.g., 120). FIG. 6 further demonstrates that the HVHS switch can not only greatly shorten the duration of the glitch, but also allows the ion beam to recover quickly. Additional explanations will be given in association with FIG. 10B. Prior to time 0.0, when an arc occurs, the detected faraday current I-faraday 660 is at a high level 660a, the positive power supply voltage for electrode voltage Vext 670 is at a high positive voltage level 670a, the negative power supply voltage for electrode voltage Vsup 690 is at a low negative voltage level 690a, and Vext trigger control signal 680 provides a switch closed 680a signal to switch 504, which produces a high Vext level 670a. At time 0.0, an arc occurs on the high voltage supply (e.g., Va 620), for example, at the Vext electrode, and the Vext 670 and Vsup 690 voltages quickly drop to zero, for example, as shown at 670b and 690b, respectively. In response, the current detected by CT 506, for example, is received by trigger control circuit 508 and provides a switch open 680b signal on Vext trigger control signal 680 to control HVHS switch 504 to open, which produces a low Vext level 670b. In addition, the detected faraday current l-faraday 660 drops to a low current level 660b. With the HVHS switch now open, and after about 0.3 ms, the Vext trigger control signal 680 returns to the 680a level indicating that the arc has been extinguished, and Vext trigger control signal 680 controls the HVHS switch to re-close, and in response Vext 670 returns to the 670a level. Thereafter, at around 0.6 ms, and with the arc extinguished, the supply voltage at the load begins to recover enough for Vsup 690 to recover to the Vsup 690a level again, and shortly thereafter at about 0.65ms-0.7 ms the beam current recovers as indicated by I-faraday 660 recovering back to the 660a level. Thus, it is shown that the arc quench circuit of the present invention is able to quench an arc in the high voltages electrodes of an ion implanter, for example, and minimize the length of an ion beam glitch to about 0.7 ms. FIG. 7 illustrates a simplified schematic representation of an exemplary arc quenching circuit 700 used in an ion implanter in accordance with several aspects of the present invention. Arc quenching circuit 700 is similar in several ways to that of FIGS. 1, 4 and 5, and as such need not be completely described again for the sake of brevity. Circuit 700 utilizes HVHS switches (A, B, and C) 704 (e.g., a series stack of MOSFET transistors) in three separate high voltage power supplies (Vext 703, −Vsup1 731, and −Vsup2 732) of the ion implanter. Arc quenching circuit 700 also comprises current transformers (CT1, 2, and 3) 706 for detecting current surges in each respective high voltage supply, and received by trigger control circuits 708 to control switches A, B, C 704 to open upon detection of the current surge indicative of an arc 725 at the respective ion beam electrode, for example, extraction electrode or arc slit 720, suppression electrodes 721 and 722, or ground electrodes 724. As illustrated, and in accordance with one aspect of the present invention, each independent electrode supply (e.g., Vext 703, −Vsup1 731, and −Vsup2 732) may independently arc to ground or another electrode, thus each HV supply may be protected by another such HVHS switch. Arc quenching circuit 700 further comprises arc protection circuits 715 having a current limiting resistor (R1, 2, and 3) 712, filter capacitor (C1, 2, and 3) 714, and flyback diode (D1, 2, and 3) 716 to protect the HVHS switches 704 from switching transients and other such overvoltage damage induced by reactive components of the circuits associated with each HV supply. Circuit 700 also utilizes a synchronization circuit 740 to sequence and synchronize the reapplication of the supply voltage to each of three respective high voltage electrodes 720, 721, and 722. For example, it may be determined that synchronization circuit 740 should re-close switches B and C 704 after re-closing switch A. Further, synchronization circuit 740 may provide time delays appropriate for reapplication of each individual HV supply. Any other sequence or timing relationships between the supplies is anticipated, including multiple switch reapplications and/or re-openings with any number of HVHS switches connected in series or parallel with each other or with each HV supply. It will be appreciated that self-adaptive switching and synchronization controls, can be used as a variation of the synchronization circuit 740, within the context of the present invention, wherein changing currents, voltages, infrared or other wavelengths of light energy, or other such changes associated with or indicative of an arc 725, are monitored and used to adjust the sequence and/or timings of the synchronization to compensate or further mitigate such arc induced supply variations. It will also be appreciated that the HVHS switches can be switched at one or more particular frequencies to modulate or otherwise provide dynamic pulse width control of the several electrode voltages, and/or the beam current in response to the detection of an arc. In addition to the detection and quenching of electrode arcs, the HV power supply modulation may also be provided in response to some known non-uniformity in the system (e.g., where a particular beam current results in a predictable non-uniformity). It may also be appreciated that while one use of such modulation is to achieve a uniform dosage on a wafer, it could be used to achieve any predetermined dopant profile, where uniformity is a subset of the general case. It is further appreciated that arc quenching circuit of the present invention may be utilized prior to the implantation as well as during implantation. Alternately, the beam current can be monitored to control the arc quenching circuit or to otherwise regulate a relatively constant beam current in response to HV supply variations when electrode arcing occurs. FIG. 8 illustrates an exemplary protection circuit 810 such as may be used across or in series with a HVHS switch 804 to absorb energy from reactive elements external to the respective HV switch 804, and to limit an over-voltage across the switch in accordance with one or more aspects of the present invention. The protection circuit 810 also protects the switch 804 and other associated components by dampening any ringing induced by switching transients from the HVHS switch 804. Protection circuit 810 is similar to the protection circuit 110 of FIG. 1 and 510 of FIG. 5. Protection circuit 810 comprises a series capacitor Cs connected in series with a resistor Rs, the protection circuit 810 being wired in parallel with a HVHS switch 804. The HVHS switch 804 comprises a HVHS switch (e.g., a series stack of MOSFET transistors) and a diode Dp connected in parallel with the switch. The HVHS switch 804 may be provided, for example, with or without the parallel diode Dp. It will be appreciated in the context of the present invention that two or more such HVHS switches may be connected in series or parallel with each other or with a HV supply to quench an arc that occurs in association with an ion source, an ion implanter, or any other such equipment utilizing high voltage power supplies, for example. FIG. 9A illustrates a simplified diagram 900, of normal 2D wafer scanning motion between an exemplary ion beam 130 and a wafer 910, wherein the wafer 910 is moved simultaneously in a vertical direction 920 and a horizontal direction 930 by a motion control system, for example. The vertical scan motion 920 illustrated, generally may occur more slowly than the horizontal scan motion 930. The resulting scan motion provided by the exemplary motion control system produces a compound motion which has a somewhat diagonal scan vector across the wafer 910. These scans repeat each horizontal row and over-scan the edge of wafer 910 each horizontal scan until a wafer exchange position 935 is achieved past the edge of the wafer 910 at one extreme end of the horizontal scan 930 and the vertical scan 920. A first scan (solid line) 936 of the ion beam 130, indicates an exemplary path starting from a first scan starting point (A) 940 and proceeds simultaneously through multiple horizontal scans motions and a vertical scan to a first scan end point (B) 945. A second scan (dotted line) 937 indicates an exemplary return path of the ion beam 130 (or wafer), which traverses to a second scan start point (C) 950 and proceeds through other multiple horizontal scan motions and a return vertical scan to a second scan end point (D) 955 and returns to the first scan start point A 940. Scanning continues in the manner until adequate dose is implanted in the wafer. Thereafter, the wafer scanning and implanting is completed and the wafer motion returns to the wafer exchange position 935 at the extreme end of the horizontal scan 930 exchange and process another wafer 910. The amount of over-travel of the relative motion between the ion beam 130 and the wafer 910 is also known as the overscan region 960. FIGS. 9B-9G illustrate 2D wafer scan motion between an exemplary ion beam 130 and a wafer 910 similar to that of FIG. 9A and two methods of re-applying the ion beam (repainting) after loss of the ion beam due to a disruption in the high voltage power supply caused by the occurrence of an arc, for example, at the high voltage electrodes. When an arc is encountered during a wafer scan, the ion beam 130 may be disrupted either for a longer duration creating a glitch stripe 972 of FIGS. 9B-9D, or briefly creating a glitch hole 987 of FIGS. 9E-9G (a short disruption) each glitch having a corresponding loss of dose. These areas of dose loss may then be repainted in accordance with one or more aspects of the present invention. In FIGS. 9B-9D, for example, when an arc occurs during implantation, the encoder positions of the horizontal and vertical motors driven by the motion control system may be monitored, so that the initial detection position X and final detection position Y associated with the arc may be recorded. The arc quenching circuit would function as previously described to quench the arc between initial detection position X and final detection position Y. Subsequent to the arc detection and quenching, the wafer scanning and ion implantation may continue as usual, or may proceed according to the example stripe repainting method as illustrated in FIGS. 9B-9D and as follows. For example, wafer scanning proceeds as usual until an arc is detected at the initial detection position X, where the HVHS switch (e.g., 504, 704) is opened to disable the ion beam 130. Meanwhile, wafer scanning continues to the final detection position Y where motion decelerates to a stop. At Y the HVHS switch is closed again to re-enable the ion beam 130 for a repaint along the glitch stripe 972 as shown in FIG. 9C. As the wafer accelerates back up to speed horizontally, the vertical motion is reversed to repaint 977 the ion beam along glitch stripe 972 from the final detection position Y back to the initial detection position X, where the HVHS switch is opened again to disable the ion beam 130 until an end of the horizontal scan Z is encountered. Wafer scanning then moves the wafer from Z to Y and may or may not stop at Y, but with the HVHS switch remaining in the OFF or disabled state. Then, at Y, the HVHS switch is closed again to enable the ion beam. Then, the normal implantation operations continue until all the required scans are complete as previously described for FIG. 9A. Similarly, in FIGS. 9E-9G, when an arc occurs during implantation, the wafer scanning and ion implantation may proceed according to the following exemplary hole repainting method illustrated and described, because the ion beam can not only be disabled quickly by the HVHS switch, but also can be enabled quickly by the HVHS switch. With the use of HVHS switch, the glitch will only leave a hole of less doped area on the wafer, and the hole of the start and ending position (initial and final position) can be recorded by a control computer. Therefore, the repaint can be done after the completion of the normal wafer scans. Thus, this repaint is so called the hole repaint. For example, wafer scanning proceeds as usual until an arc is detected at an initial detection position W, where the HVHS switch (e.g., 504, 704) is opened to disable the ion beam 130 for a predetermined period of time, if, example, the glitch length exceeds about 1 ms. For example, the predetermined period of time may be about 5-20 ms to account for all anticipated arc durations. After the predetermined time period, the HVHS switch is closed to enable the ion beam at X, the final detection position. Meanwhile, wafer scanning motion continues as usual to a first row scan position Y and normal implantation operations continue until all the required scans are complete as shown in FIG. 9F, and as previously described for FIG. 9A. Then, to perform a hole repaint 997 between W and X, the HV switch is opened to disable the ion beam 130, and the wafer is moved to a row repaint position Z corresponding to the beginning of the row wherein the glitch hole 987 occurred. The wafer scan moves toward Y, and the HVHS switch is closed again at W to re-enable the ion beam 130 for a repaint 997 along the glitch hole 987 as shown in FIG. 9G. When the wafer reaches X, the HV switch is opened to disable the ion beam 130. When the scan reaches the end of the horizontal scan Y the scan motion may stop and the hole repaint method is complete. The advantage of doing a hole repaint is, that the wafer scanning will not be interrupted during the normal scan. Thus, the overall implantation productivity can be higher than that of the stripe repaint. In addition to the stripe and hole repaint methods described above, for example, at the end of a horizontal scan 930 and vertical scan 920, at the wafer exchange position 935, or at the end of a particular row scan, a repaint process may be initiated, for example, by a modified trigger control circuit (e.g., 508 of FIG. 5, or 708 of FIG. 7), or a modified synchronization circuit (e.g., 740 of FIG. 7), as previously described. Such a modification would further make a trigger control circuit operable to receive a command (see 508 of FIG. 10A, and 1108 of FIG. 11) from the motion control system or the implanter computer to initiate the repaint process. The repaint process initially may include turning OFF the ion beam 130 via the HVHS switch (e.g., 504, 704) for the ion source (e.g., 120 of FIG. 5) and the extraction/suppression electrodes 720, 722, for example. The ion beam 130 or wafer 910 may then be moved to a first horizontal and vertical scan position (e.g., W of FIG. 9E) where the arc was initially detected, for example, by first returning the beam to the beginning of the row (horizontal) Z wherein the arc was initially detected. In this way, the scan motion can begin a row as usual prior to encountering the position of the initial arc detection (e.g., W). Thus, the wafer will then be fully accelerated up to the same speed as that which was present when the arc was initially detected. Once at the position of initial arc detection W, the ion beam 130 is enabled by closing the HVHS switch (e.g., 504, 704), while the wafer 910 is horizontally and vertically scanned 930 and 920, respectively, in front of the ion beam 130. When a second horizontal and vertical scan position (e.g., X) associated with the final detection of the arc is encountered, the HVHS switch 504, 704 is opened again to disable the ion beam 130. Alternately, the ion beam may be repainted 977 in the opposite direction of the initial scan direction, for example, starting from the final detection position Y of FIG. 9C, and proceeding in a scan direction to the initial detection position X. This direction, however, may not reproduce the dose losses as precisely as those achieved by repainting the ion beam 130 in the same direction as the original scan direction (see the original direction indicated by the arrow for short glitch hole 987 and long glitch stripe 972). FIG. 10A illustrates an exemplary arc quenching circuit 1000 utilized in association with the high voltage supply 503 of an ion source 120 such as may be used in an ion implantation system (e.g., 200 of FIG. 2) in accordance with the present invention. The trigger control circuit 508 has an external trigger or control input 1010 for externally controlling the triggering of the HVHS switch 504, for example, by a computer, or by another such device. FIG. 10B illustrates the effect of using external control of the HV switch to control the ion beam on and off. For example, external control may be provided in response to a faraday cup current (Ifaraday) waveform 1060 that provides a proportionate sample of the ion beam current (Ibeam). The faraday cup current 1060 is used to provide a trigger voltage 1070 for the external trigger input 1010 to the exemplary arc quenching circuit 1000 of FIG. 10A, such as may be used in accordance with one or more aspects of the present invention. For example, at a given instant of time, if the ion beam 130 is expected to be present at the faraday cup or on the wafer, a high current 1060a is produced, and a low level trigger voltage 1070a is applied. However, if the beam is not present at the faraday cup or on the wafer, faraday current 1060b is zero, and a high level trigger voltage 1070b is applied, which when provided to external trigger input 1010, switches OFF HVHS switch 504 to quench the arc. In this way, the switch can be used to artificially shut off or start the ion beam very quickly. FIG. 11 illustrates an exemplary arc quench controller 1100 having an arc quench circuit 1102, is used in an ion implanter (e.g., 200 of FIG. 2), utilizing a HVHS switch 504 between a high voltage supply 503 and an electrode (not shown) of the implanter 200. Arc quench controller 1100 is similar to the arc quenching circuits of FIGS. 5, 7, and 10A, and as such need not be completely described again for the sake of brevity. The arc quench controller 1100 utilizes a switch control circuit 1108 to sequence, control, and synchronize the reestablishment of the current and voltage to one or more electrodes from other switches 1110, to determine the duration of arc loss by long glitch detector 1120, and coordinate repaint commands 1130 from a motion control system 1150 or forced switch control commands 1140 from the ion implantation system in accordance with one or more aspects of the present invention. Optionally, the other HVHS switches 1110 may be from other electrode supplies of the same ion implanter, or they may be from other HVHS switches of other similar arc quenching circuits (AQCs), which are not shown. These switches need to be synchronized to ensure the desired order and timing for opening and closing the switches 1110. As indicated above, the duration of a glitch, or loss of the ion beam, may be detected by a long glitch detector 1120, to determine if the glitch is long enough to require the repaint procedure as described above. Conversely, if a glitch is short enough, a determination may be made to ignore the loss. Such a determination level may be set, for example, by the end user of the ion implantation system. If a long glitch (e.g., 972 of FIG. 9B) is detected and a repaint of the ion beam 130 is required, the HVHS switch 504 could be forced OFF during the glitch to quench the arc. Further, when the long glitch 972 is detected by long glitch detector 1120, the repaint process may be initiated upon assertion of the repaint command 1130 from the scan motion control system 1150. With the use of HVHS switches, the long glitch can be defined as the duration of the glitch longer than about 1 ms. During this repaint process, the HVHS switch control is forced ON/OFF, as illustrated in FIG. 10B, in response to the repaint command 1130, and in response to the positions achieved by the motion control system 1150, and as described previously. In addition because the HVHS switches 504 are present in the circuit of the present invention to quench arcs, the implanter system is also provided with the ability to simply turn the ion beam 130 ON or OFF at will, either manually with a switch or by way of command 1140 from one of the implanters control systems, its computer, or by an external input. As FIG. 10B illustrated, it may be particularly beneficial to be able to turn the beam 130 ON/OFF, for example, when loading or unloading a new wafer 910, during other types of wafer exchange, at the start/end of each wafer scan, or in the over-travel regions 960 of each row scan of a wafer 910. This is so called beam duty factor reduction. This is to say that by disabling the beam via the HVHS switches, the total time required of the ion beam in the beam line and the wafer process chamber is reduced. Accordingly, the glitch quench controller 1100 of the present invention facilitates reducing the “beam duty factor” should reduce the particle count on a wafer, because the beam will be used to a greater percentage usefully on the wafer 910 and less on the other surfaces of the implanter adjacent to the wafer 910 (e.g., in the over-travel regions 960). Although the glitch quench circuits and glitch quench controller of the invention has been illustrated in association with a HV power supply for an ion source and an extraction electrode, it will be appreciated that such circuits may also be used in association with the other HV supplies and electrodes of an ion implanter, or other such ion sources and accelerators, including other HV applications subject to HV arcing and are anticipated in the context of the present invention. One aspect of the present invention provides a method of quenching an arc and repainting the ion beam is presented and described. One implementation of the present invention effectively quenches such high voltage arcs which occur at an electrode of an ion implanter and by opening a high voltage high speed switch wired in series between the electrode and a high voltage supply which provides the electrode potential to control the ion beam. When the arc is then extinguished, the electrode potential is restored, before the HV supply has had a chance to fully discharge. Thereafter, any dose losses experienced during the arcing may be restored by returning the ion beam/wafer to the location where the arc occurred, repainting the ion beam over such areas, and toggling the ion beam ON/OFF during the repaint operation using the same HVHS switch. One such method 1200 is illustrated in FIG. 12, representing an exemplary method for quenching an arc in an ion implanter, using an arc quenching circuit (e.g., 500 of FIG. 5, 700 of FIG. 7, and 1000 of FIG. 10A, or the arc quenching controller 1100 of FIG. 11) of the present invention in accordance with several aspects of the present invention. Although the example method 1200 is illustrated and described hereinafter as a series of acts or events, it will be appreciated that the present invention is not limited by the illustrated ordering of such acts or events. In this regard, some acts may occur in different orders and/or concurrently with other acts or events apart from those illustrated and/or described herein, in accordance with the invention. In addition, not all illustrated steps may be required to implement a methodology in accordance with the present invention. It is further noted that the methods according to the present invention may be implemented in association with the wafers, wafer cassettes, wafer sensor, wafer handling system, and modeling system illustrated and described herein as well as in association with other apparatus and structures not illustrated. Method 1200 comprises an example arc quenching method that may be used to extinguish an arc that may occur at an ion beam controlling electrode of an ion implanter (e.g., 200 of FIG. 2), using an arc quenching circuit similar to circuit 1000 of FIG. 10A. For example, optionally, a wafer 910 may be in the process of being horizontally 930 and vertically 920 scanned at 1210 with an ion beam 130 (either by the wafer or the ion beam moving). At 1220, a current or voltage change (e.g., Va 620 of FIG. 6A) associated with the arc is detected (e.g., by CT 506 of FIGS. 5, 10A, and 11, or by a faraday cup 244 to provide l-faraday 1060 of FIG. 10B associated with the ion beam current Ibeam) at an electrode (e.g., extraction electrode 208 of FIG. 2) of the implanter 200. When an arc then occurs during implantation, the encoder positions of the horizontal and vertical motors driven by the motion control system (e.g., 1150 of FIG. 11) of the implanter 200 may optionally be monitored at 1230, so that the initial detection position (e.g., W of FIG. 9E) and final detection position (e.g., X of FIG. 9E) associated with the arc may be recorded for an optional repaint operation (e.g., method 1300 of FIG. 13). At 1240 the duration of such an arc may be optionally detected (e.g., by detector 1120 of FIG. 11) and used to determine if a subsequent repaint is desired. After the detection of an arc, the arc is quenched at 1250, by opening a HVHS switch 504 connected between a high voltage supply 503 and the electrode 208 of the implanter 200, thereby interrupting the power provided to the electrode 208 and quenching the arc. When the arc is no longer detected, the HVHS switch 504 is again closed at 1260 to reconnect the high voltage supply to the electrode and to re-establish the ion beam. Optionally, the wafer scanning and ion implantation may continue as usual, for example, until the end of a row scan, or the end of a vertical scan 920, for example, at the wafer exchange position 935, or until the end of all the scans anticipated for a wafer 910. Then, a repaint process may be initiated, for example, by switch control unit 1108, or a modified synchronization circuit (e.g., 740 of FIG. 7), as previously described. The switch control circuit 1108 is operable to receive a command (see 508 of FIG. 10A, and 1108 of FIG. 11) from the motion control system 1150 or the implanter computer to initiate the repaint process. FIG. 13 illustrates an exemplary method 1300 for repainting the ion beam 130 to recover dose loss due to arcing in an ion implanter 200, for example, using the arc quenching controller 1100 of FIG. 11 in accordance with one or more aspects of the present invention. Method 1300 comprises an example ion repainting method that may be used to restore the ion dose loss during arcing in an ion implanter (e.g., 200 of FIG. 2), using an exemplary arc quenching controller 1100 of FIG. 11. After an arc has been extinguished, and at the end of a horizontal scan 930, or at the end of a vertical scan 920, for example, the repaint process may be initiated. Initially, at 1310, the ion beam 130 is again turned OFF, for example, using the HVHS switch (e.g., 504, 704) to the ion source (e.g., 120 of FIG. 5) and the extraction/suppression electrodes 720, 722, for example. The ion beam 130 or wafer 910 is then moved at 1320, to a first horizontal and vertical scan position (e.g., W of FIG. 9E) where the arc was initially detected, for example. This movement to the initial scan position W may be done by first returning the beam to the beginning of the row Z wherein the arc was initially detected. In this way, the scan motion can begin a row as usual prior to encountering the position of the initial arc detection (e.g., W of FIG. 9E). Thus, the wafer will then be fully accelerated up to the same speed as that which was present when the arc was initially detected. Once at the position of initial arc detection (e.g., W of FIG. 9E), the ion beam 130 is also enabled at 1320 by closing the HVHS switch (e.g., 504, 704). At 1330, the wafer 910 is horizontally and vertically scanned 930 and 920, respectively, in front of the ion beam 130 until a second horizontal and vertical scan position (e.g., X of FIG. 9G) associated with the final detection of the arc is encountered, and the HVHS switch 504, 704 is opened again to disable the ion beam 130. Thereafter, other such arc detection holes 987 or stripes 972 which occurred during the ion implantation may also be repainted in a similar manner, either separately in multiple scans or collectively in a single scan as needed. Further, it is appreciated that the ion beam may be repainted in the opposite direction of the initial scan direction, for example, starting from the final detection position, and proceeding with the repaint toward the initial detection position. The HVHS switches are basically applied to the extraction systems of any ion sources. It will be appreciated that the aspects described herein are equally applicable to other ion sources including those that provide primary electron beam current in “soft ionization” ion sources, RF or microwave power in RF or microwave ion sources, as well as to non-arc discharge sources. Although the invention has been illustrated and described above with respect to a certain aspects and implementations, it will be appreciated that equivalent alterations and modifications will occur to others skilled in the art upon the reading and understanding of this specification and the annexed drawings. In particular regard to the various functions performed by the above described components (assemblies, devices, circuits, systems, etc.), the terms (including a reference to a “means”) used to describe such components are intended to correspond, unless otherwise indicated, to any component which performs the specified function of the described component (i.e., that is functionally equivalent), even though not structurally equivalent to the disclosed structure, which performs the function in the herein illustrated exemplary implementations of the invention. In addition, while a particular feature of the invention may have been disclosed with respect to only one of several implementations, such feature may be combined with one or more other features of the other implementations as may be desired and advantageous for any given or particular application. Furthermore, to the extent that the terms “includes”, “including”, “has”, “having”, “with” and variants thereof are used in either the detailed description or the claims, these terms are intended to be inclusive in a manner similar to the term “comprising”. Also, the term “exemplary” as utilized herein simply means example, rather than finest performer. |
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054066000 | abstract | The present invention provides a cask (10) for transport and short-term storage of spent nuclear fuel. The cask 10 includes a structural shell (14) defining a cavity (40) for receiving spent nuclear fuel. The shell is formed from an upper shell portion (16) formed of a first metal-and a lower shell portion (18) formed from a second metal. The first metal utilized to form the upper shell portion has a higher load bearing strength than the second metal utilized to form the lower shell portion. A bearing surface is defined on the upper shell portion by trunnions (30) mounted within sleeves (32) secured to the upper shell portion. The trunnions (30) each define a bearing projection (160) that is engageable to enable hoisting of the cask, with the tensile and shear loads of hoisting the cask being transferred from the trunnions to the trunnion mounting sleeves, and thus to the upper shell portion of the structural shell. The cask further includes a bottom closure plate (20) secured to the bottom end of the shell, and a top closure plate releasably securable to the top end of the shell. A neutron radiation absorbing shield jacket (28) is formed about the exterior of the structural shell. |
abstract | Fourier telescopes permit observations over a very broad band of energy. They generally include synthetic spatial filtering structures, known as multilayer grids or grid pairs consisting of alternate layers of absorbing and transparent materials depending on whether neutrons or photons are being imaged. For hard x-rays and gamma rays, high (absorbing) and low (transparent) atomic number elements, termed high-Z and low-Z materials may be used. Fabrication of these multilayer grid structures is not without its difficulties. Herein the alternate layers of the high-Z material and the low-Z material are inserted in a polyhedron, transparent to photons of interest, through an open face of the polyhedron. The inserted layers are then uniformly compressed to form a multilayer grid. |
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051376830 | abstract | An organometallic chromium compound in the gaseous phase is brought into contact with a substrate consisting of the inner surface of the zirconium alloy tubular cladding (5) of the fuel element, or the outer surface of the pellets of fuel material, the substrate being kept at a temperature between 300.degree. and 600.degree. C. The organo metallic compound may consist of chromium acetylacetonate. The process enables a chromium oxide coating to be obtained inside the tubular cladding (5) and/or on the outer surface of the nuclear fuel pellets. Pellet/cladding interaction is thus prevented or limited when this fuel element is used in the reactor. |
055132308 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is seen a welding apparatus having a copper electrode 2 which is assembled from two identical halves, with a through bore 3 in which a fuel rod cladding tube 4 being filled with nuclear fuel and made of a zirconium alloy is located. The two identically constructed parts of the copper electrode 2 surround the cladding tube 4 located in the through bore 3 and touch the outer surface of the cladding tube 4 over a large surface area. A carrier body 6 for a counter electrode which is also provided, is displaceable relative to the copper electrode 2 in the direction of a longitudinal axis 5 of the cladding tube 4, and includes a copper sheath 7 disposed coaxially with the cladding tube 4 and therefore with the duct 3 in the copper electrode 2. A seal plug, locking plug or stopper 8 for the fuel rod, which is likewise formed of a zirconium alloy, is loosely inserted into the copper sheath 7. The copper electrode 2 has a cylindrical step or shoulder 9 formed therein at an end of the through bore 3 facing toward the copper sheath 7. The cylindrical step 9 has a diameter which is greater than the diameter of the through bore 3. The step 9 forms a cylindrical void, into which the end of the cladding tube 4 protrudes from one end and into which the end of the seal plug 8 protrudes from the other end. Dot-dash lines in FIG. 2 indicate the original shape of the cladding tube 4 and the seal plug 8. In its original shape, the seal plug 8 has an external cone 10 at an end surface, which tapers toward the longitudinal axis 5 of the cladding tube and therefore also toward the longitudinal axis of the seal plug 8, which coincides with the longitudinal axis 5. One end of the cladding tube 4 rests with its inner edge on the cone 10. In order to weld the cladding tube 4 to the seal plug 8, an electrical current source is connected to the copper electrode 2 and to the copper sheath 7. At the same time, through the use of the carrier body 6, the seal plug 8 is pressed against the cladding tube 4, and the material at the point of contact between the seal plug 8 and the cladding tube 4 is upset by the amount of the forward feed 11 (upset distance). In FIG. 1, the initial condition of the cladding tube 4 and seal plug 8 at the beginning of the welding process is shown above the longitudinal axis 5 of the cladding tube 4 shown in dot-dashed lines. Below this longitudinal axis 5, the final state can be seen, after completion of the welding process and disconnection of the copper electrode 2 and the copper sheath 7 from the current source. As FIG. 1 shows, in this final state the seal plug 8 engages the cladding tube 4, and at a transition point between the cladding tube 4 and the seal plug 8, an annular bead 12 with a cylindrical outer jacket surface 13 is present on the outer surface of the cladding tube 4. FIG. 2 shows that an annular outwelling, outflow or squeezing out of material 14 is formed inside the cladding tube 4. Both the annular bead 12 and the outwelling of material 14 are formed of the material of the original cladding tube and the original seal plug, which in the present case is zirconium alloy. Due to the two seams or joints that are located between the two parts of the copper electrode 2, there may be two humps 17 on the outer jacket surface 13 of the annular bead 12, each of which extends along a jacket line of the jacket surface 13. The annular bead 12 as well as the outwelling of material 14 solidify again from a welding melt and have the microscopic structure shown in FIG. 3. The flow of the material caused by the welding is indicated by arrows 16. As can be learned from FIG. 3, there is a depression 15 of the cladding tube 4, as seen in the encompassing direction relative to the longitudinal axis 5, which is located in the cylindrical jacket surface 13 of the annular bead 12. This depression 15 contains the material having the same microscopic structure as in the cladding tube 4, that is the zirconium alloy of the cladding tube 4. On the inside of the cladding tube 4, an internal lining 18 of high-purity zirconium can be seen. This corrosion-sensitive lining continues on the surface of the outwelling of material 14 and is not part of the material solidified from the welding melt and thus has not penetrated as far as the outside. The cylindrical outer jacket surface 13 of the annular bead 12 can remain mechanically unmachined. As a result, not only is there a savings in production costs, but damage to the cladding tube 4 at the annular bead 12 is also avoided. Moreover, machining chips, which in the case of a zirconium alloy could even self-ignite very easily, are avoided. |
description | This application claims the benefit of priority of U.S. provisional patent application No. 62/402,460, titled “NUCLEAR EXCITATION TRANSFER VIA PHONON-NUCLEAR COUPLING,” filed on Sep. 30, 2016, which is incorporated herein in its entirety by this reference. The present disclosure relates to condensed matter and nuclear sciences. More particularly, the present disclosure relates to excitation transfer. This summary is provided to introduce in a simplified form concepts that are further described in the following detailed descriptions. This summary is not intended to identify key features or essential features of the claimed subject matter, nor is it to be construed as limiting the scope of the claimed subject matter. In at least one embodiment, an apparatus includes: a support; a radioactive source on the support, the radioactive source comprising nuclei; and an excitation element coupled to the support. Upon activation of the excitation element, radiation emission from the radioactive source is reduced. In at least one example, the excitation element includes a vibration source. In at least one example, excitation is transferred from nuclei of the radioactive source to nuclei of the support. In at least one example, the excitation transfer occurs in bulk from multiple nuclei of the radioactive source. In at least one example, the excitation transfer causes emissions from the support. These descriptions are presented with sufficient details to provide an understanding of one or more particular embodiments of broader inventive subject matters. These descriptions expound upon and exemplify particular features of those particular embodiments without limiting the inventive subject matters to the explicitly described embodiments and features. Considerations in view of these descriptions will likely give rise to additional and similar embodiments and features without departing from the scope of the inventive subject matters. Although the term “step” may be expressly used or implied relating to features of processes or methods, no implication is made of any particular order or sequence among such expressed or implied steps unless an order or sequence is explicitly stated. Any dimensions expressed or implied in the drawings and these descriptions are provided for exemplary purposes. Thus, not all embodiments within the scope of the drawings and these descriptions are made according to such exemplary dimensions. The drawings are not made necessarily to scale. Thus, not all embodiments within the scope of the drawings and these descriptions are made according to the apparent scale of the drawings with regard to relative dimensions in the drawings. However, for each drawing, at least one embodiment is made according to the apparent relative scale of the drawing. These descriptions relate to novel and non-obvious advancements in Condensed Matter Nuclear Science. Experiments have provided evidence of a number of observations: Heat energy, thought to be a nuclear effect, but without commensurate energetic nuclear radiation; He-4 commensurate with and correlated with heat energy; Tritium production; Collimated x-ray and gamma emission. An explanatory theory as described herein uses developed models to account for heat energy and other anomalies. The approach is based, according to inventive embodiments, on the notion of massive up-conversion and down-conversion. According to inventive embodiments, conversion occurs between large nuclear quanta and large numbers of low-energy vibrational quanta. FIG. 1 is a down-conversion diagram according to at least one embodiment. A simplest conceptual approach may be used, but math favors intermediate steps where many metastable nuclei with a lower energy transition are excited. FIG. 2 is an up-conversion diagram according to at least one embodiment. With respect to the simplest possible up-conversion experiment, this mechanism is proposed as responsible for collimated x-ray emission in the Karabut experiment. Parts of the model according to at least one embodiment follow. One part of the theoretical approach involves models for two-level systems coupled with a highly-excited oscillator. Prior models that may be known in the literature to up-convert and down-convert do not anticipate this. Prior approaches don't expect (macroscopic) phonon exchange with (subatomic) nuclear transitions. Relativistic interaction for this coupling is proposed. Relativistic coupling are known in the literature, but in other disparate non-analogous applications. An approach according to embodiments herein: Includes a model that results in and is capable of describing anomalies systematically; Theory is connected with experiment, one piece at a time; Includes focus on collimated x-ray emission as a test problem in recent years; and Includes phonon-nuclear coupling. Phonon-Nuclear Coupling—Relativistic Problem: H ^ = ∑ j α j · c [ p j - q j A ( r j ) ] + ∑ j β j m c 2 + ∑ j < k V ^ jk ( r k - r j ) + ∑ j q j Φ ( r j ) MR = ∑ j mr j P ^ = ∑ j p ^ j Q = ∑ j q j ξ j = r j - R π ^ j = p ^ j - P N H ^ = ∑ j α j · c [ P ^ N + π j - q j A ( R + ξ j ) ] + ∑ j β j m c 2 + ∑ j < k V ^ jk ( ξ k - ξ j ) + ∑ j q j Φ ( R + ξ j ) Foldy-Wouthuysen Type of Rotation: H ^ ′ = e i S ^ ( H ^ - i ℏ ∂ ∂ t ) e - i S ^ = H ^ + i [ S ^ , H ^ ] - 1 2 [ S ^ , [ S ^ , H ^ ] ] + … - ℏ ∂ S ^ ∂ t - i 2 [ S ^ , ℏ ∂ S ^ ∂ t ] + … S ^ = - i 1 2 M c ∑ j β j α j · [ P ^ j - Q A ( R ) ] Rotation works on the center of mass degrees of freedom.Nucleus as a Particle: H ^ ′ = P ^ - Q A 2 2 M 1 N ∑ j β j + Q Φ - ℏ Q 2 M 1 N ∑ j β j ∑ j ^ · B - ℏ 2 Q 8 M 2 c 2 ∇ · E + ℏ Q 8 M 2 c 2 ∑ j ∑ j ^ · [ ( P ^ - Q A ) × E - E × ( P ^ - Q A ) ] Internal Nuclear Structure: + ∑ j β j m c 2 + ∑ j α j · c π ^ j + ∑ j < k V ^ jk + ∑ j [ q j Φ ( R + ξ j ) - Q N Φ ( R ) ] - ∑ j α j · c [ q j A ( R + ξ j ) - Q N A ( R ) ] + ∑ j β j ( P ^ - Q A ) · π ^ j M + 1 2 M c ∑ j < k [ ( β j α j + β k α k ) · ( P ^ - Q A ) , V ^ jk ] + … H ^ ′ = P ^ 2 2 M + nucleus as a particle ∑ j β j m c 2 + ∑ j α j · c π ^ j + ∑ j < k V ^ jk internal nuclear structure + a · c P ^ + … Where a · c P ^ = { ∑ j β j π ^ j M + 1 2 M c ∑ j < k [ ( β j α j + β k α k ) , V ^ jk ] } · P ^ ↑ (—accounting for coupling between center of mass motion and internal nuclear degrees of freedom) Phonon-nuclear coupling is present in relativistic models. Interaction can be rotated out for a composite in free space. For interacting nuclei, for some examples it is inconvenient to rotate it out. Examples where this is true are connected with the anomalies. Homonuclear Diatomic Molecule: Motivation for diatomic molecule: Interested in simplest possible version of problem involving phonon-nuclear coupling; Work with nuclear transitions in two nuclei (fewest possible); Work with identical nuclei (energy levels degenerate); Make use of diatomic molecule (simplest system that can vibrate); Would like electric dipole (E1) transition if possible; and would like lowest energy nuclear transition, to maximize effect. Low Energy Nuclear Transitions: Excited stateNucleusenergy (keV)Half-lifeMultipolarity201Hg1.564881 nsM1 + E2181Ta6.246.05 μsE − 1169Tm8.410174.09 nsM1 + E283KR94,051154.4 nsM1 + E2187Os9.752.38 nsM1(+E2)73Ge13.28452.92 μsE257Fe14.412998.3 nsM1 + E2 Low Energy E1 Candidates: isotopeT1/2 (ground)E(keV)T1/2 (excited)MultipoleTa-181Stable6.2376.05 μsE1Dy-161Stable25.65129.1nsE1Pa-2291.5 d11.6(not known)E1Ac-22721.77 y27.36938.3nsE1 (+M2)Ta-1791.82 y30.71.42μsE1Ra-22514.9 d31.562.1nsE1Ir-19011.78 d36.154>2μsE1Th-22718.70 d37.063(not known)E1 FIG. 3 shows a homonuclear diatomic Ta2. Model for Diatomic Molecule: H ^ = M 1 c 2 + M 2 c 2 nuclear states + P 1 2 2 M + P 2 2 2 M + U ( R 2 - R 1 ) molecule + a 1 · c P ^ + a 2 · c P ^ phonon - nuclear coupling A model for excitation transfer according to at least one embodiment is shown in FIG. 4. Indirect coupling coefficient—Carry out a calculation for the vibrational ground state, and for degenerate nuclear states: 〈 I * M 1 , IM 2 , n = 0 V ^ 12 IM 1 , I * M 2 , n = 0 〉 = - μ c 2 ( ℏ ω 0 ) 2 Δ E 2 M 1 M 2 I ( I + 1 ) ( 2 I + 1 ) I * ( I * + 1 ) ( 2 I * + 1 ) ( I * a I ) 2 Equivalent Hamiltonian: H ^ 12 = - μ c 2 ( ℏ ω 0 ) 2 Δ E 2 ( I * a I ) 2 ( I * I ^ I ) 2 ( I 1 · R 12 ) ( I 2 · R 12 ) R 12 2 A homonuclear diatomic Ta2 presents a physics problem for phonon-nuclear coupling. Good analysis of indirect coupling for excitation transfer is provided. Excitation transfer leads to a splitting that may be observable. New splitting is different than electric field gradient quadrupole splitting, closer to, but different than, nuclear spin-spin splitting. Diatomic 57Fe in an Argon Matrix—Analog in Diatomic 57Fe: Mossbauer experiments have been done in diatomic Fe-57. Molecules formed in argon matrix near liquid helium temperature. Mossbauer spectra observed for diatomic Fe2. Large quadrupole splitting due to electric field gradient may be observed. Phonon-nuclear coupling may be observed in diatomic Ta2 Mossbauer process. FIG. 5 is a Mossbauer spectra of the prior art (P H Barrett and T K McNab, Phys Rev Lett 25 (1970) 1601)). Regarding diatomic Ta2: Mossbauer effect has been studied for Ta-181 transition at 6240 eV; Diatomic Ta2 molecule has been observed; Optical measurements have been done on Ta2 in an argon matrix; analogous Mossbauer experiments have not been done for Ta2 in an argon matrix; the ground state of diatomic Fe2 is an electronic spin singlet, but ground state of Ta2 may not be, presenting a challenge. Excitation transfer with 181Ta—Possibility of observing excitation transfer: Elegant observation of phonon-nuclear coupling in 181Ta2, Issues with 201Hg2 at 1565 eV, since not an E1 transition, are considered; and various ways to verify phonon-nuclear coupling are considered. FIG. 6 is an excitation transfer scheme according to at least one embodiment, by which to transfer excitation from one nucleus to another, where there are many others to go to. Excitation transfer with more nuclei gives a more complicated mathematical problem, but includes similar physics as for up-conversion and down-conversion models. If off-resonant loss is different than on-resonant loss, then one would expect an acceleration of excitation transfer effect. This could be observed by looking at different positions in space. FIG. 7 is a schematic of a process apparatus 100, according to at least one embodiment, that includes a radioactive W-181 source 102. FIG. 8 is a decay scheme of the prior art, in which W-181 decays to Ta-181. FIG. 9 is a schematic of the process apparatus 100 of FIG. 7, in which gamma emission emanates from the location of the radioactive W-181 source 102. Excitation transfer—stimulation by vibrations has the potential to cause excitation transfer effect. As excitation transfer occurs, less emission from the location of radioactive source 102 occurs as excitation is transferred to other nuclei which see common excited vibrational modes. Thus emissions are seen from other parts of a support plate 104. FIG. 10 is a schematic diagram of the process apparatus of FIG. 7 in which excitation transfer occurs. If the plate is thick, then most of the radiation would be absorbed internally. Embodiments for excitation transfer include: Putting the W-181 source 102 on surface of Ta-181 support plate 104; Check that 6240 eV emission occurs from location of source; Then vibrate using a vibration source 106 coupled to the support plate 104 to cause excitation transfer; Measure reduction of emission from location of source; Measure emission from locations where no source is present; Effect suggests that loss be different in off-resonant states in order to be a big effect. A candidate for observed anomalies is provided. A potential is to account systematically for all anomalies. A model for massive up-conversion and down-conversion effects is provided. A model for phonon-nuclear coupling is provided. Level shift due to phonon-nuclear coupling in a diatomic molecule may occur. Ta2 is promising. Challenges are present due to the electronic ground state not being a singlet. In an at least one embodiment, excitation transfer occurs in a W-181 source device to produce excited state Ta-181. Excitation transfer is used to move the excitation from the source location to other nuclei. Vibrations are to stimulate excitation transfer effect to reduce emission at source location. Emission is seen from other parts of the plate. Particular embodiments and features have been described with reference to the drawings. It is to be understood that these descriptions are not limited to any single embodiment or any particular set of features, and that similar embodiments and features may arise or modifications and additions may be made without departing from the scope of these descriptions and the spirit of the appended claims. |
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description | Pursuant to 35 U.S.C. § 119(a), this application claims the benefit of earlier filing date and right of priority to Korean Application No. 10-2013-0102649, filed on Aug. 28, 2013; Korean Application No. 10-2014-0036321, filed on Mar. 27, 2014; Korean Application No. 10-2014-0083848, filed on Jul. 4, 2014, the contents of which are incorporated by reference herein in their entireties. 1. Field of the Disclosure This specification relates to a safety system for securing safety of a nuclear power plant, and in particular, to a facility that may decrease the concentration of a radioactive material in a containment by a passive principle when an accident occurs in the nuclear power plant and a nuclear power plant having the same. 2. Background of the Disclosure Depending on the position of installation, nuclear reactors are classified into loop-type reactors (e.g., commercial reactors, Korea) with main components (steam generators, a pressurizer, reactor coolant pumps, etc.) installed outside the reactor vessel and integral reactors (e.g., SMART reactor, Korea) with the main components installed in the reactor vessel. Further, nuclear reactors are classified into active reactors and passive reactors depending on how the safety system is implemented. The active reactors are reactors that use an active component, such as a pump, which is powered by an emergency diesel generator in order to operate the safety system, and the passive reactors are reactors that use a passive component which is powered by a passive force such as gravity or gas pressure in order to operate the safety system. In the passive reactors, the passive safety system may safely maintain the reactors only with a natural force embedded in the system without a safety-grade AC power source such as an emergency diesel generator or an operator's action at least for a time (72 hours) required by the regulations when an accident occurs, and after 72 hours, the passive safety system may be treated by the operator or assisted by a non-safety system. A containment (containment building, reactor building, containment vessel or safeguard vessel) that plays a role as a final protection barrier to prevent radioactive materials from releasing to the external environment are classified into the containment building (or reactor building) formed of reinforced concrete and the containment vessel and safeguard vessel formed of steel depending on the material constituting a pressure boundary. The containment vessel is a large vessel that is designed to have a low pressure like the containment building, and the safeguard vessel is a small vessel designed to be rendered to have a small size and having the higher design pressure. Unless mentioned specially, as used herein, the terms “containment building,” “reactor building,” “containment vessel,” or “safeguard vessel” are collectively referred to as a containment. Various forms of active and passive systems, such as a containment spray system, a containment cooling system, a suppression tank or suppression pool, are put to use in order to decrease the density of radioactive material, the pressure and temperature in the containment at accidents. Hereinafter, such facilities are described below one by one. The active containment spray system (Korean commercial reactor, SMART reactor, etc.) sprays a large amount of cooling water using containment spray pumps when an accident occurs, recollects the cooling water to an in-containment refueling water storage tank or sump, and re-sprays the cooling water to decrease the pressure and temperature of the containment and the concentration of radioactive material for a long time. The active containment spray system may perform a long-term spraying function and requires a power system to be available for activating the pumps. The passive containment spray system (Canadian CANDU, etc.) has a cooling water storage tank at an upper side of the containment and sprays a large amount of cooling water when an accident takes place to decrease the pressure and temperature inside the containment and the concentration of the radioactive material. Since the passive containment spray system has a limited storage capacity of cooling water, and thus, cannot be operated more than a predetermined time. Accordingly, the cooling water storage tank needs to be periodically made up using a pump for long-term use of the passive containment spray system. This means that the passive containment spray system also needs to use a pump and a power system for activating the pump in order for a long-term operation. The suppression tank (commercial BWR, CAREM: Argentina, IRIS: Westinghouse, U.S. et. al.) guides the steam discharged into the containment to the suppression tank using a difference in pressure between the containment and the inside of the suppression tank and condenses the steam to decrease the pressure and temperature in the containment and the concentration of the radioactive material. The suppression tank operates only when the pressure in the containment is higher than the pressure in the suppression tank. The passive containment cooling system has heat exchangers and a cooling water tank installed in or outside the containment and condenses the steam in the containment using the heat exchangers to decrease the pressure and temperature in the containment and the concentration of the radioactive material. The passive containment cooling system uses the natural circulation in the containment and thus has a lower performance in reducing the pressure and temperature and concentration of radioactive material as compared with the active containment spray system. Besides, there is a sort of passive containment cooling system (AP1000: Westinghouse, U.S.) that applies a steel containment vessel to cool (spray, air) the external wall and that condenses the steam in the containment vessel on the internal wall of the containment vessel to thus decrease the pressure and temperature in the containment vessel and the concentration of radioactive material. This system uses the natural circulation in the containment similarly to the passive containment cooling system and thus shows a relatively low performance in reducing pressure and temperature and the concentration of radioactive material as compared with the active containment spray system. Most of the above-described systems show a relatively excellent performance in decreasing the pressure and temperature inside the containment. However, among the radioactive materials that may spread to the external environment when an accident occurs in the nuclear power plant, iodine may have a highest proportion of concentration. Iodine, when contacts water, is mostly dissolved in the water (solubility 0.029 g/100 g(20° C.)). Accordingly, among the containment-related safety systems, the active containment spray system (which is adopted for the Korean commercial reactors), which uses an active pump to spray a great amount of cooling water and to recirculate the cooling water for a long time, may show the most excellent performance in decreasing the concentration of radioactive material in the containment. However, the active safety system necessarily requires supply of emergency AC power for operating the active components such as pumps when an accident occurs in the nuclear power plant, and without supply of emergency AC power, does not operate. In this point of view, demand for the passive safety system with relatively high safety is on the rise. This is why the passive safety system does not require a power system nor continuous operation of the active components. However, in case the passive safety system is adopted as safety system of the containment, the concentration of radioactive material in the containment would be relatively higher due to a lower performance in containment cooling as compared with the active safety system. Further, an exclusion area boundary (EAB) is set for the public safety to restrict the public access in preparation for an accident that may occur in the nuclear power plant. In case the passive safety system is applied to the nuclear power plant, the safety of nuclear power plant may be increased relatively further than the active safety system is applied, but it needs to secure a relatively broader EAB. The expansion of EAB may result in a significantly increased cost of constructing the nuclear power plant. Accordingly, an increasing need exists for a facility for reducing radioactive materials, which allows for application of a passive safety system to enhance the safety of nuclear power plant by resolving the problem of an expanding EAB. Therefore, an aspect of the detailed description is to provide a facility for reducing radioactive material in a containment, which may contribute to increasing safety of a nuclear power plant. In particular, an aspect of the detailed description proposes a facility for reducing radioactive material, which may reduce the concentration of radioactive material that is discharged in the containment when an accident occurs in the nuclear power plant. Another aspect of the detailed description is to provide a facility for reducing radioactive material which is configured to suppress an increase in the number of valves that may occur due to an introduction thereof and to prevent re-volatilization of radioactive material and a nuclear power plant having the same. Still another aspect of the detailed description is to provide a facility for reducing radioactive material, which may resolve the problem of an increasing EAB that may be caused as a passive safety system is adopted in a nuclear power plant and a nuclear power plant having the same. To achieve these and other advantages and in accordance with the purpose of this specification, as embodied and broadly described herein, there is provided a facility for reducing radioactive material. The facility comprises a cooling water storage unit installed inside a containment and formed to store cooling water; a boundary unit forming a boundary of radioactive material inside the containment and surrounding a reactor coolant system installed inside the containment to prevent a radioactive material from releasing from the reactor coolant system or a pipe connected with the reactor coolant system to the containment; a connecting pipe connected with an inner space of the boundary unit and the cooling water storage unit to guide a flow of a fluid caused by a pressure difference between the boundary unit and the cooling water storage unit from the boundary unit to the cooling water storage unit; and a sparging unit disposed to be submerged in the cooling water stored in the cooling water storage unit and connected with the connecting pipe to sparge the fluid that has passed through the connecting pipe and the radioactive material contained in the fluid to the cooling water storage unit. According to an embodiment of the present invention, the cooling water storage unit may include an inlet through which the connecting pipe passes, and the highest part of the connecting pipe may be formed at a predetermined height from a bottom of the cooling water storage unit to prevent the cooling water stored in the cooling water storage unit from flowing back to an inside of the boundary unit. According to another embodiment of the present invention, the facility may further comprise a check valve formed to allow for a flow only in one direction and installed at the connecting pipe to prevent the cooling water in the cooling water storage unit from flowing back to the boundary unit through the connecting pipe. To achieve these and other advantages and in accordance with the purpose of this specification, as embodied and broadly described herein, there is provided a facility for reducing radioactive material. The facility comprises a boundary unit forming a boundary of a radioactive material inside a containment and surrounding a reactor coolant system installed inside the containment to prevent the radioactive material from releasing from the reactor coolant system or a pipe connected with the reactor coolant system to the containment; a discharging unit installed at the boundary of the radioactive material to form a fluid path that runs from the boundary unit to the containment and configured to guide a flow of a fluid caused by a pressure difference between the containment and the boundary unit from the containment to the boundary unit through the fluid path; and a filter facility installed in the fluid path of the discharging unit to capture the radioactive material contained in the fluid passing through the discharging unit in the boundary unit. According to an embodiment of the present invention, at least a portion of the boundary unit may be expanded to a region adjacent to the containment while surrounding a penetration pipe penetrating the containment to prevent a loss-of-coolant accident from occurring due to breakage of the penetration pipe in a region between the containment and the boundary unit. According to another embodiment of the present invention, the boundary unit may form a sealing structure around the reactor coolant system to prevent release of the radioactive material. According to another embodiment of the present invention, at least a portion of the boundary unit may be formed by a concrete structure inside the containment or a coating member installed on the concrete structure. According to another embodiment of the present invention, the boundary unit may comprise a barrier formed to surround the reactor coolant system; and a cover formed to cover an upper part of the reactor coolant system and coupled with the barrier. According to another embodiment of the present invention, the filter facility may comprise at least one of: a filter configured to form iodic silver by reacting silver nitrate with iodine contained in the fluid and formed to remove the iodic silver from the fluid; and an absorbent configured to remove the iodine contained in the fluid through chemisorption that is performed by charcoal. According to another embodiment of the present invention, the facility may further comprise a cooling water storage unit installed inside the containment, the cooling water storage unit formed to store cooling water for dissolving the radioactive material. According to another embodiment of the present invention, the discharging unit may be extended from the boundary unit to an inside of the cooling water storage unit to discharge the fluid into the cooling water storage unit. According to another embodiment of the present invention, the facility may further comprise a cooling water recollecting portion forming a fluid path that runs from the containment to the cooling water storage unit to recollect cooling water present inside the containment to the cooling water storage unit; and an opening portion formed by opening at least a portion of the cooling water storage unit to maintain pressure balance between the cooling water storage unit and an inside of the containment. According to another embodiment of the present invention, the facility may further comprise an additive injection unit supplying an additive for maintaining a pH of cooling water to a predetermined value or more to prevent volatilization of the radioactive material dissolved in the cooling water storage unit. According to another embodiment of the present invention, the additive injection unit may be installed at a predetermined height inside the cooling water storage unit to be submerged in the cooling water as a water level of the cooling water storage unit increases, and as the additive injection unit is submerged in the cooling water, the additive may be dissolved in the cooling water. According to another embodiment of the present invention, the additive injection unit may be installed on a fluid path of the cooling water recollecting portion to dissolve the additive in the cooling water recollected to the cooling water recollecting portion. According to another embodiment of the present invention, the facility may further comprise a sparging unit installed at an end of the discharging unit to be submerged in the cooling water of the cooling water storage unit and configured to sparge a fluid that has passed through the discharging unit, to condense steam and to dissolve soluble radioactive materials in the discharged air contained in the fluid. According to another embodiment of the present invention, the sparging unit may have a flow resistance therein to induce an even distribution of the fluid into a plurality of fine fluid paths. According to another embodiment of the present invention, the facility may further comprise a pressure balance line passing through at least a portion of the boundary unit and extended to an inside of the containment to form a fluid path of atmosphere passing through the boundary of the radioactive material, and the pressure balance line, when a pressure inside the containment is higher than a pressure inside the boundary unit, introduces atmosphere inside the containment to the inside of the boundary unit to prevent the cooling water in the cooling water storage unit from flowing back to the inside of the boundary unit. According to another embodiment of the present invention, the facility may further comprise a check valve formed to allow for a flow only in one direction and installed at the pressure balance line to prevent the atmosphere inside the boundary unit from being discharged to the inside of the containment through the pressure balance line. According to another embodiment of the present invention, the cooling water storage unit may be connected with a pipe forming a fluid path that runs to a safety injection line of a safety injection system to inject the cooling water stored in the cooling water storage unit to the inside of the reactor coolant system. Further scope of applicability of the present application will become more apparent from the detailed description given hereinafter. However, it should be understood that the detailed description and specific examples, while indicating preferred embodiments of the disclosure, are given by way of illustration only, since various changes and modifications within the spirit and scope of the disclosure will become apparent to those skilled in the art from the detailed description. Description will now be given in detail of the exemplary embodiments, with reference to the accompanying drawings. For the sake of brief description with reference to the drawings, the same or equivalent components will be provided with the same reference numbers, and description thereof will not be repeated. In describing the present invention, if a detailed explanation for a related known function or construction is considered to unnecessarily divert the gist of the present disclosure, such explanation has been omitted but would be understood by those skilled in the art. The accompanying drawings are used to help easily understood the technical idea of the present invention and it should be understood that the idea of the present disclosure is not limited by the accompanying drawings. It will be understood that although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are generally only used to distinguish one element from another. It will be understood that when an element is referred to as being “connected with” another element, the element can be connected with the other element or intervening elements may also be present. In contrast, when an element is referred to as being “directly connected with” another element, there are no intervening elements present. A singular representation may include a plural representation unless it represents a definitely different meaning from the context. Terms such as “include” or “has” are used herein and should be understood that they are intended to indicate an existence of several components, functions or steps, disclosed in the specification, and it is also understood that greater or fewer components, functions, or steps may likewise be utilized. FIG. 1A is a concept view illustrating a facility 1100 for reducing radioactive material and a nuclear power plant 110 having the same according to an embodiment of the present invention. The nuclear power plant 110 includes a containment 112, a reactor coolant system 111, and a core 111a. In addition to the components shown in FIG. 1A, the nuclear power plant 110 may include a reactor coolant pump, a pressurizer, a steam generator, other systems for normal operation of the nuclear power plant 110 and various systems for securing safety of the nuclear power plant 110. The containment 112 is installed outside the reactor coolant system 111 to prevent release of radioactive material. The containment 112 serves as a final barrier to prevent the radioactive material from releasing from the nuclear power plant 110 to the external environment. Containments 112 may be classified into a containment building (or also referred to as a reactor building) formed of reinforced concrete, a containment vessel formed of steel, and a safeguard vessel formed of steel depending on the material constituting the pressure boundary. The containment vessel is a large vessel designed to have a low pressure like the containment building, and the safeguard vessel is a small vessel designed to have a small size and having the higher design pressure. Unless mentioned otherwise, as used herein, the term “containment 112” includes all of the containment building, the reactor building, the containment vessel, and the safeguard vessel. The reactor coolant system 111 is installed in the containment 112. The reactor coolant system 111 is a coolant system that delivers and conveys heat energy generated by nuclear fission of fuel in the core 111a. A primary fluid fills the inside of the reactor coolant system 111. When an accident, such as a loss of coolant accident, occurs, steam may be discharged from the reactor coolant system 111 to an atmosphere of the containment 112, and an isolation system of the containment 112 shuts off the external release of the atmosphere and the radioactive material contained in the atmosphere. A reactor coolant pump (not shown) induces the circulation of the primary fluid, and a pressurizer (not shown) maintains a pressurized state that exceeds a saturated pressure in order to control the pressure of the coolant at normal plant operation. The facility 1100 for reducing radioactive material is installed inside the containment 112. The facility 1100 for reducing radioactive material is configured to sparge into the cooling water, i) steam discharged from the reactor coolant system 111 installed in the containment 112 or a pipe 113 connected with the reactor coolant system 111 when an accident occurs, ii) an atmosphere in a boundary unit 1120, and iii) radioactive material contained in the steam and the air. The facility 1100 for reducing radioactive material includes a cooling water storage unit 1110, a boundary unit 1120, a connecting pipe 1130, and a sparging unit 1140. The cooling water storage unit 1110 is installed in the containment 112. The cooling water storage unit 1110 is formed to store cooling water dissolving the radioactive material therein. The cooling water storage unit 1110 may be configured as a tank or pool. The cooling water storage unit 1110 may be shared by the cooling water storage unit 1110 and other systems of the nuclear power plant 110. For example, the facility 1100 for reducing radioactive material and a passive safety injection system (not shown) and a passive residual heat removal system (not shown) share the cooling water storage unit 1110. The cooling water storage unit 1110 may be installed at an upper side or lower side of the containment 112. The cooling water storage unit 1110 may be installed at an upper side of the containment 112 to receive cooling water that is condensed and falls in the containment 112 as shown in FIG. 1A. In case a containment spray system (not shown) is installed in the nuclear power plant 110, the cooling water storage unit 1110 may be installed at an upper side or lower side of the containment 112 to receive the sprayed cooling water. A cooling water recollecting portion 1110a and an opening portion 1110b may be installed in the cooling water storage unit 1110. The cooling water recollecting portion 1110a forms a fluid path that runs from the containment 112 to the cooling water storage unit 1110 to recollect the condensed water generated in the containment 112 to the cooling water storage unit 1110. The opening portion 1110b is formed as at least a portion of the cooling water storage unit 1110 is opened to maintain the pressure balance between the cooling water storage unit 1110 and the containment 112. The cooling water recollecting portion 1110a and the opening portion 1110b may share the same fluid path. The cooling water storage unit 1110 has an inlet 1111 through which the connecting pipe 1130 passes. The highest part of the connecting pipe 1130 may be formed at a predetermined height from the bottom of the cooling water storage unit 1110 to prevent backflow of the cooling water retained in the cooling water storage unit 1110. The boundary unit 1120 is installed between the reactor coolant system 111 and the containment 112 to form a radioactive material boundary. The boundary unit 1120 surrounds the reactor coolant system 111 to prevent radioactive material from releasing from the reactor coolant system 111 or pipe 113 connected with the reactor coolant system 111 to the containment 112. The boundary unit 1120 forms a sealing structure around the reactor coolant system 111 to prevent the radioactive material from releasing along a path other than the connecting pipe 1130. The boundary unit 1120 is designed to have a design pressure that may withstand the pressure of a head difference or more between the cooling water storage unit 1110 and the sparging unit 1140. At least a portion of the boundary unit 1120 may be formed by a concrete structure inside the containment 112 and a coating member (1123) such as a steel liner et. al. installed on the concrete structure. The boundary unit 1120 may include a barrier 1121 formed to surround the periphery of the reactor coolant system 111 and a cover 1122 formed to cover an upper part of the reactor coolant system 111 and may form a sealing structure around the reactor coolant system 111 by i) the bottom surface or dual bottom surface of the containment 112, ii) the barrier 1121, and iii) the cover 1122. The connecting pipe 1130 is connected with an inner space of the boundary unit 1120 and the cooling water storage unit 1110 to guide a flow of the fluid generated by a pressure difference between the boundary unit 1120 and the cooling water storage unit 1110 from the boundary unit to the cooling water storage unit 1110. The connecting pipe 1130 forms a fluid path that runs from an inner space of the boundary unit 1120 to the cooling water storage unit 1110. If the pressure in the boundary unit 1120 is larger than the pressure in the cooling water storage unit 1110, the fluid in the boundary unit 1120 flows through the fluid path of the connecting pipe 1130 to the cooling water storage unit 1110. The cooling water storage unit 1110 has an inlet 1111 that allows the connecting pipe 1130 to pass therethrough. The connecting pipe 1130 extends through the inlet 1111 of the cooling water storage unit 1110 to the inside of the cooling water storage unit 1110 to form a fluid path that runs to the sparging unit 1140 and is connected with the sparging unit 1140. The atmosphere (steam and air) and radioactive material in the boundary unit 1120 are delivered to the sparging unit 1140 through the connecting pipe 1130. The sparging unit 1140 is disposed to be submerged in the cooling water contained in the cooling water storage unit 1110 and is connected with the connecting pipe 1130 to sparge the fluid that has passed through the connecting pipe 1130 and the radioactive material contained in the fluid to the cooling water storage unit 1110. The sparging unit 1140 may have a plurality of sparging holes 1141 formed to sparge the fluid and radioactive material finely. The sparging unit 1140 may have a plurality of fine fluid paths (not shown) that run the plurality of sparging holes 1141. The sparging unit 1140 may have a flow resistance therein, to allow the fluid to be evenly distributed through the plurality of fine fluid paths. The nuclear power plant 110 may include pipes 113 for connecting the systems operated as the nuclear power plant 110 is in normal operation, other than the facility 1100 for reducing radioactive material, to the reactor coolant system 111. The pipe 113 may pass through the containment 112 and the boundary unit 1120 of the facility 1100 for reducing radioactive material. The pipe 113 may have a plurality of isolation valves 113a, 113b, 113c, 113e, 113f, 113g, and 113h or a check valve 13d arranged to be spaced apart from each other to close both sides of a broken line when a break occurs. The facility 1100 for reducing radioactive material, contrary to when double containments 112 are installed, does not form a high-pressure boundary with the containment 112, thus minimizing an increase in the economical expense due to added facilities. The facility 1100 for reducing radioactive material is a low-pressure facility. Hereinafter, the operations of the facility 1100 for reducing radioactive material when the nuclear power plant 110 is in normal operation and when an accident occurs are described with reference to FIGS. 1B to 1F. FIG. 1B is a concept view illustrating a normal operation state of the nuclear power plant 110 shown in FIG. 1A. When the nuclear power plant 110 is in normal operation, the isolation valves 113a, 113b, 113c, 113e, 113f, 113g, and 113h installed on the pipe 113 connecting the systems (not shown) for normal operation of the nuclear power plant 110 with the reactor coolant system 111 may remain opened. The fluids circulating for normal operation of the nuclear power plant 110 may flow through the pipe 113. The facility 1100 for reducing radioactive material is a facility passively operated by a pressure difference formed between the boundary unit 1120 and the cooling water storage unit 1110, and since there is little pressure difference between the boundary unit 1120 and the cooling water storage unit 1110 when the nuclear power plant 110 is in normal operation, the facility 1100 for reducing radioactive material remains in the standby state. Hereinafter, the operations of the facility 1100 for reducing radioactive material i) when pipe breakage occurs in the facility 1100 for reducing radioactive material and ii) when pipe breakage occurs between the facility 1100 for reducing radioactive material and the containment 112 are described separately from each other. FIG. 1C is a concept view illustrating the operation of the facility 1100 for reducing radioactive material when an accident occurs in the nuclear power plant 110 shown in FIG. 1A. When an accident such as pipe breakage occurs in the facility 1100 for reducing radioactive material, the reactor coolant and radioactive material may be discharged through the broken line 113i to the inside of the boundary unit 1120. When the accident occurs, the isolation valves 113a, 113b, 113c, 113e, 113f, 113g, and 113h installed on the pipe 113 passing through the boundary unit 1120 are closed by a related signal. In case a check valve 113d forming a fluid path is installed towards the reactor coolant system 111, the flow in an opposite direction is shut off, and the facility 1100 for reducing radioactive material maintains the sealing structure. The isolation valves 113a, 113b, 113c, 113e, 113f, 113g, and 113h may share the operation signal, and thus, the facility 1100 for reducing radioactive material may be operated even without a separate signal. The nuclear power plant 110 may include a passive residual heat removing system 114 to remove sensible heat in the reactor coolant system 111 and residual heat of the core 111a and a passive safety injection system 15 to inject cooling water to the inside of the reactor coolant system 111 to maintain the water level of the reactor coolant system 111. The passive residual heat removing system 114 and the passive safety injection system 115 start their operation to secure safety of the nuclear power plant 110 when an accident occurs like the facility 1100 for reducing radioactive material. If steam is discharged from the broken line 113i, the radioactive material, together with the steam, is discharged into the boundary unit 1120, and the pressure inside the boundary unit 1120 gradually increases. As the pressure inside the boundary unit 1120 increases, a pressure difference of H1 or more is generated between the boundary unit 1120 and the cooling water storage unit 1110, and the fluid is rendered to flow by the pressure difference from the boundary unit 1120, which has a relatively high pressure, to the cooling water storage unit 1110, which has a relatively low pressure. The connecting pipe 1130 guides the flow caused by the pressure difference to the inside of the cooling water storage unit 1110, and the fluid (steam, air, and radioactive material) that has passed through the cooling water storage unit 1110 is sparged into the cooling water through the sparging unit 1140 submerged in the cooling water. Accordingly, the steam is sparged into the cooling water and is condensed, and the air is cooled to go up. The soluble radioactive material is dissolved in the cooling water and is collected in the cooling water storage unit 1110. As the steam is condensed and the radioactive material is collected into the cooling water, a limited amount of the radioactive material discharged from the broken line 113i is discharged to the containment 112. A small amount of radioactive material discharged to the inside of the containment 112 is suppressed from releasing to the external environment by the containment 112. In particular, iodine that may be spread to the external environment with the highest concentration among radioactive materials is soluble and is mostly dissolved in the cooling water. The facility 1100 for reducing radioactive material stays in operation if the amount of cooling water in the cooling water storage unit 1110 is maintained to be a predetermined value or more and the pressure difference between the facility 1100 for reducing radioactive material and the inside of the containment 112 is not less than H1. The cooling water storage unit 1110 receives the condensed water introduced through the cooling water recollecting portion 1110a. Accordingly, the facility 1100 for reducing radioactive material may maintain the cooling water level that is required for its operation. The sparging unit 1140 sparges the steam that may cause the pressure inside the containment 112 to increase into the cooling water storage unit 1110 and condenses the steam. Accordingly, the cooling water storage unit 1110 may suppress an increase in the pressure inside the containment 112 and may decrease the design pressure of the containment 112. FIG. 1D is a concept view illustrating a state in which a nuclear power plant 110 operates when an accident occurs at a location different from the position shown in FIG. 1C. The pipe passing through the boundary unit 1120 may be broken in the boundary unit 1120 as described above in connection with FIG. 1C, but may be broken in a space between the containment 112 and the boundary unit 1120 as described above in connection with FIG. 1D. If pipe breakage occurs in the space between the containment 112 and the boundary unit 1120, steam is discharged from the broken line 113i to result in the pressure in the containment 112 increasing. However, since the steam stops being discharged when all of the isolation valves 113a, 113b, 113c, 113e, 113f, 113g, and 113h and the check valve 113d are closed by a related signal, the pressure inside the containment 112 does not steadily go up. Accordingly, when an accident occurs to cause the pressure outside the facility 1100 for reducing radioactive material to increase, the accident is suppressed early. As the pressure inside the containment 112 increases, the cooling water in the cooling water storage unit 1110 may be pressurized to cause the water level of the connecting pipe 1130 to partially go up. However, the highest part of the connecting pipe 1130 of the cooling water storage unit 1110 is formed at a predetermined height from the bottom of the cooling water storage unit 1110 so as to prevent backflow of the cooling water despite the increased pressure in the containment 112. Accordingly, in case there is no significant pressure difference between the inside of the containment 112 and the boundary unit 1120 (<H2), the cooling water retained in the cooling water storage unit 1110 does not flow back to the inside of the boundary unit 1120. FIGS. 1E and 1F are concept views illustrating a nuclear power plant 110 including passive safety systems other than a facility 1100 for reducing radioactive material. The nuclear power plant 110 includes a passive containment cooling system 116 that reduces pressure inside the containment 112 through cooling. The passive containment cooling system 116 has a heat exchanger (not shown). The cooling fluid passing through the heat exchanger exchanges heat with the fluid inside the containment 112. Accordingly, heat is delivered from the inside of the containment 112 to the cooling fluid, and the cooling fluid is discharged to the outside along the fluid path connected with the heat exchanger. Such process is repeated to suppress the increasing pressure inside the containment 112. Referring to FIG. 1E, the heat exchanger (not shown) provided in the passive containment cooling system 116 may be installed inside the cooling water storage unit 1110. If the cooling fluid passing through the heat exchanger exchanges heat with the atmosphere or cooling water and/or atmosphere inside the cooling water storage unit 1110, the cooling water storage unit 1110 is cooled. Since the cooling water storage unit 1110 and the inside of the containment 112 are formed to communicate with each other through the cooling water recollecting portion 1110a or opening portion 1110b, cooling the cooling water storage unit 1110 leads to the containment 112 being cooled, and the increasing pressure inside the containment 112 may be suppressed. Referring to FIG. 1F, the heat exchanger (not shown) provided in the passive containment cooling system 116 may be installed at an upper side of the containment 112. The cooling fluid flowing through the inner fluid path of the heat exchanger exchanges heat with the atmosphere inside the containment 112. If the facility 1100 for reducing radioactive material and the passive containment cooling system 116 are both adopted in the nuclear power plant 110, the steam discharged to the inside of the containment 112 is cooled and condensed by the passive containment cooling system 116. Since the condensed water formed as the steam is condensed may be recollected to the cooling water storage unit 1110, the water in the cooling water storage unit 1110 may be maintained at a proper level or more. Another embodiment of the present invention is now described. FIG. 2 is a concept view illustrating a facility 1200 for reducing radioactive material and a nuclear power plant 120 having the same according to another embodiment of the present invention. The facility 1200 for reducing radioactive material includes a cooling water storage unit 1110, a boundary unit 1120, a sparging unit 1140, and a pressure balance line 1250. The description of the cooling water storage unit 1210, the boundary unit 1220, the connecting pipe 1230, and the sparging unit 1240 is not repeated and replaced with the above description thereof. The boundary unit 1220 forms a boundary of radioactive material. The pressure balance line 1250 passes through at least a portion of the boundary unit 1220 to form a flow path of atmosphere passing through the boundary of radioactive material and extends to the inside of the containment 122. The pressure balance line 1250, in case the pressure inside the containment 122 is higher than the pressure inside the boundary unit 1220, introduces the atmosphere inside the containment 122 to the inside of the boundary unit 1220. By doing so, the pressure balance line 1250 prevents backflow of the cooling water in the cooling water storage unit 1210 to the inside of the boundary unit 1220. The inflow of atmosphere through the pressure balance line 1250 is passively made by the pressure difference between the containment 122 and the boundary unit 1220. The pressure balance line 1250 may be split from the connecting pipe 1230 as shown in FIG. 2. The pressure balance line 1250 may pass through an upper side of the cooling water storage unit 1210 and may extend to the inside of the containment 122. The atmosphere inside the containment 122 is introduced to the inside of the boundary unit 1220 through the pressure balance line 1250. Since the pressure balance line 1250 suppresses an increase in the pressure difference in an opposite direction of the boundary unit 1220, the mechanical integrity of the boundary unit 1220 may be more safely maintained. The check valve 1251 is installed in the pressure balance line 1250. The check valve 1251 is formed to allow for a flow only in one direction. The check valve 1251 prevents the atmosphere inside the boundary unit 1220 from being discharged to the inside of the containment 122 through the pressure balance line 1250. According to the conditions of accident, the pressure inside the boundary unit 1220 may be higher than the pressure inside the containment 122. In such case, the atmosphere inside the boundary unit 1220 may be discharged to the inside of the containment 122 through the pressure balance line 1250 to lose the unique functions of the facility 1200 for reducing radioactive material. The check valve 1251 cuts off the flow to the pressure balance line 1250 to thus prevent the atmosphere inside the boundary unit 1220 from being discharged to the inside of the containment 122. FIG. 3 is a concept view illustrating a facility 1100 for reducing radioactive material and a nuclear power plant 130 having the same according to still another embodiment of the present invention. The facility 1300 for reducing radioactive material includes a cooling water storage unit 1310, a boundary unit 1320, a connecting pipe 1330, and a sparging unit 1340. The description of the similar components is not repeated and replaced with the above description thereof. The cooling water storage unit 1310 of the facility 1300 for reducing radioactive material may be installed at a lower region in the inner space of the containment 132. The cooling water storage unit 1310 has a cooling water recollecting portion 1310a and an opening portion 1310b. A space is formed between the outer wall of the containment 132 and the inner structure of the containment 132. The fluid inside the containment 132 may flow to the cooling water storage unit 1310 through the space between the outer wall and the structure, the cooling water recollecting portion 1310a, and the opening portion 1310b. Likewise, the fluid inside the cooling water storage unit 1310 may flow to the inside of the containment 132 through the cooling water recollecting portion 1310a, the opening portion 1310b, and the space between the outer wall and the structure. Comparison between the facility 1100 for reducing radioactive material shown in FIG. 1A and the facility 1300 for reducing radioactive material shown in FIG. 3 shows that the positions where the cooling water storage units 1110 and 1310 are installed may vary depending on the requirements for the internal design of the containments 112 and 132. Even when the positions where the cooling water storage units 1110 and 1310 are installed differently, the facilities 1100 and 1300 for reducing radioactive material may be configured to not cause a deterioration of their functions. FIG. 4 is a concept view illustrating a facility 1400 for reducing radioactive material and a nuclear power plant 140 having the same according to still another embodiment of the present invention. The facility 1400 for reducing radioactive material includes a cooling water storage unit 1410, a boundary unit 1420, a connecting pipe 1430, a sparging unit 1440, and a pressure balance line 1450. Unlike the facility 1200 for reducing radioactive material shown in FIG. 2, the pressure balance line 1450 is not split from the connecting pipe 1430 but is formed independently. The pressure balance line 1450 passes through a boundary of the radioactive material, which is formed by the boundary unit 1420, and extends up to the inside of the containment 142. The pressure balance line 1450, in case the pressure inside the containment 142 is higher than the pressure of the boundary unit 1420, introduces atmosphere to decrease the pressure inside the containment 142. The pressure balance line 1450 prevents the cooling water in the cooling water storage unit 1410 from flowing back to the inside of the boundary unit 1420. The pressure balance line 1450 has a check valve 1451. The check valve 1451 is formed to allow for a flow only in a direction. The check valve 1451 prevents atmosphere from being discharged from the inside of the boundary unit 1420 to the inside of the containment 142. The mechanical integrity of the boundary unit 1420 may be more safely maintained by the pressure balance line 1450 and the check valve 1451. FIG. 5 is a concept view illustrating a facility 1500 for reducing radioactive material and a nuclear power plant 150 having the same according to yet still another embodiment of the present invention. The facility 1500 for reducing radioactive material includes a cooling water storage unit 1510, a boundary unit 1520, a connecting pipe 1530, and a sparging unit 1540. The connecting pipe 1530 has a check valve 1531 to prevent the cooling water inside the cooling water storage unit 1510 from flowing back to the boundary unit 1520 through the connecting pipe 1530. The check valve 1531 allows for only flow that is formed from the boundary unit 1520 to the cooling water storage unit 1510 and cuts off flow in an opposite direction. Even when the pressure inside the cooling water storage unit 1510 is higher than the pressure inside the boundary unit 1520 due to an accident, the check valve 1531 may prevent the cooling water retained in the cooling water storage unit 1510 from flowing back to the boundary unit 1520. FIG. 6 is a concept view illustrating a facility 1600 for reducing radioactive material and a nuclear power plant 160 having the same according to yet still another embodiment of the present invention. The facility 1600 for reducing radioactive material is installed inside the containment 162, and when an accident occurs, is configured to sparge, to the cooling water storage unit 1610, the radioactive material discharged from a reactor coolant system 161 or pipes 163, 163′, and 165c connected with the reactor coolant system 161. The cooling water storage unit 1610 is installed inside the containment 162. The cooling water storage unit 1610 may be formed as a tank or pool to store cooling water therein. Further, as the cooling water storage unit 1610, an in-containment refueling water storage tank may be used as well. When an accident occurs, the atmosphere (steam and air) inside the boundary unit 1620 is sparged into the cooling water as the facility 1600 for reducing radioactive material operates. The cooling water storage unit 1610 may be shared by other safety systems of the nuclear power plant 160 than the facility 1600 for reducing radioactive material. For example, the facility 1600 for reducing radioactive material and a safety injection system 165 may share the cooling water storage unit 1610. As another example, the facility 1600 for reducing radioactive material and a passive residua heat removal system (not shown) may share the cooling water storage unit 1610. The cooling water storage unit 1610 may be installed at an upper side or lower side of an inner space of the containment 162. Condensed water may be formed inside the containment 162 and may fall. The cooling water storage unit 1610 may be installed at an upper side of the inner space of the containment 162 to collect the falling condensed water. In case the nuclear power plant 160 has a containment spray system (not shown), the cooling water storage unit 1610 may be installed at an upper or lower side of the containment 162 to receive the sprayed cooling water. The cooling water storage unit 1610 has an inlet 1611 through which a connecting pipe 1630 to be described below passes. The highest part of the connecting pipe 1630 may be formed at a predetermined height from the bottom of the cooling water storage unit 1610 to prevent backflow of the cooling water retained in the cooling water storage unit 1610. The boundary unit 1620 is installed between the reactor coolant system 161 and the containment 162 to form a boundary of the radioactive material. The boundary unit 1620 surrounds the reactor coolant system 161 to prevent release of the radioactive material from the pipes 163, 163′, and 165c connected with the reactor coolant system 161 to the containment 162. The boundary unit 1620 forms a sealing structure around the reactor coolant system 161 to prevent release of the radioactive material along a path other than the connecting pipe 1630 to be described below. The pipes 163, 163′, and 165c passing through the boundary unit 1620 has isolation valves 163a, 163b, 163a′, and 165c′ and check valves 163b′, 163b″, 165c″. The isolation valves 163a, 163b, 163a′, and 165c′ and the check valves 163b′, 163b″, 165c″ are closed when an accident occurs to maintain the sealing structure. The boundary unit 1620 is formed to have a design pressure to withstand a pressure of a head difference or more between the cooling water storage unit 1610 and the sparging unit 1640. At least a portion of the boundary unit 1620 may be formed by a concrete structure inside the containment 162 and a coating member (1623) such as steel liner et. al. installed on the concrete structure. The boundary unit 1620 may include a barrier 1621 and a cover 1622. The barrier 1621 is formed to surround the periphery of the reactor coolant system 161. The cover 1622 is formed to cover an upper portion of the reactor coolant system 161. The barrier 1621, the cover 1622, and the bottom surface (or dual bottom surface) of the containment 162 may form a sealing structure around the reactor coolant system 161. The nuclear power plant 160 includes penetration pipes 163 and 163′ passing through the containment 162. The penetration pipes 163 and 163′ are connected with the reactor coolant system 161 or a secondary system. The penetration pipes 163 and 163′ may include a plurality of isolation valves 163a, 163b, and 163a′ or check valve 163b′ arranged to be spaced apart from each other to close both sides of the broken line when breakage occurs. In case the boundary unit 1620 and the containment 162 are spaced apart from each other and the penetration pipes 163 and 163′ pass through a region between the boundary unit 1620 and the containment 162, a loss-of-coolant accident may occur in the region between the boundary unit 1620 and the containment 162. In case a loss-of-coolant accident occurs in the region between the boundary unit 1620 and the containment 162, the radioactive material might not be trapped in the inside of the facility 1600 for reducing radioactive material. Accordingly, in case a loss-of-coolant accident occurs in the region between the boundary unit 1620 and the containment 162, the penetration pipes 163 and 163′ should have isolation valves to prevent additional release of the radioactive material. However, since the isolation valves have the mechanism to be opened or closed by a related safety system signal, the isolation valves may abnormally operate or might not operate. Addition of isolation valves is not preferred in view of simplifying the facility. To resolve such issue, the present invention has a structure that may prevent release of radioactive material even without installation of additional isolation valves. Specifically, at least a portion of the boundary unit 1620 is expanded up to a region adjacent to the containment 162 while wrapping around the penetration pipes 163 and 163′ in order to prevent the loss-of-coolant accident that may occur due to breakage of the penetration pipes 163 and 163′ between the boundary unit 1620 and the containment 162. As a result, the penetration pipes 163 and 163′ passing through the containment 162 up to the inside of the containment 162 is caused to be positioned inside the boundary unit 1620. Accordingly, the present invention may significantly lower the possibility that a loss-of-coolant accident occurs, e.g., by breakage of the pipes 163 and 163′ in the region between the boundary unit 1620 and the containment 162 and may prevent release of radioactive material even without installation of additional isolation valves. The connecting pipe 1630 is connected with the boundary unit 1620 and the cooling water storage unit 1610 to guide the fluid flow caused in the boundary unit 1620 to the cooling water storage unit 1610. The atmosphere inside the boundary unit 1620 includes steam or air, and when a loss-of-coolant accident occurs, may be caused to contain radioactive material. If an accident that causes the pressure inside the boundary unit 1620 to rise occurs and thus the difference between the pressure inside the boundary unit 1620 and the pressure inside the containment 162 is increased to H1 or more, the atmosphere inside the boundary unit 1620 is passively caused to flow to the cooling water storage unit 1610 through the connecting pipe 1630. The connecting pipe 1630 passes through the inlet 1611 of the cooling water storage unit 1610 up to the inside of the cooling water storage unit 1610 to deliver the atmosphere inside the boundary unit 1620 and radioactive material contained in the atmosphere to the sparging unit 1640. The sparging unit 1640 is connected with the connecting pipe 1630 to receive the atmosphere inside the boundary unit 1620 and the radioactive material contained in the atmosphere from the connecting pipe 1630. At least a portion of the sparging unit 1640 is submerged in the cooling water of the cooling water storage unit 1610 so that the sparging unit 1640 sparges the atmosphere and the radioactive material contained in the atmosphere to the cooling water. The sparging unit 1640 has a plurality of sparging holes 1641 formed to sparge the atmosphere inside the boundary unit 1620 and the radioactive material contained in the atmosphere finely. Further, the sparging unit 1640 has a plurality of fine fluid paths (not shown) that run to the plurality of sparging holes 1641. The sparging unit 1640 has a flow resistance in its inner fluid path to evenly distribute the fluid into the plurality of fine fluid paths. The steam sparged through the sparging unit 1640 to the cooling water storage unit 1610 is condensed, and the air is cooled to rise. The soluble radioactive material is mostly dissolved in the cooling water. In case the cooling water in the cooling water storage unit 1610 maintains a predetermined water level, and the pressure difference between the boundary unit 1620 and the containment 162 is H1 or more, the facility 1600 for reducing radioactive material remains in steady operation. In case a single connecting pipe 1630 and a single sparging unit 1640 are provided, the facility 1600 for reducing radioactive material may be caused to be impossible to operate as the connecting pipe 1630 or the sparging unit 1640 is blocked. Accordingly, it is preferable to provide a plurality of connecting pipes 1630 and a plurality of sparging units 1640 considering redundancy. The facility 1600 for reducing radioactive material may further include a cooling water recollecting portion 1610a and an opening portion 1610b. The opening portion 1610b prevents overpressure in the cooling water storage unit 1610. In contrast, the cooling water recollecting portion 1610a recollects the steam discharged from the cooling water storage unit 1610. The cooling water in the cooling water storage unit 1610 is evaporated as its temperature goes up, turning into steam. The steam may be discharged through the opening portion 1610b to the inside of the containment 162. The steam discharged to the inside of the containment 162 is cooled, turning into condensed water. The cooling water recollecting portion 1610a forms a fluid path at an upper part of the cooling water storage unit 1610 to recollect the condensed water to the cooling water storage unit 1610. The connection between the cooling water recollecting portion 1610a and the cooling water storage unit 1610 may be made by way of a pipe or structure. As shown in FIG. 6, the cooling water recollecting portion 1610a and the opening portion 1610b may be formed at an upper part of the cooling water storage unit 1610. More specifically, a portion of the upper structure forming the cooling water storage unit 1610 may form the cooling water recollecting portion 1610a and the opening portion 1610b. The cooling water recollecting portion 1610a and the opening portion 1610b are installed at separate regions from each other. However, the cooling water recollecting portion 1610a and the opening portion 1610b may be formed to share the same fluid path. The nuclear power plant 160 may have various safety systems other than the facility 1600 for reducing radioactive material. For example, as shown in FIG. 6, a passive safety injection system 165 may be installed in the nuclear power plant 160. The passive safety injection system 165 is a system form maintaining the water level of the reactor coolant system 161 by injecting a coolant to the inside of the reactor coolant system 161 when an accident, such as loss of coolant accident, occurs. The passive safety injection system 165 may include various types of tanks such as a core makeup tank 165a or safety injection tank 165b. The core makeup tank 165a or the safety injection tank 165b is connected with the reactor coolant system 161 by way of a safety injection line 165c and the pressure balance line 165d. The coolant is injected from the tanks 165a and 165b through the safety injection line 165c to the reactor coolant system 161. In case the facility 1600 for reducing radioactive material and the passive safety injection system 165 are both installed in the nuclear power plant 160, the passive safety injection system 165 may be installed in the inside of the boundary unit 1620 to prevent release of radioactive material. The facility 1600 for reducing radioactive material proposed herein, unlike when double containments 162 are installed, does not form a high-pressure boundary with the containment 162 and thus may minimize an increase in the economical expense that may occur due to added facilities. The facility 1600 for reducing radioactive material may minimize an increase in the number of isolation valves. FIG. 7A is a concept view illustrating a facility 1700 for reducing radioactive material and a nuclear power plant 170 having the same according to yet still another embodiment of the present invention. An opening portion 1710b is formed to protrude from an upper part of a cooling water storage unit 1710 to the inside of a containment 172. The opening portion 1710b forms a fluid path by way of a pipe or structure. A filter facility 1770 is disposed on the fluid path to capture the radioactive material that is about to exit the cooling water storage unit 1710. If the pressure in the cooling water storage unit 1710 goes up, the steam or air inside the cooling water storage unit 1710 is discharged through the opening portion 1710b. During the course, some of the radioactive materials dissolved in the cooling water storage unit 1710 are re-volatilized, and together with the steam or air, may be thus discharged through the opening portion 1710b to the containment 172. If the radioactive materials are discharged to the containment 172, the concentration of the radioactive material in the containment 172 may increase. The filter facility 1770 is disposed on the fluid path of the opening portion 1710b to capture the radioactive material that, together with the steam, is about to be discharged to the containment 172. The filter facility 1770 includes at least one of a filter and an absorbent. The filter and the absorbent are adapted to pass steam or air while capturing the radioactive material. As the filter, a high-efficiency particulate air (HEPA) filter may be adopted. The gaseous radioactive material contained in the steam or air is filtered out when passing through the filter. For example, in case the radioactive material is iodine, iodine is combined with silver nitrate while passing through the filter to thus turn into iodic silver, and is thus removed from the steam or air. As the absorbent, charcoal may be employed. Organic iodine compounds are combined with the materials impregnated in the charcoal to turn into quaternary ammonium salt and are absorbed into the charcoal. Molecular iodine is combined with the charcoal through chemisorption. The charcoal is typically utilized as an absorbent material thanks to its large internal surface area. Either or both of the filter and the absorbent may be disposed. However, the above-described filter and absorbent are offered merely as an example, and according to the present invention, the type of the filter and absorbent is not necessarily limited thereto. The cooling water recollecting portion 1710a, like the opening portion 1710b, has a fluid path formed by a pipe or structure. The fluid path of the cooling water recollecting portion 1710a may be formed to be submerged into the cooling water storage unit 1710. However, the cooling water storage unit 1710 and the cooling water recollecting portion 1710a, rather than separated from each other, are connected with each other. Hereinafter, the normal operation of the nuclear power plant 170 and the operation under accident of the nuclear power plant 170 are described with reference to FIGS. 7B and 7C. FIG. 7B is a concept view illustrating a normal operation state of the nuclear power plant 170 shown in FIG. 7A. The pipes 173 and 173′ connected with a system (not shown) relating to the normal operation of the nuclear power plant 170 have isolation valves 173a, 173b, and 173a′ or a check valve 173b′. When the nuclear power plant is in normal operation, the isolation valves necessary for the normal operation remain opened. When the nuclear power plant 170 is in normal operation, the water in the reactor coolant system 171 remains at a normal level. Accordingly, the passive safety injection system 175 remains in the standby state. The facility 1700 for reducing radioactive material is a facility that is passively operated by a pressure difference between a boundary unit 1720 and a containment 172. When the nuclear power plant 170 is in normal operation, little pressure difference is created between the boundary unit 1720 and the cooling water storage unit 1710, and thus, the facility 1700 for reducing radioactive material remains in the standby state. FIG. 7C is a concept view illustrating the operation under accident of the nuclear power plant 170 shown in FIG. 7A. If an accident such as a loss-of-coolant accident occurs in the nuclear power plant 170 due to, e.g., pipe breakage, steam and radioactive material are discharged through the broken line 173f. A number of safety systems installed in the nuclear power plant 170 start operations. When an accident occurs, the isolation valves 173a, 173b, and 173a′ relating to the normal operation of the nuclear power plant 170 are closed by a related signal. In case check valves 173b′ and 175c″ are installed to form a fluid path in a direction toward the reactor coolant system 171, the flow in the direction coming from the reactor coolant system 171 is shut off, and the boundary unit 1720 of the facility 1700 for reducing radioactive material maintains a sealing structure. The isolation valves 173a, 173b, 173a′, and 175c′ may share an operation signal. Accordingly, even when no separate signal is applied for the operation signal of the facility for reducing radioactive material, the operation of the isolation valves 173a, 173b, 173a′, 175c′ may allow the facility 1700 for reducing radioactive material to be operated. The nuclear power plant 170 may include a passive residual heat removing system 174 and a passive safety injection system 175. The passive residual heat removing system 174 removes sensible heat in the reactor coolant system 171 and residual heat in the core 171a. The passive safety injection system 175 injects a coolant into the reactor coolant system 171 to maintain the water level of the reactor coolant system 171. The passive safety injection system 175 is first described. The pipe connected with the core makeup tank 175a has an isolation valve 175a′ and a check valve 175a″. If the isolation valve 175a′ and the check valve 175a″ are opened, the coolant in the core makeup tank 175a is swiftly injected into the reactor coolant system 171. If the isolation valve 175d′ installed in the pressure balance line 175d is opened, steam is introduced from the high-pressure reactor coolant system 171 through the pressure balance line 175d to the safety injection tank 175b. As time goes by, the reactor coolant system 171 and the safety injection tank 175b form a pressure balance. If the reactor coolant system 171 and the safety injection tank 175b form the pressure balance, the coolant in the safety injection tank 175b is also injected into the reactor coolant system 171 by gravity water head. The coolant in the core makeup tank 175a and the safety injection tank 175b is injected through the safety injection line 175c to the reactor coolant system 171. Next, the passive residual heat removing system 174 is described. The passive residual heat removing system 174 may remove sensible heat from the reactor coolant system 171 and residual heat from the core 171a. A steam generator (not shown) is installed at the boundary between the primary system and the secondary system. The passive residual heat removing system is configured to circulate the coolant to the steam generator. As the coolant circulates, the sensible heat from the reactor coolant system 171 and the residual heat from the core 171a are removed to the outside. The nuclear power plant 170, as necessary, may further include other systems than the above-mentioned safety systems. If steam is discharged from the broken line 173f, the radioactive material, together with the steam, is discharged to the inside of the boundary unit 1720. As the steam and radioactive material are continuously discharged from the broken line 173f, the pressure inside the boundary unit 1720 is gradually increased. As the pressure inside the boundary unit 1720 is increased to H1 or more, a flow of the fluid (including steam, air, and radioactive material) is caused by a pressure difference from the boundary unit 1720 that has a relatively high pressure to the cooling water storage unit 1710 that has a relatively low pressure. The connecting pipe 1730 guides the flow of the fluid caused by the pressure difference to the cooling water storage unit 1710. The atmosphere that has passed through the connecting pipe 1730 is sparged into the cooling water through the sparging unit 1740 submerged in the cooling water storage unit 1710. Accordingly, the steam is sparged into the cooling water and is condensed. The air is cooled to rise. The soluble radioactive material is dissolved in the cooling water and is collected. Accordingly, the facility 1700 for reducing radioactive material may suppress the radioactive material from releasing from the containment 172 to the external environment. In particular, iodine, which has the highest concentration among the radioactive materials spread to the external environment, is soluble and thus is mostly dissolved in the cooling water. The facility 1700 for reducing radioactive material remains in steady operation when the amount of cooling water in the cooling water storage unit 1710 maintains a predetermined value or more and the pressure difference between the boundary unit 1720 and the containment 172 is H1 or more. The sparging unit 1740 sparges the steam that may lead to an increase in the pressure inside the containment 172 to the cooling water storage unit 1710 to condense the steam. Accordingly, the facility 1700 for reducing radioactive material may suppress the increasing pressure inside the containment 172 and may reduce the design pressure in the containment 172. As time goes by, the steam may be discharged from the cooling water storage unit 1710 through the opening portion 1710b. However, the radioactive material contained in the steam is captured while passing through the filter facility 1770 and is not discharged to the containment 172. A portion of the steam discharged to the inside of the containment 172 is re-condensed and is recollected to the cooling water storage unit 1710 through the cooling water recollecting portion 1710a. FIG. 8 is a concept view illustrating a facility 1800 for reducing radioactive material and a nuclear power plant 180 having the same according to yet still another embodiment of the present invention. The opening portion 1810b and the cooling water recollecting portion 1810a share the same fluid path. The steam or air in the cooling water storage unit 1810 is discharged overtime through the opening portion 1810b to the containment 182. The condensed water created in the containment 182 is recollected to the cooling water storage unit 1810 through the cooling water recollecting portion 1810a. The filter facility 1870, as shown in FIG. 8, is disposed inside the cooling water storage unit 1810. Specifically, the filter facility 1870 is installed at an upper part of the inner space in the cooling water storage unit 1810. Accordingly, the radioactive material contained in the steam or air is captured while passing through the filter facility 1870 and is restricted for being discharged to the containment 182. FIG. 9 is a concept view illustrating a facility 1900 for reducing radioactive material and a nuclear power plant 190 having the same according to yet still another embodiment of the present invention. The facility 1900 for reducing radioactive material may further include an additive injection unit 1980. The additive injection unit 1980 supplies the cooling water storage unit 1910 with an additive to maintain the pH of the coolant to a predetermined value or more (typically pH 7 or more) so as to prevent volatilization of the radioactive material dissolved in the cooling water storage unit 1910. The additive injection unit 1980 may be installed in the fluid path of the cooling water recollecting portion 1910a as shown in FIG. 4. Radioactive iodine dissolved in the cooling water exists in the form of negative ions. In case the pH of the cooling water in which iodine is dissolved is low, the amount of radioactive iodine that is to be re-volatilized may be significantly increased. This is why the amount of radioactive iodine that is converted into volatilizable elemental iodine (12) is sharply increased in the cooling water of pH 7 or less. The additive injection unit 1980 injects an additive to the cooling water (or condensed water) to prevent the radioactive material dissolved in the cooling water from being re-volatilized. For example, the additive may be sodium phosphate. Sodium phosphate adjusts the pH of the cooling water to prevent corrosion inside the containment 192 and re-volatilization of a radioactive nuclide. However, the type of additives according to the present invention is not limited thereto. The additive may include materials to passively manage the water quality of the cooling water storage unit 1910. For example, boric acid to suppress reactivity of the core 191a or other additives for suppressing corrosion of the device may be added. Referring to FIG. 9, the condensed water in the containment 192 is recollected through the cooling water recollecting portion 1960 to the cooling water storage unit 1910. The additive injection unit 1980 may be installed in the fluid path of the cooling water recollecting portion 1910a to dissolve the additive in the recollected condensed water. Accordingly, if the additive is dissolved in the condensed water introduced to the cooling water recollecting portion 1910a, the additive increases the pH of the condensed water to prevent re-volatilization of the radioactive material. If the condensed water is introduced into the cooling water storage unit 1910 and is mixed with the cooling water, the mixture of the cooling water and the condensed water may be kept at a pH of 7 or more. FIG. 10A is a concept view illustrating a facility 2000 for reducing radioactive material and a nuclear power plant 200 having the same according to yet still another embodiment of the present invention. The nuclear power plant 200 may have a passive containment cooling system along with the facility 2000 for reducing radioactive material. The passive containment cooling system is a system for cooling the inside of the containment 202 to suppress a rise in the pressure inside the containment 202. The passive containment cooling system includes a heat exchanger 206a. The atmosphere inside the containment 202 and the cooling water in the cooling water storage unit 2010 are cooled by the heat exchanger 206a. The steam and air contained in the atmosphere inside the containment 202 may be condensed or cooled. If the temperature inside the containment 202 is decreased, a portion of the steam inside the containment 202 is decreased. Accordingly, the rise in the pressure inside the containment 202 may be suppressed by the passive containment cooling system. The heat exchanger 206a of the passive containment cooling system may be installed in an inner space of the containment 202. Unlike this, the heat exchanger 206a may be installed to be submerged in the cooling water in the cooling water storage unit 2010. The heat exchanger 206a may be installed in both side an inner space of the containment 202 and the cooling water storage unit 2010. Referring to FIG. 10A, a portion of the heat exchanger 206a is disposed in the inner space of the containment 202 and another portion of the heat exchanger 206a is disposed inside the cooling water storage unit 2010. FIG. 10B is a concept view illustrating an example where an accident occurs in the nuclear power plant 200 shown in FIG. 10A. If pipe breakage occurs in a pipe connected with the reactor coolant system 201, steam and radioactive material are discharged through the broken line 203″. The passive safety injection system 205 installed inside the boundary unit 2020 injects a coolant into the reactor coolant system 201. The passive residual heat removing system 204 removes sensible heat in the reactor coolant system 201 and residual heat in the core 201a. As steam is discharged, the pressure inside the boundary unit 2020 is increased to be higher than the pressure inside the containment 202, and a fluid flow is created due to the pressure difference between inside the boundary unit 2020 and inside the containment 202. The connecting pipe 2030 guides the fluid flow to the cooling water storage unit 2010. The sparging unit 2040 sparges, into the cooling water, the fluid and the radioactive material contained in the fluid delivered from the connecting pipe 2030. The soluble radioactive material is collected in the cooling water storage unit 2010. The passive containment cooling system 206 cools at least one of the containment 202 and the cooling water storage unit 2010. As time goes by, the steam or air in the cooling water storage unit 2010 is discharged to the inside of the containment 202 through the opening portion 2010b. However, the radioactive material is captured by the filter facility 2070 installed in the fluid path of the opening portion 2010b and is not discharged to the containment 202. A portion of the steam that has been discharged to the containment 202 is re-condensed to form condensed water. The condensed water is recollected to the cooling water storage unit 2010 through the cooling water recollecting portion 2010a. FIG. 10C is a concept view illustrating a variation to the nuclear power plant 200 shown in FIG. 10B. The passive containment cooling systems 206 and 206′ are formed to cool the atmosphere in the containment 202 and the cooling water in the cooling water storage unit 2010. The heat exchanger (not shown) of the passive containment cooling system may be installed in an inner space of each of the cooling water storage unit 2010 and the containment 202. When an accident occurs, the operation of the facility 2000 for reducing radioactive material, the passive safety injection system 205, and the passive residual heat removing system 204 is the same as that described above in connection with FIG. 10B. The passive containment cooling systems 206 and 206′ cool the atmosphere in the containment 202. Accordingly, the steam evaporated from the cooling water storage unit 2010 to the containment 202 or the atmosphere inside the containment 202 may be cooled or condensed. The condensed water generated as the steam is condensed is collected through the cooling water recollecting portion 2010a, and this has been described above. FIG. 10D is a concept view illustrating another variation to the nuclear power plant 200 shown in FIG. 10B. The passive containment cooling system 206″ is formed to cool the atmosphere in the containment 212 and the cooling water in the cooling water storage unit 2010. The heat exchanger (not shown) of the passive containment cooling system 206″ is formed to penetrate an upper structure of the cooling water storage unit 2010 to simultaneously cool the containment 202 and the cooling water storage unit 2010. Other configurations are the same as those described above in connection with FIG. 10C. FIG. 11 is a concept view illustrating a facility 2100 for reducing radioactive material and a nuclear power plant 210 having the same according to still another embodiment of the present invention. The cooling water storage unit 2110 may be installed at a lower region of an inner space in the containment 212. As in the embodiments described above, the connecting pipe 2130 passes through the inlet 2111 of the cooling water storage unit 2110 and extends to a lower part of the cooling water storage unit 2110. The sparging unit 2140 is connected with the connecting pipe 2130 to receive the radioactive material that has passed through the connecting pipe 2130. The opening portion 2110b is formed to project to an inner space of the containment 212. A filter facility 2170 is installed in a fluid path of the opening portion 2110b. The cooling water recollecting portion 2110a is formed to collect condensed water. The heat exchanger 216c of the passive containment cooling system is installed in the cooling water storage unit 2110 to cool the cooling water in the cooling water storage unit 2110. FIG. 12 is a concept view illustrating a facility 2200 for reducing radioactive material and a nuclear power plant 220 having the same according to yet still another embodiment of the present invention. The facility 2200 for reducing radioactive material further includes a pressure balance line 2250. The pressure balance line 2250 of the facility 2200 for reducing radioactive material needs to be distinguished from the pressure balance line 215d of the passive safety injection system 215. The pressure balance line 2250 of the facility 2200 for reducing radioactive material forms a fluid path that runs from the inside of the containment 222 to the inside of the boundary unit 2220. In case the pressure inside the containment 222 is higher than the pressure inside the boundary unit 2220, the pressure balance line 2250 introduces the atmosphere inside the containment 222 to the inside of the boundary unit 2220. Accordingly, the cooling water in the cooling water storage unit 2210 may be prevented from flowing back to the inside of the boundary unit 2220. The pressure balance line 2050 may be branched from the connecting pipe 2230 and may extend up to the inside of the containment 222. The pressure balance line 2250, as shown, may pass through the upper structure of the cooling water storage unit 2210. The pressure balance line 2250 may have a check valve 2251 that allows for a flow only in one direction. The check valve 2251 prevents the atmosphere inside the boundary unit 2220 from being discharged through the pressure balance line 2250 to the inside of the containment 222. FIG. 13 is a concept view illustrating a facility 2300 for reducing radioactive material and a nuclear power plant 230 having the same according to yet still another embodiment of the present invention. The pressure balance line 2350 forms a fluid path that runs from the inside of the containment 232 to the inside of the boundary unit 2320. The inner space of the boundary unit 2320 and the inner space of the containment 232 are connected with each other by way of the pressure balance line 2350. The pressure balance line 2350, rather than branched from the connecting pipe 2330, is formed independently from the connecting pipe 2330. In this point of view, the pressure balance line 2350 shown in FIG. 13 differs from the pressure balance line 2250 shown in FIG. 12. The pressure balance line 2350 passes through the upper part of the boundary unit 2220 and may extend to the inside of the boundary unit 2220. The check valve 2351 may be installed in the pressure balance line 2350, and the function of the check valve 2351 is the same as that described above in connection with FIG. 7. FIG. 14 is a concept view illustrating a facility 2400 for reducing radioactive material and a nuclear power plant 240 having the same according to yet still another embodiment of the present invention. The connecting pipe 2430 has a check valve 2431 that allows for a flow only in one direction. The check valve 2431 prevents the cooling water in the cooling water storage unit 2410 from flowing back to the boundary unit 2420 through the connecting pipe 2430. FIG. 15 is a concept view illustrating a facility for reducing radioactive material and a nuclear power plant having the same according to yet still another embodiment of the present invention. The cooling water storage unit 2500 may be connected with the safety injection line 255c. The pipe 2532 connecting the cooling water storage unit 2500 with the safety injection line 255c has an isolation valve 2532a and a check valve 2532b. If the isolation valve 2532a and the check valve 2532b are opened, the cooling water stored in the cooling water storage unit 2510 is injected into the reactor coolant system 251. FIG. 16 is a concept view illustrating a facility 2600 for reducing radioactive material and a nuclear power plant 260 having the same according to yet still another embodiment of the present invention. The passive safety injection system 265 may be installed selectively in or outside the boundary unit 2620. Referring to FIG. 16, the passive safety injection system 265 is installed outside the boundary unit 2620. The safety injection line 265c may have an isolation valve 265c1. The isolation valve 265c1 may be installed inside the boundary unit 2620. The pressure balance line 265d may also have isolation valves 265d1 and 265d2. The isolation valves 265d1 and 265d2, respectively, may be installed in and outside the boundary unit 2620. Further, isolation valves or check valves may be added to the inside or outside of the boundary unit 2620. FIG. 17 is a concept view illustrating a facility 2700 for reducing radioactive material and a nuclear power plant 270 having the same according to yet still another embodiment of the present invention. The nuclear power plant 270 includes a feed water system 277 and a feed water supply line 277a. The feed water supply line 277a has an isolation valve 277b. Further, the nuclear power plant 270 includes a turbine system 278 and a steam line 278a. The steam line 278a also has an isolation valve 278b. When the nuclear power plant is in normal operation, water is supplied through the water supply line 277a to the reactor coolant system 271. The water receives heat from the core 271a while passing through the steam generator 271b, and generates steam. The steam may be supplied through the steam line 278a to the turbine system 278. The feed water supply line 277a and the steam line 278a also pass through the boundary unit 2720 and the containment 272. Accordingly, the feed water supply line 277a and the steam line 278a are also examples of the penetration line described above. The boundary unit 2720 extends up to a region adjacent to the containment 272 while surrounding the steam line 278a, the feed water supply line 277a, and the pipes 273 and 273′ penetrating the containment. Accordingly, even when pipe breakage occurs in the boundary unit 2720, the radioactive material cannot exit the boundary unit 2720. Further, the boundary unit 2720 is expanded to the region adjacent to the containment 272, and the possibility that an accident such as feed line or steam line break accident occurs in the region between the boundary unit 2720 and the containment 272 may be significantly lowered. Accordingly, no isolation valve needs to be installed in the region between the boundary unit 2720 and the containment 272. Resultantly, the present invention may reduce the number of isolation valves for closing the pipe line when an accident occurs. FIG. 18 is a concept view illustrating a facility 2800 for reducing radioactive material and a nuclear power plant 280 having the same according to yet still another embodiment of the present invention. Additive injection units 2880 include a first additive injection unit 2881 and a second additive injection unit 2882. The first additive injection unit 2881 is installed inside the cooling water storage unit 2810. The second additive injection unit 2882 may be installed in a fluid path of the cooling water recollecting portion 2810a. The first additive injection unit 2881 may be installed at a predetermined height from the bottom of the cooling water storage unit 2810 to be submerged in the cooling water as the water level of the cooling water increases. If the fluid in the boundary unit 2820 is continuously sparged into the cooling water storage unit 2810, the water level of the cooling water storage unit 2810 gradually increases. If the water level of the cooling water storage unit 2810 is higher than the water level of the additive injection unit 2881, the additive injection unit 2881 is submerged in the cooling water. As the additive injection unit 2881 is submerged in the cooling water, the additive is dissolved in the cooling water. Further, the second additive injection unit 2882 dissolves the additive in the condensed water recollected through the cooling water recollecting portion 2810a as described above. FIG. 19 is a concept view illustrating a facility 2900 for reducing radioactive material and a nuclear power plant 290 having the same according to yet still another embodiment of the present invention. The facility 2900 for reducing radioactive material is formed to configure a boundary of radioactive material between the containment 292 and the reactor coolant system 291. The facility 2900 for reducing radioactive material is configured to capture radioactive material that may be discharged to the containment 292 when an accident occurs in the nuclear power plant 290. The facility 2900 for reducing radioactive material includes a boundary unit 2920, a discharging unit 2930, and a filter facility 2970. The boundary unit 2920 is installed inside the containment 292. The boundary unit 2920 forms a boundary of radioactive material in the containment 292. When an accident occurs, radioactive material may release from the reactor coolant system 291 or pipes 293, 293′, and 295c connected with the reactor coolant system 291 to the inside of the containment 292. The boundary unit 2920 wraps around the reactor coolant system 291 and the pipes 293, 293′ and 295c to prevent release of radioactive material to the containment 292. The design pressure for radioactive material formed by the boundary unit 2920 is designed to withstand the pressure difference of a flow discharged from the discharging unit 2930 when an accident occurs. At least a portion of the boundary unit 2920 may be formed by a concrete structure inside the containment 292. Further, at least a portion of the boundary unit 2920 may be formed by a coating member such as a steel liner et. al. installed on the concrete structure. The boundary unit 2920 may include a barrier 2921 and a cover 2922. The barrier 2921 is formed to wrap around the reactor coolant system 291. As shown in FIG. 19, the barrier 2921 is configured to wrap around the remaining part except the upper part of the reactor coolant system 291 at a position spaced apart from the reactor coolant system 291. The cover 2922 is formed to cover the upper part of the reactor coolant system 291 and is coupled with the barrier 2921. Accordingly, at the time the reactor coolant system 291 disposed inside the boundary unit 2920 needs maintenance, the cover 2922 may be separated from the barrier 2921 to expose the reactor coolant system 291. The nuclear power plant 290 includes penetration pipes 293, 293′ and 295c penetrating the containment 292. The terms “penetration pipes 293, 293′ and 295c” may be used to denote all the pipes that have the feature of penetrating the containment 292. For example, if a pipe used to make a primary fluid flow and a pipe used to make a secondary fluid flow penetrate the containment 292, the pipes belong to the penetration pipes 293, 293′ and 295c. Further, the safety injection line 295c that runs to the reactor coolant system 291 to form a safety injection fluid path also belongs to the penetration pipes 293, 293′ and 295c. The penetration pipes 293, 293′ and 295c are connected to the reactor coolant system 291 or connected to a secondary system. The penetration pipes 293 and 203′ may have isolation valves 293a, 293b, and 293a′ or check valves 293b′ at positions spaced apart from each other to doubly close the containment 292 and the boundary unit 2920 when breakage occurs. If the boundary unit 2920 and the containment 292 are spaced apart from each other and the penetration pipes 293 and 293′ pass through the region between the boundary unit 2920 and the containment 292, an accident may occur due to breakage of the penetration pipes 293 and 293′ in the region between the boundary unit 2920 and the containment 292. In such case, the radioactive material might not be trapped in the reactive boundary unit 2920. The isolation valves 293a, 293b, and 293a′ have a mechanism to be opened and closed in response to a related safety system signal and thus may be likely to malfunction or halt. The check valve 293b′ has a moving part and thus it is impossible to remove the possibility of malfunctioning or halting. For the above reasons, the possibility of occurrence of a single failure may be granted an exception for some high-reliability devices, but the nuclear power plant 290 is basically designed to assume occurrence of a single failure when an accident occurs. Accordingly, considering a single failure, the isolation valves 293a, 293b, and 293a′ or check valves 293b′ should be installed at the portions of the penetration pipes 293 and 293′ disposed between the containment 292 and the boundary unit 2920 to prevent additional release of radioactive material. However, addition of the isolation valves 293a, 293b, and 293a′ or check valves 293b′ is not preferred in view of simplifying the facility. To address such issue, the present invention provides a structure that may prevent release of radioactive material even without installation of additional isolation valves 293a, 293b, and 293a′ or check valves 293b′. Hereinafter, the structure is described in detail. At least a portion of the boundary unit 2920 is expanded up to a region adjacent to the containment 292 while surrounding the penetration pipes 293 and 293′ to prevent an accident from occurring due to breakage of the penetration pipes 293 and 293′ in a region between the boundary unit 2920 and the containment 292. Due to such expanded structure of the boundary unit 2920, the portions of the penetration pipes 293 and 293′, which pass through the containment 292 to the inside of the containment 292 are mostly positioned inside the boundary unit 2920. Accordingly, the present invention may significantly lower, by the expanded structure of the boundary unit 2920, the possibility that a loss-of-coolant accident, feed line break accident or steam line break accident occurs due to, e.g., breakage of the penetration pipes 293 and 293′ in the region between the boundary unit 2920 and the containment 292. The penetration pipes 293 and 293′ may have a portion (first portion) disposed outside the containment 292, a portion (second portion) disposed inside the boundary unit 2920, and a portion (third portion) disposed between the containment 292 and the boundary unit 2920. Under accident, as a combination of valves for isolating the containment 292 from the boundary unit 2920, isolation valves 293a, 293b, and 293a′ or check valve 293b′ may be selectively adopted considering the direction of a flow in the penetration pipes 293 and 293′ and flow resistance according to the characteristics of the nuclear power plant. The expanded structure of the boundary unit 2920 is configured to minimize the gap between the containment 292 and the boundary unit 2920. Accordingly, the expanded structure of the boundary unit 2920, even without additional installation of the isolation valves 293a, 293b, and 293a′ at the third portion, may exclude the possibility that the penetration pipes 293 and 293′ are broken at the third portion. The discharging unit 2930 is installed at the boundary of radioactive material to form a fluid path that runs from the boundary unit 2920 to the containment 292. If a pressure difference is created between the containment 292 and the boundary unit 2920, the fluid flows from a place with a relatively high pressure to a place with a relatively low pressure. For example, when a loss-of-coolant accident occurs due to, e.g., pipe breakage, steam may be discharged from the reactor coolant system 291 or pipe connected with the reactor coolant system 291. In such case, the pressure inside the boundary unit 2920 is rendered to be higher than the pressure inside the containment 292. Accordingly, the fluid inside the containment 2920 is caused to flow to the containment 292. As used herein, the term “pressure inside the containment 292” refers to the pressure in the remaining space except the inner space in the containment 292 of the boundary unit 2920. The discharging unit 2930 is configured to guide the fluid flow caused by the pressure difference between the containment 292 and the boundary unit 2920 from the boundary unit 2920 through the fluid path to the containment 292. The boundary unit 2920 forms a sealing structure around the reactor coolant system 291 to prevent the fluid from flowing from the boundary unit 2920 to the containment 292 through a path other than the fluid path formed by the discharging unit 2930. For example, the boundary unit 2920 may be configured to surround the reactor coolant system 291 at the position spaced apart from the reactor coolant system 291. Accordingly, the fluid inside the boundary unit 2920 may be discharged into the containment 292 only through the fluid path formed by the discharging unit 2930 but cannot be discharged via other paths. As used herein, the term “inside the containment 292” refers to the remaining space in the containment 292 other than the inner space of the boundary unit 2920. The filter facility 2970 is installed in the fluid path of the discharging unit 2930 to capture the radioactive material contained in the fluid passing through the discharging unit 2930 in the boundary unit 2920. The filter facility 2970 is configured to capture radioactive material in the boundary unit 2920 while the atmosphere inside the boundary unit 2920 is discharged through the fluid path of the discharging unit 2930 to the inside of the containment 292. The filter facility 2970 includes at least one of a filter and an absorbent. The term “additive” may be interchangeably used with the term “absorbent.” As the filter, a high-efficiency particulate air (HEPA) filter may be adopted. The gaseous radioactive material contained in the fluid is removed while passing through the filter. For example, in case the radioactive material is iodine, iodine is combined with silver nitrate while passing through the filter to thus turn into iodic silver. Iodic silver may be separated from the fluid. Accordingly, the filter is configured to allow silver nitrate react with iodine contained in the fluid to form iodic silver. The filter is formed to eliminate iodic silver from the fluid. As the absorbent, charcoal may be employed. Organic iodine compounds are combined with the materials impregnated in the charcoal to turn into quaternary ammonium salt and are absorbed into the charcoal. Molecular iodine is combined with the charcoal through chemisorption. The charcoal is utilized as an absorbent material thanks to its large internal surface area. Accordingly, the absorbent is configured to remove iodine contained in the fluid through chemisorption that is made by charcoal. However, the above-described filter and the absorbent are merely an example, and the type of filter and absorbent according to the present invention is not limited thereto. In order to prevent damage to the containment 292 that may occur due to a significant increase in the pressure inside the containment 292 and occurrence of an accident and to decrease the concentration of radioactive material discharged to the external environment, AREVA, France, and Westinghouse, U.S., have developed a filtered containment ventilation system (FCVS). The FCVS has a filter facility at the boundary between the inside and outside of the containment 292 and opens the boundary (using a breaking plate or valve) when an accident occurs to significantly increase the pressure inside the containment 292, and discharges the atmosphere inside the containment 292 through the filter facility. In case a beyond design basis accident (the beyond design basis accident refers to an accident that causes the pressure inside the containment 292 to be significantly increased to a design pressure or more) occurs in the nuclear power plant 290 adopting the FCVS, the breaking plate or valve installed between the inside of the containment 292 and the filter facility is opened, and a flow is caused by the pressure difference between the inside and outside of the containment 292 (between the high pressure created inside the containment 292 and the atmospheric pressure outside the containment 292). The flow causes the atmosphere (air and steam) inside the containment 292 to pass through the filter facility and to be then discharged to the outside of the containment 292. However, the above-described, conventional FCVS is not operated when the design basis accident occurs, and the radioactive material is directly discharged to the inside of the containment 292. Accordingly, the conventional FCVS, upon occurrence of a design basis accident, cannot lower the concentration of radioactive material inside the containment 292 and cannot resultantly suppress a certain amount of radioactive material releasing to the outside of the containment 292. In contrast, the present invention is configured to operate even when all types of accidents occur including a design basis accident and beyond design basis accident. The present invention is configured to force radioactive material in the boundary unit 2920 and to discharge a fluid having a low concentration of radioactive material to be discharged to the containment 292. The radioactive material is captured in the boundary unit 2920 while passing through the filter facility 2970. The present invention may reduce the concentration of radioactive material in the containment 292 in a very efficient manner, thus leading to a significant reduction in the amount of radioactive material releasing to the outside of the containment 292. The nuclear power plant 290 may further include a passive safety injection system 295 configured to inject a coolant into the reactor coolant system 291 using a natural force when an accident occurs. The passive safety injection system 295 may include a core makeup tank 295a and a safety injection tank 295b. The core makeup tank 295a is formed to store a coolant such as low-temperature boric acid solution. The core makeup tank 295a is installed to have a height gap from the reactor coolant system 291. The core makeup tank 295a and the reactor coolant system 291 may be connected with each other by the pressure balance line 295d. The pressure balance line 295d is configured to form a pressure balance between the reactor coolant system 291 and the core makeup tank 295a and is for allowing for coolant injection from the core makeup tank 295a by gravity. The safety injection tank 295b is formed to store a coolant such as low-temperature boric acid solution. The safety injection tank 295b and the reactor coolant system 291 may be connected with each other through the pressure balance line 295d. The safety injection tank 295b may be filled with some gas (typically, nitrogen gas). The pressure of the gas is set to be lower than the pressure of the reactor coolant system 291 that is in normal operation. When the nuclear power plant 290 is in normal plant operation, the safety injection tank 295b is isolated by the check valve, so that the coolant inside the safety injection tank 295b is not injected to the reactor coolant system 291. The passive safety injection system 295 includes a safety injection line 295c connected with the reactor coolant system 291 to form an injection fluid path for coolant. The core makeup tank 295a and the safety injection tank 295b are connected with the reactor coolant system 291 through the safety injection line 295c. The safety injection line 295c forms a fluid path for the coolant injected from the core makeup tank 295a and the safety injection tank 295b to the reactor coolant system 291. The safety injection line 295c may penetrate the containment 292. Accordingly, the safety injection line 295c may be configured of an example of the above-described penetration pipes 293 and 293′. The expanded structure of the boundary unit 2920 may apply to the safety injection line 295c as well. At least a portion of the boundary unit 2920 may be expanded up to a region adjacent to the containment 292 while surrounding the safety injection line 295c to prevent a loss-of-coolant accident that may occur due to breakage of the safety injection line 295c between the boundary unit 2920 and the containment 292. The other description of the expanded structure of the boundary unit 2920 is replaced by what has been described above therefor. FIG. 19 is a view illustrating a normal operation state of the nuclear power plant 290. Accordingly, the isolation valves 293a, 293b, and 293a′ installed on the pipes 293 and 293′ for normal operation of the nuclear power plant 290 stay opened. In the normal operation of the nuclear power plant 290, no steam is discharged from the reactor coolant system 291, and thus, the pressure balance is maintained between the boundary unit 2920 and the containment 292. FIG. 20 is a concept view illustrating a facility 3000 for reducing radioactive material and a nuclear power plant 300 having the same according to yet still another embodiment of the present invention. The facility for reducing radioactive material includes a boundary unit 3020, a discharging unit 3030, and a filter facility 3070. The facility for reducing radioactive material further includes a cooling water storage unit 3010, a cooling water recollecting portion 3010a and an opening portion 3010b. The cooling water storage unit 3010 is installed inside the containment 302. For example, the cooling water storage unit 3010 may be installed in an upper or lower part of an inner space of the containment 302. The cooling water storage unit 3010 is formed to store cooling water and may be formed as a tank or pool. Among other radioactive materials spread to the external environment when an accident occurs in the nuclear power plant 300, iodine may have a highest concentration. Iodine, when contacting water, is mostly dissolved in the water. The cooling water storage unit 3010 retains cooling water that may dissolve iodine. Most of the radioactive materials are captured in the boundary unit 3020 by the filter facility 3070 while passing through the discharging unit 3030. However, a small amount of radioactive material is not captured by the filter facility 3070 and may be discharged to the containment 302 or a small amount of radioactive material may leak from the boundary unit 3020. However, a small amount of radioactive material discharged to the containment 302, if dissolved by sprayed or condensed water of other containment 302 safety systems (for example, a containment spray system or cooling system) that may be employed as per the characteristics of the nuclear power plant 300 to be captured in the cooling water of the cooling water storage unit 3010, may be cut off from releasing to the external environment. Accordingly, the cooling water storage unit 3010 may support the function of the filter facility 3070. The cooling water recollecting portion 3010a forms a fluid path that runs from the containment 302 to the cooling water storage unit 3010 to recollect the condensed water created from the fluid discharged through the discharging unit 3030 to the containment 302 to the cooling water storage unit 3010. However, in case the safety system is configured in combination with the spray system (not shown), the sprayed cooling water is also recollected to the cooling water recollecting portion 3010a. For example, the cooling water recollecting portion 3010a may be disposed to be adjacent to the inner wall of the containment 302 so that the condensed water flowing down the inner wall of the containment 302 is collected to the cooling water storage unit 3010. However, the shape of the cooling water recollecting portion 3010 may be selectively adopted according to the characteristics of the nuclear power plant 300. According to the present invention, the cooling water recollecting portion 3010a has a structure of introducing the cooling water inside the containment 302 such as sprayed water or condensed water to the cooling water storage unit 3010 and is not limited to a special shape of the cooling water recollecting portion 3010a. A portion of the fluid discharged through the discharging unit 3030 to the containment 302 is condensed to form condensed water. The concentration of boric acid in the condensed water is low, and the condensed water may contain a small amount of radioactive material. The condensed water is recollected from the containment 302 through the cooling water recollecting portion 3010a to the cooling water storage unit 3010. The opening portion 3010b is formed by opening at least a portion of the cooling water storage unit 3010 to maintain a pressure balance between the inside of the containment 302 and the cooling water storage unit 3010. If a pressure difference is created between the cooling water storage unit 3010 and the containment 302, the cooling water storage unit 3010 and the containment 302 may re-form a pressure balance by the opening portion 3010b. The cooling water storage unit 3010 may be configured of a single facility for the facility 3000 for reducing radioactive material only, but may be shared with other systems (for example, passive safety injection system 305, residual heat removing system, etc.). Hereinafter, an example where the facility 3000 for reducing radioactive material and the passive safety injection system 305 share the cooling water storage unit 3010 is described. The cooling water storage unit 3010 is connected to the safety injection line 305c to inject the cooling water retained therein to the inside of the reactor coolant system 301. The cooling water storage unit 3010 is installed at a higher position than the reactor coolant system 301. The pipe 3012 connecting the cooling water storage unit 3010 with the safety injection line 305c has an isolation valve 3012a and a check valve 3012b. If the isolation valve 3012a is opened by a related signal when an accident occurs, a cooling water flow is generated from the cooling water storage unit 3010 to the reactor coolant system 301. The check valve 3012b is opened by the flow of cooling water, and the cooling water is injected through the safety injection line 305c to the reactor coolant system 301. FIG. 21 is a concept view illustrating a facility 3100 for reducing radioactive material and a nuclear power plant 310 having the same according to yet still another embodiment of the present invention. The facility 3100 for reducing radioactive material further includes an additive injection unit 3180. The additive injection unit 3180 supplies an additive to increase the pH of the coolant to a predetermined value or more (typically pH 7 or more) to prevent volatilization of the radioactive material dissolved in the cooling water storage unit 3110. As illustrated, the additive injection unit 3180 may be installed in a fluid path of the cooling water recollecting portion 3110a to dissolve the additive in the cooling water, such as sprayed or condensed water, to the cooling water storage unit 3110. Radioactive iodine dissolved in the cooling water exists in the form of negative ions. In case the pH of the cooling water in which iodine is dissolved is low, the amount of radioactive iodine that is to be re-volatilized may be significantly increased. This is why the amount of radioactive iodine that is converted into volatilizable elemental iodine (12) is sharply increased in the cooling water of pH 7 or less. Besides, the amount that turns into elemental iodine is associated with the temperature of the cooling water and the concentration of iodine in the solution. The elemental iodine may be re-volatilized in the atmosphere according to a separation coefficient defined as a ratio in concentration of iodine in the atmosphere to iodine in the cooling water. According to related regulations, in case the pH of the cooling water is higher than 7.0, the amount that turns into elemental iodine is significantly reduced, so that re-volatilization may be negligible. The additive injection unit 3180 supplies an additive to the cooling water, such as sprayed or condensed water, recollected to the cooling water storage unit 3110 to prevent re-volatilization of radioactive material. As the additive, sodium phosphate may be adopted. Sodium phosphate adjusts the pH of the cooling water to prevent re-volatilization of the radioactive nuclide or corrosion of the inside of the containment 312 upon accident. However, the type of additive according to the present invention is necessarily limited thereto. Boric acid to suppress the reactivity of the core 311a or other additives to suppress corrosion of the device may be added so that the water quality of the cooling water storage unit 3110 is passively managed. FIG. 22 is a concept view illustrating a facility 3200 for reducing radioactive material and a nuclear power plant 320 having the same according to yet still another embodiment of the present invention. The nuclear power plant 320 includes a steam generator 321b. The steam generator 321b is installed at the boundary between the primary system and the secondary system and generates steam through heat transfer of the primary fluid and secondary fluid. The steam generator 321b forms a pressure boundary between the fluid path of the primary fluid and the fluid path of the secondary fluid path and thus the primary fluid and the secondary fluid are not mixed with each other. In the normal operation of the nuclear power plant 320, the feed water system 327 supplies water (secondary fluid) through the feed water supply line 327a to the steam generator 321b. The heat generated in the core 321a is transferred to the primary fluid, and the primary fluid transfers heat to the secondary fluid while passing through the steam generator 321b. The supplied water receives heat from the primary fluid while passing through the steam generator 321b, and turns into steam. The steam discharged from the steam generator 321b is delivered through the steam line 328a to the turbine system 328. In normal operation of the nuclear power plant 320, the isolation valves 327b and 328b installed on the feed water supply line 327a and the steam line 328a remain opened. The feed water supply line 327a and the steam line 328a may pass through the containment 322. Accordingly, the feed water supply line 327a and the steam line 328a may be configured as examples of the above-described penetration pipes 323 and 323′. At least a portion of the boundary unit 3220 may be expanded up to a region adjacent to the containment 322 while surrounding the feed water supply line 327a and the steam line 328a as well as the penetration pipes 323 and 323′ to prevent an accident from occurring due to breakage of the feed water supply line 327a and the steam line 328a between the boundary unit 3220 and the containment 322. Accordingly, the present invention may significantly lower the possibility that a steam line break accident or feed line break accident occurs due to breakage of the water supply line 327a or steam line 328a in the region between the boundary unit 3220 and the containment 322 by way of the expanded structure of the boundary unit 3220. The facility 3200 for reducing radioactive material may minimize the gap between the boundary unit 3220 and the containment 322 by the expanded structure of the boundary unit 3220 to exclude the possibility that the penetration pipes 323 and 323′ occur at the portion therebetween. Unlike described above, various tanks 325a and 325b of the passive safety injection system 325, rather than positioned inside the boundary unit 3220, may be disposed between the boundary unit 3220 and the containment 322. The safety injection line 325c may be split into a portion disposed inside the boundary unit 3220, a portion disposed between the boundary unit 3220 and the containment 322, and a portion disposed outside the containment 322. Since the passive safety injection system 325 is disposed outside the boundary unit 3220, check valves 325f and 325f′ are added to the safety injection line 325c, and isolation valves 325e and 325e′ are added to the pressure balance line 325d. However, the check valves 325f and 325f′ or isolation valves 325e and 325e′ may be selectively adopted considering the conditions such as direction of flow and flow resistance. FIG. 23 is a concept view illustrating a facility 3300 for reducing radioactive material and a nuclear power plant 330 having the same according to yet still another embodiment of the present invention. The nuclear power plant 330 further includes a containment cooling system configured to suppress a rise in pressure inside the containment 332. The containment cooling system may be a passive containment cooling system that suppresses a rise in the pressure inside the containment 332 using natural circulation. The passive containment cooling system has a heat exchanger 336b. The heat exchanger 336b, as shown in FIG. 23, may be installed in the atmosphere of the containment 332. However, the position of the heat exchanger 336b is not necessarily limited thereto, and may be disposed in the cooling water storage unit 3310. The cooling fluid is heat-exchanged with the atmosphere inside the containment 332 while passing through the heat exchanger 336b and is heated. The density of the heated cooling fluid is reduced, and the cooling fluid goes up along the fluid path of the heat exchanger 336b. The cooling fluid is discharged from the heat exchanger 336b to the outside of the containment 332. The steam discharged from the boundary unit 3320 to the containment 332 is condensed in the heat exchanger 336b by natural circulation. The phenomenon that the steam is condensed to turn into condensed water reduces the steam partial pressure inside the containment 332 and thus functions to suppress a rise in the pressure inside the containment 332. Typically, the passive containment cooling system shows a lower efficiency of reducing radioactive material as compared with the typical active containment spray system. However, in case the facility 3300 for reducing radioactive material proposed herein is adopted along with the passive containment cooling system, the concentration of radioactive material discharged by the facility 3300 for reducing radioactive material to the inside of the containment 332 may be remarkably reduced to solving the problems of the passive containment cooling system. The cooling water recollecting portion 3310a is disposed at a lower part of the heat exchanger 336b to recollect the condensed water created by the operation of the heat exchanger 336b to the cooling water storage unit 3310. The condensed water generated in the heat exchanger 336b may be dropped and recollected to the cooling water storage unit 3310 through the fluid path of the cooling water recollecting portion 3310a. In the process of recollecting the condensed water, the condensed water may be supplied with an additive from the additive injection unit 3380. Accordingly, the pH of the condensed water may be adjusted, and the condensed water may be prevented from re-volatilization. FIG. 24A is a concept view illustrating the normal operation of a facility 3400 for reducing radioactive material and a nuclear power plant 340 having the same according to yet still another embodiment of the present invention. The heat exchanger 346a of the passive containment cooling system may be formed to cool both the cooling water in the cooling water storage unit 3410 and the atmosphere in the containment 342. At least a portion of the heat exchanger 346a is submerged in the cooling water storage unit 3410 and may be extended up to the inner space of the containment 342 from the cooling water storage unit 3410. Among the pipes 343, 343′ and 345c penetrating the containment 342, the pipes 343 and 343′ for normal operation of the nuclear power plant 340 allow the fluid to flow therethrough. The isolation valves 343a, 343b, and 343a′ and the check valve 343b′ installed on the pipes 343 and 343′ are required for normal operation of the nuclear power plant 340. During the normal operation of the nuclear power plant 340, the isolation valves 343a, 343b, and 343a′ and the check valve 343b′ installed on the pipes 343 and 343′ remain opened. FIG. 24B is a concept view illustrating an example in which an accident occurs in a facility 3400 for reducing radioactive material and a nuclear power plant 340 having the same according to yet still another embodiment of the present invention. If an accident occurs in the nuclear power plant 340, the nuclear power plant 340 may remain in safe shutdown condition by the operation of various safety systems. The passive residual heat removing system 344 removes sensible heat in the reactor coolant system 341 and residual heat in the core 341a. The passive containment cooling system suppresses a rise in the pressure inside the containment 342. The passive safety injection system 345 maintains the water level of the reactor coolant system 341. The facility 3400 for reducing radioactive material captures radioactive material in the boundary unit 3420. If an accident such as a loss-of-coolant accident occurs, steam is discharged from the broken line 343f. The discharged steam may be mixed with the atmosphere present inside the boundary unit 3420. Since upon accident the isolation valves 343a, 343b, and 343a′ and the check valve 343b′ are closed, the fluid does not flow any longer through the pipes 343 and 343′ for normal operation of the nuclear power plant 340. As steam is continuously discharged from the reactor coolant system 341, the pressure inside the boundary unit 3420 is gradually increased, and a pressure difference is generated between the inside of the boundary unit 3420 and the inside of the containment 342. The fluid created as the atmosphere and steam are mixed with each other forms a flow by the pressure difference. The fluid is discharged from the boundary unit 3420 through the discharging unit 3430 to the containment 342. The radioactive material contained in the fluid is captured in the boundary unit 3420 while passing through the filter facility 3470 installed in the discharging unit 3430. The remaining fluid except the radioactive material is discharged to the inner space of the containment 342. The fluid discharged to the containment 342 is mixed with the atmosphere in the containment 342. Accordingly, the pressure and temperature of the containment 342 are gradually increased. However, the heat exchanger 348 of the passive containment cooling system is operated to suppress a rise in the pressure of the containment 342. The atmosphere in the containment 342 (including the fluid discharged to the containment 342) and the fluid supplied from the outside of the containment 342 exchange heat with each other while flowing through different fluid paths from each other. Accordingly, the atmosphere in the containment 342 is cooled and condensed in the heat exchanger 346a by natural circulation. The atmosphere in the containment 342 is cooled and condensed by the operation of the heat exchanger 346a. The air contained in the atmosphere of the containment 342 is discharged back to the inside of the containment 342, and the condensed water generated as the steam is condensed is recollected to the cooling water storage unit 3410 through the cooling water recollecting portion 3410a. In this process, the condensed water is supplied with an additive for preventing re-volatilization from the additive injection unit 3480. Accordingly, the condensed water is recollected to the cooling water storage unit 3410, and the condensed water may be prevented from re-volatilization. The cooling water stored in the cooling water storage unit 3410 is cooled by the heat exchanger 346a by natural circulation. The passive safety injection system 345 injects cooling water to the reactor coolant system 341. If upon accident a phenomenon such as reduction in pressure of the reactor coolant system 341 occurs, the isolation valve 345a′ installed on the pipe connecting the core makeup tank 345a with the safety injection line 345c is opened in response to a related signal. A flow of the cooling water is caused by gravity water head from the core makeup tank 345a to the reactor coolant system 341, and the check valve 345a″ is opened by the flow of the cooling water. The cooling water is injected from the core makeup tank 345a through the safety injection line 345c to the reactor coolant system 341. If, upon accident, a phenomenon in which the pressure inside the reactor coolant system 341 is reduced to a predetermined value or less, for example, the isolation valve 345d′ installed on the pressure balance line 345d is opened by a related signal. The steam supplied from the reactor coolant system 341 is injected to the safety injection tank 345b through the pressure balance line 345d, and the pressure inside the safety injection tank 345b increases. If the pressure balance is formed between the reactor coolant system 341 and the safety injection tank 345b, the cooling water inside the safety injection tank 345b is injected to the reactor coolant system 341 by gravity water head. The check valve 345b′ is opened by the flow of the cooling water, and the cooling water is injected to the reactor coolant system 341 through the safety injection line 345c. The cooling water retained in the cooling water storage unit 3410 may be used for safety injection. The isolation valve 3412a installed on the pipe 3412 connecting the cooling water storage unit 3410 with the safety injection line 345c is opened by a related signal, and as the reactor coolant system 341 is cooled after accident, the pressure inside the reactor coolant system 341 and the pressure inside the cooling water storage unit 3410 form a pseudo-balanced state, a flow of the cooling water is caused by gravity from the cooling water storage unit 3410. The term “pseudo-balance” refers to a state that is not the theoretically complete balanced state but is close to the balanced state enough to form a flow of the cooling water. As the cooling water flows, the check valve 3412b is opened, and the cooling water may be injected to the reactor coolant system 341. FIG. 25A is a concept view illustrating the normal operation of a facility 3500 for reducing radioactive material and a nuclear power plant 350 having the same according to yet still another embodiment of the present invention. The nuclear power plant 350 further includes an extended path 3531 and a circulation enhancement facility 359. The extended path 3531 is extended from the discharging unit 3530 up to an upper part of the heat exchanger 356b to discharge the fluid from the discharging unit 3530 to the heat exchanger 356b. The fluid in the boundary unit 3520 flows along the extended path 3531 and is discharged through the outlet of the extended path 3531. The circulation enhancement facility 359 is installed at the outlet of the extended path 3531. The fluid is discharged through the circulation enhancement facility 359. The circulation enhancement facility 359 may be configured in the form of a jet pump, for example. The circulation enhancement facility 359 is configured to introduce the atmosphere included in the containment 352 by a pressure decrease that is caused as the fluid is discharged with high velocity. The circulation enhancement facility 359 is configured to inject the introduced atmosphere together with the fluid. The circulation enhancement facility 359 includes a zet nozzle unit 359a and an atmosphere entrainment unit 359b. The zet nozzle unit 359a is connected with the outlet of the extended path 3531 to receive the fluid from the extended path 3531. The zet nozzle unit 359a is formed to inject the received fluid to the heat exchanger 356b. The atmosphere entrainment unit 359b wraps around the zet nozzle unit 359a at the position spaced apart from the zet nozzle unit 359a to form an atmosphere inlet space around the zet nozzle unit 359a. For example, the atmosphere entrainment unit 359b may form a ring-shaped atmosphere inlet space around the zet nozzle unit 359a. The atmosphere entrainment unit 359b is configured to inject the atmosphere introduced through the atmosphere inlet space, together with the fluid. The atmosphere in the containment 352 may be circulated more actively by the circulation enhancement facility 359. This means that a small amount of the remaining radioactive material and the steam released to the inside of the containment 352 may be guided to the heat exchanger 356b. Accordingly, the steam may be condensed, and the soluble radioactive material may be dissolved in the condensed water and recollected to the cooling water storage unit 3510. Further, the circulation enhancement facility 359 mitigates a decrease in the efficiency of the heat exchanger 356b that occurs due to accumulation of a noncondensable gas (air) around the heat exchanger 356b. The performance of the heat exchanger 356b may be enhanced through forced circulation by the circulation enhancement facility 359. Further, the circulation enhancement facility 359 may increase the speed of flow at the periphery of the heat exchanger 356b to assist in enhancing the heat transfer coefficient. FIG. 25B is a concept view illustrating an example in which an accident occurs in a facility 3500 for reducing radioactive material and a nuclear power plant 350 having the same according to yet still another embodiment of the present invention. When an accident occurs in the nuclear power plant 350, the nuclear power plant 350 may stay in safe shutdown condition by the operation of various safety systems. The passive residual heat removing system removes sensible heat in the reactor coolant system 351 and residual heat in the core 351a. The passive containment cooling system 355 maintains the water level of the reactor coolant system 351. The facility 3500 for reducing radioactive material captures radioactive material in the boundary unit 3520. The fluid inside the boundary unit 3520 flows along the extended path 3531 connected with the discharging unit 3530 and is injected to the heat exchanger 356b through the zet nozzle unit 359a. If the fluid is injected with a high speed, a pressure drop phenomenon locally occurs. Accordingly, the atmosphere inside the containment 352 is introduced to the atmosphere entrainment unit through the atmosphere inlet space, and the atmosphere entrainment unit 359b injects the introduced atmosphere, together with the fluid, to the heat exchanger 356b. The atmosphere and fluid are cooled and condensed in the heat exchanger 356b. The air is discharged, and the condensed water created by the operation of the heat exchanger is recollected to the cooling water storage unit 3510 through the cooling water recollecting portion 3510a. Since the additive injection unit 3580 injects an additive to the condensed water during the process of recollecting the condensed water, the condensed water may be prevented from re-volatilization. The description of the others is replaced with what has been described above. FIG. 26A is a concept view illustrating the normal operation of a facility 3600 for reducing radioactive material and a nuclear power plant 360 having the same according to yet still another embodiment of the present invention. The discharging unit 3630 is extended from the boundary unit 3620 up to the inside of the cooling water storage unit 3610 to discharge the atmosphere inside the boundary unit 3620 to the cooling water storage unit 3610. The outlet of the discharging unit 3630 is submerged in the cooling water of the cooling water storage unit 3610. Accordingly, the fluid in the boundary unit 3620 is not directly discharged to the containment 362 and is discharged to the cooling water in the cooling water storage unit 3610. The facility 3600 for reducing radioactive material further includes a sparging unit 3640. The sparging unit 3640 is installed at an end of the discharging unit 3630 to be submerged in the cooling water of the cooling water storage unit 3610. The sparging unit 3640 sparges the fluid that has passed through the discharging unit 3630. The fluid contains steam and air, and the sparging unit 3640 is configured to sparge the air while condensing the steam. In case the facility 3600 for reducing radioactive material includes the sparging unit 3640, the design pressure for the boundary of radioactive material formed by the boundary unit 3620 is designed considering water head. The sparging unit 3640 may have a flow resistance in its inner fluid path to induce an even distribution of the fluid to the plurality of fine fluid paths. The fluid may be relatively evenly distributed to each fine fluid path by the flow resistance. As the steam is condensed, the pressure inside the containment 362 may be suppressed from increasing. Since the non-condensed air inside the boundary unit 3620 is discharged to the inside of the containment 362, the pressure inside the containment 362 may increase. However, since the volume of the inside of the boundary unit 3620 is relatively smaller than the volume of the inside of the containment 362, the pressure inside the containment 362 is not greatly increased. The containment 362 and the boundary unit 3620 may be connected with each other via a pressure balance line (not shown). The pressure balance line may have a check valve (not shown), and the sparging unit 3640, unlike shown, may be installed on the pressure balance line. In case as long-term cooling or a loss-of-coolant accident occurs outside the boundary unit 3620, the pressure inside the containment 362 is higher than the pressure inside the boundary unit 3620, the check valve of the pressure balance line is opened, and the containment 362 and the boundary unit 3620 form a pressure balance. Since the atmosphere inside the containment 362 is introduced to the inside of the boundary unit 3620 through the pressure balance line, the pressure balance line may prevent the cooling water in the cooling water storage unit 3610 from flowing back to the inside of the boundary unit 3620. The additive injection unit 3680 may be installed in each of the cooling water storage unit 3610 and the cooling water recollecting portion 3610a. The first additive injection unit 3681 is installed in the cooling water storage unit 3610. The second additive injection unit 3682 is installed in the cooling water recollecting portion 3610a. The first additive injection unit 3681 is installed at a predetermined height of the cooling water storage unit 3610 to be submerged in the cooling water by a rise in the water level of the cooling water storage unit 3610. As the first additive injection unit 3681 is submerged in the cooling water, the additive is dissolved in the cooling water, and thus, the first additive injection unit 3681 may prevent the radioactive material from volatilizing. The second additive injection unit 3682 injects an additive to the condensed water recollected through the cooling water recollecting portion 3610a to the cooling water storage unit 3610. The description of the functions of the second additive injection unit 3682 is replaced by what has been described above. FIG. 26B is a concept view illustrating an example in which an accident occurs in a facility 3600 for reducing radioactive material and a nuclear power plant 360 having the same according to yet still another embodiment of the present invention. When an accident occurs, the fluid inside the boundary unit 3620 is sparged through the fluid path of the discharging unit 3630 to the inside of the cooling water storage unit 3610. As the fluid is sparged from the sparging unit 3640, the steam is condensed and the air is cooled. The air may be discharged through the opening portion 3610b to the inner space of the containment 362. The atmosphere inside the containment 362 is introduced into the heat exchanger 366a by way of natural circulation. The atmosphere in the containment 362 is cooled and condensed by the heat exchanger 366a. The air is discharged back to the inner space of the containment 362, and the condensed water is introduced through the cooling water recollecting portion 3610a to the cooling water storage unit 3610. While passing through the cooling water recollecting portion 3610a, an additive is supplied from the second additive injection unit 3682. As the water level of the cooling water storage unit 3610 gradually increases, the first additive injection unit 3681 is submerged in the cooling water, and the additive is dissolved in the cooling water. The condensed water and the cooling water may be prevented from re-volatilizing by the first additive injection unit 3681 and the second additive injection unit 3682. The above-described facility for reducing radioactive material has been proposed to solve the problems with expanding exclusion area boundary (EAB) that may occur when a passive safety system is introduced. In case an accident occurs in the nuclear power plant (except some limited quantities of leakage), a majority of radioactive materials discharged from the reactor coolant system or pipe line connected with the reactor coolant system is configured to be sparged into a cooling water storage unit such as a large pool or tank through a sparging unit, thus significantly decreasing the concentration of the radioactive material in the containment. Further, release of the radioactive material to the external environment may be minimized. Use of the facility for reducing radioactive material may resolve the issue of expanding EAB that may be caused by adopting the passive safety system in the nuclear power plant and allows for easy introduction of a passive safety system with excellent effects in enhancing safety. A reduction in the EAB may save economical expense, and the facility for reducing radioactive material may maintain the function of reducing radioactive material for a long time as long as the cooling water storage unit maintains a predetermined water level or more, thus contributing to enhanced safety of the nuclear power plant. According to the present invention, when a loss-of-coolant accident occurs, a filter facility may be used to capture the radioactive material in the boundary unit and may suppress a rise in the concentration of the radioactive material in the containment. Further, according to the present invention, the concentration of the radioactive material in the containment is suppressed from increasing to remarkably reduce the EAB, and release of radioactive material to the external environment may be minimized. Accordingly, the nuclear power plant may enjoy significantly enhanced safety, as well as savings in the economical expense. According to the present invention, further, the problem with expanding EAB may be resolved, and a passive safety system with excellent safety enhancing effects may be applied to the nuclear power plant. Further, according to the present invention, the pH of the cooling water in the cooling water storage unit may be controlled by a passive manner to suppress re-volatilization of radioactive material while hardly increasing the number of isolation valves, and even when the radioactive material is re-volatilized, the radioactive material may be suppressed from being discharged to the inside of the containment. As the present features may be embodied in several forms without departing from the characteristics thereof, it should also be understood that the above-described embodiments are not limited by any of the details of the foregoing description, unless otherwise specified, but rather should be construed broadly within its scope as defined in the appended claims, and therefore all changes and modifications that fall within the metes and bounds of the claims, or equivalents of such metes and bounds are therefore intended to be embraced by the appended claims. |
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summary | ||
description | The present invention relates to a tie plate for use in a nuclear fuel assembly and more particularly in an upper nozzle of a fuel assembly for a boiling water reactor (BWR). A nuclear fuel assembly for boiling water reactor (BWR) conventionally comprises a bundle of fuel rods and at least one tubular water channel encased in a tubular fuel channel. Each water channel replaces at least one fuel rod in the bundle. Each fuel rod comprises a tubular cladding receiving a stack of nuclear fuel pellets and closed at its ends by end plugs. The fuel assembly comprises a plurality of spacer grids distributed along the fuel assembly for maintaining the fuel rods transversely in spaced relationship. Each spacer grid comprises guide cells through each of which a fuel rod usually extends. The fuel assembly comprises an upper nozzle at the upper end of the fuel channel and a lower nozzle at the lower end of the fuel channel. Fuel rods extend from the upper nozzle to the lower nozzle. Each nozzle has guide cells each for receiving a respective fuel rod end plug and channels for allowing coolant to flow through the nozzle. In operation, the fuel assembly is oriented vertically in a nuclear reactor core and a coolant is caused to flow upwardly between the fuel rods. The coolant flows in the water channel and the fuel channel, from lower end to upper end of the fuel assembly. Coolant enters the fuel channel through the lower nozzle and exits the fuel channel through the upper nozzle. An object of the invention is to provide an upper nozzle tie plate limiting the pressure drop of the upper nozzle while being of sufficient strength. To this end, a tie plate is provided which is formed by intersecting strips delimiting between them tubular guide cells each for allowing a fuel rod to extend through the tie plate, wherein the strips delimit between them tubular flow cells separate from the guide cells, each flow cell for allowing coolant flow through the tie plate and wherein guide cells and flow cells are arranged at nodes of a lattice defined by a repeating pattern comprising four corner nodes in a square lattice arrangement and a central node at the center of the four corner nodes, with one guide cell at each corner nodes, separated by a pair of parallel spaced strips intersecting a pair of parallel spaced strips, the two pairs of strips delimiting a four-walled central flow cell at the center node. The tie plate may comprise one or several of the following features, taken in isolation or in any technically feasible combination: the central flow cell has a cross-section area superior to that of each guide cell; each guide cell has a square-shaped cross-section; each central flow cell has a square-shaped cross-section; the repeating pattern is a 3×3 array of nodes in a square or rectangular lattice arrangement with the four guide cells at corner nodes, the central flow cell at the center node and four intermediate flow cells at four intermediate side nodes between the corner nodes; each intermediate flow cell has a cross-section area superior to that of the guide cells; each intermediate flow cell has a rectangular-shaped cross-section; the guide cells and flow cells positioned at the nodes of the repeating pattern are delimited between a group of four parallel strips intersecting a group of four parallel strips, wherein in each group the spacing between the pairs of strips delimiting the central flow cell between them is superior to the spacing between the pair of strips delimiting the guide cells between them; it comprises strips having planar guide cell portions delimiting guide cell side walls; it comprises strips having guide cell portions delimiting guide cell side walls curved inwardly with respect to the guide cell; it comprises strips having guide cell portions delimiting guide cell side walls curved outwardly with respect to the guide cell; it comprises strips having guide cell portions delimiting guide cell side walls provided with dimples protruding from the guide cell portion inwardly in the corresponding guide cell. A nuclear fuel upper nozzle is also provided comprising a tie plate as defined above. A nuclear fuel assembly is also provided comprising a tie plate as defined above. A nuclear fuel assembly is also provided comprising an upper nozzle as defined above and a bundle of fuel rods wherein at least one guide cell receives a fuel rod with a transverse clearance between the fuel rod and the guide cell side walls. As illustrated on FIG. 1, the nuclear fuel assembly 2 is elongated along a longitudinal central axis L. In use, the fuel assembly 2 is placed in the core of a nuclear reactor with the axis L extending substantially vertically. In the following, the terms “lower” and “upper”, “axial” and “transverse” refer to the position of the fuel assembly 2 in the reactor. The fuel assembly 2 is for a boiling water reactor (BWR) and comprises a bundle of part-length (not shown) and full-length nuclear fuel rods 4 and a tubular water channel 6 encased in a tubular fuel channel 8. The fuel rods 4, the water channel 6 and the fuel channel 8 extend longitudinally parallel to axis L. Each fuel rod 4 comprises a tubular cladding 10 filled with stacked nuclear fuel pellets and closed at its lower end and upper end by a lower end plug 12 and an upper end plug 14 respectively. The fuel rods 4 are arranged in a lattice and the water channel 6 replaces some of the fuel rods 4 in the lattice. The fuel assembly 2 comprises fuel rod supporting spacer grids 16 distributed along the fuel rods 4, only one spacer grid 16 being illustrated on FIG. 1. The function of the spacer grids 16 is to maintain the fuel rods 4 axially and transversally with a transverse spacing between them. The fuel assembly 2 comprises a lower nozzle 18 and an upper nozzle 20 at the lower end and the upper end respectively of the fuel channel 8. The fuel rods 4 extend from the lower nozzle 18 to the upper nozzle 20. In a boiling water reactor fuel assembly, the fuel assembly bottom nozzle and upper nozzle are configured to aid in channeling the coolant through the fuel assembly during operation, the bottom nozzle accepting coolant flow and the upper nozzle discharging the coolant from the fuel assembly. A BWR lower nozzle typically includes an inlet nozzle, a lower tie plate and a transition region between the inlet nozzle and the lower tie plate whereby coolant entering the inlet nozzle flows upwardly successively through the transition region, through the tie plate and about the individual fuel rods secured or supported freely at their lower ends by the lower tie plate. A debris filter is typically included in the lower nozzle, usually below the lower tie plate, in the transition zone. The lower nozzle 18 is adapted to position the fuel assembly 2 in the lower core plate, to allow coolant to flow through the lower nozzle 18 to enter the fuel channel 8, to catch debris and to support the fuel rods 4. A BWR upper nozzle typically includes a tie plate with holes providing an exit for coolant to flow out of the fuel assembly and a handle bar. One or more of the water and/or fuel rods are usually used as structural members to be rigidly fasten by some means to both the lower and upper tie plates for the purpose of lifting the BWR fuel assembly via the handle bar and maintaining a fixed distance between the lower and upper nozzles. The upper nozzle is usually connected to the fuel channel so that the fuel channel is lifted out of the nuclear reactor core along with the fuel assembly. The upper nozzle and the fuel channel can be disconnected allowing the fuel channel to be lifted off the fuel assembly. The upper end plugs of the fuel rods typically extend through tie plate holes which restrain the end plugs laterally. The upper nozzle 20 is adapted to allow coolant to flow through the tie plate 22 to exit the fuel channel 8 and to allow the upper end plugs 14 of the fuel rods to extend through the tie plate 22. As illustrated on FIG. 2, the upper nozzle 20 comprises a tie plate 22 formed by intersecting strips 24, 26 delimiting between them individual separate tubular guide cells 28 and tubular flow cells 30, 32 centered at nodes of a regular lattice. Each guide cell 28 is for receiving the upper end plug 14 of a respective fuel rod 4 (only one being represented for the sake of clarity). Each flow cell 30, 32 is for allowing coolant to flow through the tie plate 22. The tie plate 22 comprises a central aperture 34 for allowing the water channel 6 to extend through the tie plate 22. The central aperture 34 replaces guide cells 28 and flow cells 30, 32 in the lattice. The tie plate 22 comprises two sets of strips 24, 26. The strips 24, 26 of each set are parallel and mutually spaced. The strips 24, 26 of each set intersect the strips of the other set at right angles thus delimiting the guide cells 28 and the flow cells 30, 32. The tie plate 22 comprises a set of first strips 24 extending in a first transverse direction T1 and a set of second strips 26 extending in a second transverse direction T2 perpendicular to the first transverse direction T1. Each guide cell 28 and each flow cell 30, 32 is delimited by the intersection of two adjacent first strips 24 with two adjacent second strips 26. Each guide cell 28 and each flow cell 30, 32 has four side walls each formed by a respective strip 24, 26. As shown on FIG. 3 illustrating an enlarged partial view of the tie plate of FIG. 2, the guide cells 28 and flow cells 30, 32 are centered at nodes of a regular lattice defined by a repeating pattern P comprising four corner nodes in a square lattice arrangement and a center node at the center of the four corner nodes, with one guide cell at each corner node of the pattern P, separated by a pair of parallel spaced first strips 24 intersecting a pair of parallel spaced second strips 26 and a central flow cell 32 at the center node of the pattern P. The central flow cell 32 is delimited by the intersection of the pair of first strips 24 and the pair of second strips 26. The central flow cell 32 has four side walls. Each side wall is defined by a portion of a respective one of the strips of the two intersecting pairs of first strips 24 and second strips 26. The tie plate 22 further includes intermediate flow cells 30 between successive guide cells 28 in the repeating pattern P. More specifically, the pattern P comprises a 3×3 array of nodes in a square or rectangular lattice arrangement with four guide cells 28 at corner nodes, the central flow cell 32 at the center node and four intermediate flow cells 30 at intermediate side nodes. Each intermediate flow cell 30 is delimited by two successive guide cells 28 in one transverse direction T1, T2 and by two successive central flow cells 32 in the other transverse direction T2, T1. Each central flow cell 32 is surrounded by four guide cells 28 in a centered square pattern P: the four guide cells 28 form a square and the central flow cell 32 is at the center of the square. It is also surrounded by four intermediate flow cells 30 each delimiting one side of the central flow cell 32. Guide cells 28 and flow cells 30, 32 of the pattern P are separated by two first strips 24 and two second strips 26 defining at their intersection at right angle the four-walled central flow cell 30. Each guide cell 28 and each flow cell 30, 32 of the tie plate 22 is delimited by two pairs of intersecting strips 24, 26. Each first strip 24 and each second strip 26 separates on a side a row of guide cells 28 alternating with intermediate flow cells 30 and on the other side a row of intermediate flow cells 30 alternating with central flow cells 32. The pairs of parallel strips 24, 26 delimiting guide cells 28 between them have a small spacing E1 and the pair of parallel strips 24, 26 delimiting between them central flow cells 32 have a large spacing E2 larger that the small spacing E1. Due to the different spacings E1, E2, the guide cells 28 and flow cells 30, 32 have different cross-sections. Guide cells 28 are identical and have a square-shaped cross-section. Central flow cells 32 are identical and have a square-shaped cross-section. The central flow cells 32 have a side dimension superior to that of the guide cells 28. Each central flow cell 32 has a cross-section area superior to that of each guide cell 28. Each intermediate flow cell 30 has two parallel side walls each in common with an adjacent central flow cell 32 and two parallel side walls each in common with an adjacent guide cell 28. The intermediate flow cells 30 have rectangular-shaped cross-section. The short side walls are the side walls shared with the guide cells 28. Each intermediate flow cell 30 has a cross-section area superior to that of the guide cells 28. Referring to FIG. 2, the guide cells 28 and flow cells 30, 32 of the tie plate 22 define in each transverse direction T1, T2 mixed rows 36 each comprising guide cells 28 alternating with intermediate flow cells 30, said mixed rows 36 alternating with flow rows 38 each exclusively comprising flow cells 30, 32, namely intermediate flow cells 30 alternating with central flow cells 32. The tie plate 22 comprises first mixed row 36 and first flow rows 38 extending in the first transverse direction T1 and alternating in the second transverse direction T2 and second mixed rows 42 and second flow rows 40 extending in the second transverse direction T2 and alternating in the first transverse direction T1. Each intermediate flow cell 30 is inserted between two guide cells 28 in a mixed row 36, 42 in one of the first and second transverse directions T1, T2. Each central flow cell 32 is located at the intersection of a first flow row 38 and a second flow row 40. Each strip 24, 26 separates a mixed row 36, 42 and an adjacent flow row 38, 40. Each first strip 24 separates a first mixed row 36 and a first flow row 38 and each second strip 26 separates a second mixed row 42 and a second flow row 40. The tie plate 22 comprises two strips between each pair of adjacent mixed rows in the first direction T1 and in the second direction T2 . So, each pair of adjacent first mixed rows 36 are separated by two parallel spaced first strips 24 and each pair of adjacent second mixed rows 42 are separated by two parallel spaced second strips 26. In each transverse direction T1 , T2 , the spacing between the strips 24, 26 is regular and alternatively large and small: in each transverse direction T1 , T2 the tie plate 22 has the small spacing E1 between each pair of parallel adjacent strips 24, 26 delimiting between them a mixed row 36, 42 and the large spacing E2 between each pair of parallel adjacent strips 24, 26 delimiting between them a flow row 38, 40. The small spacing E1 is smaller than the large spacing E2. FIG. 4 illustrates a front view of a first strip 24 and a second strip 26 of the tie plate 22. First strips 24 and second strips 26 have upper edges 44 and lower edges 46. First strips 24 and second strips 26 are provided with interconnection slots 48, 50 formed in their edges. The first strips 24 are provided with slots 48 in their upper edge 44 and the second strips 26 are provided with slots 50 in their lower edge 46. When the tie plate 22 is assembled, strips 24, 26 are connected by interfitting the slots 48, 50. Slots 48, 50 define along each strip 24, 26 guide cell portions 52 alternating with flow cell portions 54. Each guide cell portion 52 forms a side wall separating a guide cell 28 from an intermediate flow cell 30 and each flow cell portion 54 forms a side wall separating an intermediate flow cell 30 and a central flow cell 32 when the tie plate 22 is assembled. Each guide cell 28 is delimited by four guide cell portions 52, each central flow cell 32 is delimited by four flow cell portions 54 and each intermediate flow cell 30 is delimited by two opposed guide cell portions 52 intersecting perpendicularly two opposed flow cell portions 54. As illustrated on FIG. 1, the fuel rod upper end plugs 14 have a diameter inferior to that of the fuel rods 4. Hence, the tie plate 22 can be provided with guide cells 28 of small cross-section and flow cells 30, 32 of large cross-section. This namely provides central flow cells 32 of large cross-section delimited by two pairs of intersecting strips at the center of each group of four guide cells 28 defining a square and in addition intermediate flow cells 30 between the guide cells 28 around the central flow cell 32. This pattern P allows low pressure drop. The tie plate 22 is formed by mutually spaced intersecting strips. This allows obtaining a tie plate 22 with high strength. It is possible to use strips made of metal with high mechanical characteristic, such as Ni-based alloy, martensitic or precipitation-hardening stainless steel, thus allowing the use of thin strips and reducing even more the pressure drop. To even decrease the pressure loss of the tie plate 22, the lower edges 46 and/or upper edges 44 of the strips 24, 26 may be chamfered or rounded for instance mechanically or via electron beam/laser. The tie plate 22 further allows efficient guiding of the fuel rod end plugs between four side walls with easier manufacturing operations and geometrical controls with respect to providing guide cells and flow cells by drilling holes and flow channels in a massive plate. In addition, owing to the reduced cross-section areas of the flow cells 30, 32 compared to the flow channels drilled in a massive plate, the tie plate 22 allows mitigating the entry of debris and foreign material into the fuel assembly 2 through the upper nozzle 20 especially during non-operating conditions such transport, storage and handling or at the shut-down of the reactor when the coolant stops flowing upwardly. It prevents specifically large debris (more than several mm) to fall into the fuel assembly 2. In the embodiment of FIGS. 2-4, the guide cell portions 52 are planar. Each guide cell 28 is thus delimited by four planar side walls. Preferably, each guide cell 28 has transverse dimensions allowing to receive a fuel rod end plug 14 with a transverse clearance. Alternatively, each guide cell 28 has transverse dimensions to allow a linear contact of at least two opposed side walls with the fuel rod end plug 14. Advantageously, each strip 24, 26 is planar to limit flow resistance. Strips 24, 26 can thus be obtained economically e.g. by cutting a metal sheet while obtaining a tie plate 22 exhibiting low pressure drop and high strength. In the embodiment of FIG. 5 illustrating a partial top view of a tie plate 22, each guide cell portion 52 is curved outwardly with respect to the guide cell 28 delimited by the guide cell portion 52 such that each guide cell 28 exhibits a substantially circular cross-section closely matching the outer surface of a fuel rod upper end plug 14 extending through the guide cell 28. With respect to the embodiment of FIGS. 2-4, the cross-section area of the central flow cells 32 is increased relatively to the cross-section areas of the intermediate flow cells 30 and guide cells 28. In the embodiment of FIG. 6 illustrating a partial top view of a tie plate 22, each guide cell portion 52 is curved inwardly with respect to the guide cell 28 delimited by the guide cell portion 52 such as to be in linear contact with a fuel rod upper end plug 14 extending through the guide cell 28. The cross-sections of the central flow cells 32 are reduced, improving debris and foreign material retaining on the tie plate 22. In the embodiment of FIGS. 7 and 8 illustrating a partial top view of a tie plate 22 and front views of strips 24, 26 of the tie plate 22, guide cell portions 52 of the strips 24, 26 are formed with dimples 56 protruding inwardly with respect to the guide cells 28. Each dimple 56 is formed (e.g. stamped) in the corresponding guide cell portion 52. Each dimple 56 is formed to define contact area of the guide cell portion 52 with a fuel rod upper end plug 14 extending through the corresponding guide cell 28. Each dimple 56 is formed in the corresponding guide cell portion 52 between the lower edge 46 and the upper edge 44 of the strip 24, 26. The dimples 56 allow increased transverse support of the upper end plugs 14 of the fuel rods 4. In the embodiment of FIG. 9 illustrating front views of strips 24, 26 of a tie plate 22, the strips 24, 26 are formed with elastically deformable springs 58 extending from edges of the strips 24, 26, here upper edges 44, in register with the guide cell portions 52. Each spring 58 extends from a guide cell portion 52 and is for contacting a fuel rod end plug extending through a guide cell 28 delimited by guide cell portion 52 and transversely positioning the corresponding fuel rod 4. Each spring 58 comprises an elastically flexible tab 60 extending from a guide cell portion 52 and a rigid contact protrusion 62 formed in the tab 60 for contacting a fuel rod end plug extending through the corresponding guide cell 28 delimited by the guide cell portion 52. This embodiment may be used for instance to secure the upper end plugs 14 of the fuel rods 4 in tie plate 22. Embodiments allowing a linear contact or a contact area of each side wall of a guide cell 28 with the fuel rod en plug 14 improve guiding of the fuel rod end plug 14 and prevent fretting between the end plugs 14 and the tie plate 22. If fuel rods of different lengths are provided in the fuel assembly 2, the part-length fuel rods are usually secured at their lower end by the lower tie plate, while their upper end held in position by the corresponding spacer grids 16. The guide cells 28 positioned above the part-length fuel rods have the geometry of the other guide cells 28 of the tie plate 22 but act as flow cells. The invention applies in particular to an upper tie plate, specifically for a BWR nuclear fuel assembly tie plate. More generally, it applies to nozzles, namely lower nozzles and upper nozzles. |
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description | This disclosure relates to a radiation generator, and, more particularly, to a radiation generator having electrodes with roughly opposite potentials. A neutron generator may include an ion source and a target. An electric field is generated within the neutron generator that accelerates the ions toward the target at a speed sufficient such that, when the ions are stopped by the target, neutrons are generated and emitted into a formation into which the neutron generator is placed. The neutrons interact with atoms in the formation, and those interactions can be detected and analyzed in order to determine information about the formation. While well logging instruments utilizing these neutron generators are useful, they suffer from some unfortunate drawbacks. For example, commonly used ion sources may emit conductive particles that may build up on insulating surfaces inside the neutron generator, thereby changing the characteristics of those insulating surfaces. This in turn may undesirably affect the electric field inside the neutron generator, and therefore alter the focus point of the ion beam, which may result in the ion beam not striking the intended portion of the target. The foregoing serves to degrade the performance of the neutron generator, and thus the performance of the well logging instrument utilizing the neutron generator. Another drawback is that some ions generated by the ion generator may be neutralized by interactions with gases inside the neutron generator. These energetic neutral particles may impinge on a conductive electrode surface, ejecting charged particles such as electrons, and conductive particles such as sputtered metal that could land on an insulator, creating a layer on the insulator which may be charged and may be conductive. As such, further advances in the area of neutron generators are desirable. It is desired for such new neutron generators to reduce the buildup of undesirable charged or conductive particles on insulating surfaces, and thus provide a high degree of stability and consistency, such that they can deliver a tightly focused ion beam to the target and consistently generate neutrons. This summary is provided to introduce a selection of concepts that are further described below in the detailed description. This summary is not intended to identify key features of the claimed subject matter, nor is it intended to be used as an aid in limiting the scope of the claimed subject matter. According to a first aspect, a radiation generator may include an insulator, and a ion source carried within the insulator and to directly generate ions and indirectly generate undesirable particles. An extractor electrode may be carried within the insulator downstream of the ion source and having a first potential. In addition, an intermediate electrode may be carried within the insulator downstream of the extractor electrode at a ground potential and may be shaped to capture the undesirable charged or conductive particles indirectly generated by the ion source. A suppressor electrode may be carried within the insulator downstream of the intermediate electrode and having a second potential opposite in sign to the first potential. A target may be carried within the insulator downstream of the suppressor electrode, and the extractor electrode and the suppressor electrode may have a voltage therebetween such that an electric field generated in the insulator accelerates the ions generated by the ion source toward the target. Another aspect is directed to a well logging instrument. The well logging instrument may include a sonde housing, with a radiation generator carried by the sonde housing. A solid insulator may be carried by the sonde housing between an inner surface of the sonde housing and an outer surface of the radiation generator. There may be an insulating gas in the sonde housing. The radiation generator may include a sealed generator tube, a charged particle source carried within the sealed generator tube and to emit charged particles, an extractor electrode carried within the sealed generator tube downstream of the charged particle source at a first potential, an intermediate electrode carried within the sealed generator tube downstream of the extractor electrode, a suppressor electrode carried within the sealed generator tube downstream of the intermediate electrode at a second potential opposite in sign to the first potential, and a target within the sealed generator tube downstream of the suppressor electrode. The intermediate electrode may be at an intermediate potential between the first and second potential. The difference in the first and second potentials may be such that an electric field generated in the sealed generator tube accelerates the charged particles emitted by the charged particle source toward the target. A method aspect is directed to a method of generating radiation. The method may include generating ions and indirectly generating undesirable particles, the undesirable particles being generated on a trajectory toward an insulator, using an ion source. The method may also include accelerating the ions toward a target within the insulator using an extractor electrode downstream of the ion source at a first potential and a suppressor electrode downstream of the extractor electrode at a second potential opposite in sign to the first potential. The method may further include shielding the insulator from the undesirable particles that would otherwise strike the insulator, using an intermediate electrode downstream of the extractor electrode and upstream of the suppressor electrode at a ground potential. One or more embodiments of the present disclosure will be described below. These described embodiments are only examples of the presently disclosed techniques. Additionally, in an effort to provide a concise description, some features of an actual implementation may not be described in the specification. It should be appreciated that in the development of any such actual implementation, as in any engineering or design project, numerous implementation-specific decisions may be made to achieve the developers' specific goals, such as compliance with system-related and business-related constraints, which may vary from one implementation to another. Moreover, it should be appreciated that such a development effort might be complex and time consuming, but would nevertheless be a routine undertaking of design, fabrication, and manufacture for those of ordinary skill having the benefit of this disclosure. When introducing elements of various embodiments of the present disclosure, the articles “a,” “an,” and “the” are intended to mean that there are one or more of the elements. The terms “comprising,” “including,” and “having” are intended to be inclusive and mean that there may be additional elements other than the listed elements. Additionally, it should be understood that references to “one embodiment” or “an embodiment” of the present disclosure are not intended to be interpreted as excluding the existence of additional embodiments that also incorporate the recited features. Referring initially to FIG. 1, a radiation generator 100 is now described. The radiation generator 100 includes a housing 101 having an interior surface, with an insulator 105 on the interior surface. The housing 101 carries a vacuum envelope formed by the insulator 103 and the various electrodes attached thereto. The insulator 103 may be a high voltage insulator constructed from ceramic material, such as Al2O3. An ionizable gas is contained within the housing, such as deuterium or tritium, at a pressure of 2 mTorr to 20 mTorr for example. An insulating gas, for example SF6, is contained within the housing 101. An ion source 104 is carried within the housing. The ion source 104 includes a cathode 106, a cathode grid 108 downstream of the cathode, and an extractor grid 109 downstream of the cathode grid. During operation of the radiation generator 100, the cathode 104 emits electrons. The cathode 106 and the cathode grid 108 have a voltage therebetween such that the electrons emitted by the cathode are accelerated toward the cathode grid. The cathode grid 108 and the extractor grid 109 have a voltage therebetween less than the voltage between the cathode 106 and cathode grid 108. As the electrons pass the cathode grid 108 on a trajectory toward the extractor grid 109, they slow down due to the lesser voltage between the cathode grid and extractor grid. Some electrons then strike the atoms of the ionizable gas, resulting in ionization. Although the structure of this ion source 104 has been described herein, those of skill in the art will readily appreciate that other types of ion sources, such as those that operate at a lower temperature and based upon a Penning discharge, may be used. Indeed, the disclosure herein is applicable to any sort of radiation generator, regardless of cathode type. The radiation generator 100 also includes an extractor electrode 110 carried within the housing downstream of the ion source 104 that, during operating, is at a first potential. The extractor electrode 110 is curved inwardly toward a longitudinal axis of the insulator, which provides advantages that will be discussed below. An intermediate electrode 112 is carried within the housing downstream of the extractor electrode 110. A suppressor electrode 118 is carried within the housing downstream of the intermediate electrode 112 and, during operation, is at a second potential. The suppressor electrode 118 is curved inwardly toward a longitudinal axis of the insulator 103, which also provides advantages that will be discussed below. During operation, the intermediate electrode is at a potential between that of the extractor and the suppressor. The intermediate electrode may be substantially at ground potential while the suppressor and extractor are at potentials with opposite signs but not necessarily of equal magnitude. This may be achieved by having a first power source (not shown) coupled to the extractor electrode 110 to drive it to the first potential, and a second power source (not shown) coupled to the suppressor electrode 118 to drive it to the second potential. Those skilled in the art will appreciate that there may be other extractor electrodes downstream of the extractor electrode 110 shown, and that there may be other suppressor electrodes downstream of the suppressor electrode 118. There may be a first voltage divider circuit (not shown) coupled to the first power source and to each extractor electrode 110 so as to provide an increasing absolute voltage difference between the extractor 110 and each successive extractor electrode. In addition, there may be a second voltage divider circuit (not shown) coupled to each suppressor electrode 118 so as to provide an increasing absolute voltage difference between the intermediate electrode 112 and each successive suppressor electrode. A target 120 is carried within the housing downstream of the suppressor electrode 118. There is a voltage difference between the extractor electrode 110 and the suppressor electrode 118 such that an electric field generated in the housing accelerates the ions emitted by the ion source 104 toward the target 120. When the ions strike the target 120, neutrons or gamma rays, depending upon the selection of the target material, are generated. The neutrons or gamma rays can be emitted into a material, such as a formation in a borehole. The neutrons react with nuclei in the formation, and can be either reflected back, or can cause photons such as gamma ray photons to be reflected back. These reflected neutrons or gamma ray photons can be captured by a detector (not shown). Monitoring of the detector, together with analysis of the data collected thereby, can then be used to determine properties of the material in the formation. It should be noted that there is a negative difference in potential between the suppressor electrode 118 and the target 120 such that secondary electrons formed when the ions strike the target or gas between the suppressor electrode and target are directed back toward the target instead of toward the ion source 104. If the electrons were allowed to fly back toward the ion source 104, they could strike the cathode 106, heating the surface thereof and potentially generating unwanted x-rays which could damage the insulators 103 or 105. The electrons could also strike the insulator 103 and charge it up, causing asymmetrical potential distribution. A limiting factor in prior radiation generator 100 designs is the length of the acceleration gap between the extractor electrode 110 and the suppressor electrode 118. The pressure of the ionizable gas in the housing causes a variety of undesirable reactions between the accelerated ions and the ionizable gas itself, and the longer the acceleration gap, the greater the chance of these undesirable reactions. These reactions can include the formation of neutral, accelerated particles that can impinge metal surfaces inside the accelerator and the resulting creation of undesirable charged or conductive particles via sputtering, which can strike the insulator 103 and build up thereon. If enough undesirable charged or conductive particles build upon the insulator, portions of the surface of the insulator 103 may become charged and/or conductive. This would serve to alter the potential distribution between the extractor electrode 110 and suppressor electrode 118, as well as other components. This could alter the electric field in the housing, and thus alter the path or cohesiveness of the ion beam, which would degrade performance of the radiation generator 100. Worse, with enough undesirable conductive particles building up the insulator 103, a short could form between the extractor electrode 110 and suppressor electrode 118, or between other components, for example. Such a short could result in damage to the radiation generator 100 rendering it inoperable. Another concern is the creation of undesirable neutral particles. These undesirable neutral particles are formed when ions strike or interact with molecules of the ionizable gas in the acceleration gap. In this situation, an electron from the ionizable gas jumps to the ion, turning the ion into a neutral particle. The energy and direction of the newly formed neutral particle remains, yet because the particle is neutral, the electric field in the housing does not influence its trajectory. If this particle strikes a metallic surface in the radiation generator 100 it may sputter material therefrom as well as cause secondary electron emission. The material sputtered would be in the form of undesirable conductive particles, the undesirable properties of which have been described above. As also explained above, the secondary electrons could strike the insulator 103 and charge it up, or could strike a metallic surface and cause the generation of x-rays, which could in turn damage the high voltage insulator 105 between the generator 100 and the grounded housing 101. Also, secondary electron emission can lead to erroneous current flow, which could overload the power supplies. Yet another reason why it is desirable for the acceleration gap to be kept as small as possible is to reduce the likelihood of a charge exchange reaction between an initially accelerated ion and an atom of ionizable gas. In the charge exchange reaction, the initially positively charged ion picks up an electron from an atom of ionizable gas, creating a neutral particle (the negatives of which are explained above), as well as creating an ion from the ionizable gas atom. This new ion is an undesirable ion, as it is accelerated by but part of the available potential difference. The undesirable ion may or may not strike the target 120. If it strikes the target 120, its diminished energy makes it more likely to cause target erosion through sputtering and much less likely to cause a neutron generating reaction. It is therefore desirable to keep charge exchange to a minimum by using an acceleration gap of minimal length as charge exchange is more likely at low ion energies. Those of skill of art will appreciate that since the ion source 104 generates the ions which ultimately generate the undesirable conductive or undesirable neutral particles, which in turn can cause the secondary electron emission, the ion source can be said to indirectly generate the undesirable particles in the radiation generator 100. By having the extractor electrode 110 and the suppressor electrode 118 at potentials opposite in sign and with a well-defined potential distribution due to the presence of the intermediated electrode(s), the acceleration gap therebetween can be shortened. By shortening the acceleration gap, the number of charge exchange reactions can be reduced. This reduces the number of particles hitting the electrodes and therefore the amount of secondary electron emission. Since the extractor electrode 110 and suppressor electrode 118 are at potentials opposite in sign with respect to the intermediate electrode, the largest potential difference between separate electrodes is reduced compared to conventional radiation generators where the insulating material 103 is to hold off the full potential difference, while the potential difference between the extractor electrode and suppressor electrode can remain the same. Further, if the intermediate electrode is substantially at ground potential the largest potential difference between the electrodes and the grounded housing, and thus the electric field therebetween is reduced (by a factor of two, in some applications), allows the thickness of the insulation (not shown) surrounding the generator tube 100 to be reduced, as with the lesser electric field comes a lesser chance of arcing and other undesirable effects. Although the shortened acceleration gap helps reduce these undesirable effects, it may not completely do so. Therefore, to help mitigate performance degradation caused by the undesirable conductive particles and secondary electron emission, the intermediate electrode 112 is shaped to capture the undesirable charged or conductive particles that would otherwise strike the insulator 103. Indeed, the intermediate electrode 112 is T-shaped, comprising a base 114 extending along the longitudinal axis of the insulator 103, and a projection 116 extending outwardly from the base. The projection 116 illustratively has a concave triangular shape. Since the shape of the intermediate electrode 112 captures the undesirable conductive or neutral particles, as well as charged particles, that would otherwise strike the insulator 103, and forces such particles to ground, the electric field in the housing remains unchanged. In addition, the suppressor electrode 118 can be shaped such that secondary electrons formed on the downstream surface thereof are forced toward the intermediate electrode 112 where they can be forced to ground. This may result in the creation of x-rays, albeit at a lesser energy level than if the x-rays had been created by the secondary electrons striking the extractor electrode 110, because the potential difference between the suppressor electrode 118 and the intermediate electrode 112 is about half the potential difference between the extractor electrode 110 and suppressor electrode 118. Thus, although these x-rays are created, they are less damaging than if they had been formed by the secondary electrons instead striking the extractor electrode 110. Also, in some applications, the intermediate electrode 112 can be shaped such that the secondary electrons formed on the upstream surface thereof are forced toward the extractor electrode 110, resulting in the creation of x-rays lesser in energy than x-rays that would be created by secondary electrons created on the surface of the suppressor 118 electrode striking the extractor electrode 110 in the absence of the intermediate electrode. In addition, a portion of the x-rays generated may be absorbed by the intermediate electrode 112 before they damage the insulators 103 or 105. Thus, the intermediate electrode 112 shields the insulator 103 not only from x-rays but also undesirable charged or conductive particles. It should be appreciated that since the x-rays result from the undesirable charged or conductive particles striking electrodes, the x-ray photons themselves can be considered to be undesirable particles indirectly generated by the ion source. Furthermore, the extractor electrode 110, intermediate electrode 112, and suppressor electrode 118 can be shaped so as to capture the undesirable charged or conductive particles that would otherwise strike the insulator. In addition, the intermediate electrode can be made of or coated with a low-Z material, such as beryllium, to reduce the creation of x-rays produced by secondary electrons striking the electrode. FIG. 2 illustrates lines of constant potential in and around the acceleration gap and the trajectories of secondary electrons in and around the acceleration gap. Here, secondary electron emission from the upstream surface of the suppressor electrode 218 and the upstream concave surface of the intermediate electrode 212 is shown. In the case of the suppressor electrode 218, the secondary electrons are generated and leave the surface due to neutral particles striking that surface. As shown, these electrons are then captured by the intermediate electrode 212 and do not fly upstream toward the ion source. In the case of the secondary electrons being generated on the surface of the intermediate electrode 212, also due to neutral particles striking that surface, the secondary electrons, as shown, strike the extractor electrode 210. As explained above, due to the fact that these secondary electrons are accelerated at less than the full potential difference between the extractor electrode 210 and suppressor electrode 218 due to the presence of the intermediate electrode 212, the damage from the resulting x-rays is lessened. While the disclosure has been described with respect to a limited number of embodiments, those skilled in the art, having benefit of this disclosure, will appreciate that other embodiments can be envisioned that do not depart from the scope of the disclosure as disclosed herein. |
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abstract | Laser protection arrangement with a safety cutoff comprising a passive laser protection wall which stores the radiation energy of impinging radiation of a laser of a laser material processing installation, a laser protection foil which causes a detectable change when struck by laser radiation and which is arranged in front in direction of the laser radiation, and at least one sensor which is connected to the laser by a threshold switch in order to switch off the laser when the received detector signal exceeds or falls below a threshold value. |
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047059494 | abstract | Disclosed is an improved specimen cell for maintaining a scanning electron microscope specimen under nearly physiological conditions during observation when said specimen includes liquids having a relatively high vapor pressure. A cavity in the specimen cell mounts an open or closed specimen module which is scanned by the electron beam through a small aperture. During preparation of the electron microscope for observation, the aperture is closed by a door so as to prevent evaporation of liquids from the specimen. The door is mechanically or electronically opened to facilitate observation thus minimizing the exposure of the specimen to the desiccation and/or destructive vacuum effects. Furthermore, the aperture is sized so as to provide a resistance to vapor flow through the aperture while permitting bidirectional electron flow facilitating the electron microscopic observation of the specimen. In a further embodiment, an open specimen module is provided which can be replenished with liquid or fluid, which can be vibrated, raised and/or lowered and which can be heated or cooled as desired during observation of the specimen. |
042279682 | description | SPECIFIC DESCRIPTION FIGS. 1 and 2 show a prestressed substantially cylindrical cast metal pressure vessel 1, hereinafter referred to as the reactor vessel, which has a central chamber 2 receiving a helium-cooled high-temperature nuclear reactor using ball-shaped nuclear fuel elements (not shown). Such reactors are conventional in the art. The reactor vessel 1 has a copper plate 1a, a bottom plate 1b and a cylindrical wall structure 1c which is traversed by vertical stressing cables 3. The stressing means for the horizontal stressing of the individual receptacles is not shown in these Figures. It should be understood that the pressure vessels of the present invention can be composed of cast iron or cast steel blocks which are assembled together while the cover 1a and the plate 1b may likewise be composed of segments or sectors held in interfitting and tight relationship. Such structures are known in the art. Referring now to FIG. 1A, it will be seen that a typical pressure vessel, whether used as the central pressure vessel or one of the satellite vessels, as represented at 101 in FIG. 1A, can have a cover 101a which rests upon the wall structure 101c and is clamped by nuts 103a thereagainst, the nut being threaded onto the stressing cable 103. Corresponding nuts 103b engage the opposite end of the stressing cable 103. The cable 103 passes through a bore 101d in the wall structure 101c and through corresponding registering bores in the cover 101a and the base 101b. The prestressing cables 103c which extend around the vessel in horizontal planes provide the inward and individual prestress previously mentioned. In a partial circle around the reactor vessel 1 there are provided a plurality of angularly equispaced component vessels 4 which are likewise formed from prestressed cast material as pressure vessels and which are likewise substantially cylindrical (see the foregoing discussion as to FIG. 1A). Here the vertical prestressing is applied by tension cables 5 which pass through the cover and the bottom plates of the component vessel 4. Horizontal prestress is effected via cables 103c in the manner already described and as shown, for example, in FIG. 1B. Each component vessel 4 communicates with the reactor vessel 1 via a horizontal gas passage 6 which can be provided with a pair of coaxial conduits one of which delivers gas to the component vessel while the other conducts the gas back into the reactor vessel. Preferably, the hot gas from the high-temperature reactor traverses the inner of the coaxial conduits while the returning gas, relatively cool, traverses the outer conduit of the coaxial conduit system. All of the horizontal gas passages 6 lie in a common horizontal plane and extend radially with respect to the vertical axis of the assembly. The horizontal plane of the passages 6 lies in the lower portion of the cylindrical wall 1c of the reactor vessel 1. In the region of the horizontal gas passages 6, the reactor vessel 1 is provided with a plurality of vertical planar surfaces 7 which together form a polygonal cross section for the reactor vessel and lie generally along tangents to circles centered on the axis of the reactor vessel. Thus in the region of the plane mentioned previously, the reactor vessel has a polygonal cross section. The vertical planar surfaces 7 are so arranged that some of the surfaces 7a lie at right angles and are pierced by the horizontal radial gas passages 6. The component vessels 4 in the same region are also provided with vertical planar surfaces, i.e. at least three and preferably five such surfaces angularly adjoining one another. One of the planar surfaces of each of these component vessels, shown at 8a, lies directly against the corresponding surface 7a and is coextensive therewith, being also perpendicular to the gas passage 6 which communicates between this component vessel and the reactor vessel 1. The gaps 9 between the individual component vessels 4 are filled only in the region of the vertical planar surfaces 7 and 8 into a complete circular disk with support blocks 10 of grey cast iron. The support blocks 10 are as shown hollow and hence constituted by cast webs. The support blocks 10 abut the vertical planar surfaces 7b of the reactor vessel 1 between the surfaces 7a and against the lateral planar surfaces 8b of the component vessels 9. As a result, in the region of the horizontal plane mentioned previously, the pressure vessels 1 and 4 form together with the support blocks 10 a disk-shaped compound body 11. A horizontal prestressing member 12 as shown generally in FIG. 1 and FIG. 2 in which it is constituted as a band but as cables 11a and 11b in FIG. 1B, prestresses the composite disk inwardly. The manner in which the peripheral prestressing member 11a, 11b or 12 passes around the disk can be seen from FIG. 2. The vertical prestressing of the individual vessels by means of the tension cables 3 and 5 or 103 is thus not effected in any way by the support blocks 10. The component vessels 4 preferably receive the steam-generating heat exchangers and any waste-heat recovery system in the manner described. FIG. 3 shows an embodiment of the invention in which the hollow support blocks 10 do not form a single composite body but rather form two composite bodies represented at 13 and 14 respectively above and below the gas passage 6. Each of the composite bodies 13 and 14 can be stressed inwardly by a respective band 12 or set of cables 11a, 11b. |
abstract | A treatment couch having a tabletop that has a radiolucent region, and a lower-torso assembly that has a first slide member coupled to a slot of the tabletop, a second slide member coupled to a slot of the first slide member, and a link member coupled to the first and second slide members. The lower-torso assembly may be configured to adjust a patient along a longitudinal direction relative to a shoulder line of the tabletop while the lower-torso assembly remains outside the radio lucent region of the tabletop. |
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abstract | A passively cooled storage module for spent nuclear fuel includes an elongated body including a top end, bottom end, sidewall, baseplate, detachable lid, and cavity for holding a fuel canister containing heat-emitting spent nuclear fuel assemblies. Cooling air inlet ducts spaced draw ambient cooling air radially inwards into a lower portion of the cavity. The air flows upwards in the cavity along the canister and is discharged from the top end of the module to atmosphere via natural circulation. The air inlet ducts may have a multi-angled and recurving configuration comprising one or more obliquely angled sections in one embodiment. The exterior inlet end openings of the inlet ducts are arranged at a higher elevation than the interior outlet end openings to prevent the ingress of standing and flood-related waters. The ducts and lid include radiation shielding features. |
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06207962& | summary | FIELD OF THE INVENTION The present invention pertains, inter alia, to charged-particle-beam (CPB) microlithography apparatus and methods as used for transferring a pattern, defined on a reticle, to a sensitized substrate. Such apparatus and methods have especial utility in the manufacture of integrated circuits, displays, and the like. The invention also pertains to methods and apparatus for calibrating and adjusting a CPB projection-optical system and for aligning the substrate and reticle with each other for accurate pattern transfer. The invention also pertains to methods and apparatus for reducing thermal deformation of a member, such as the reticle or a movable stage, that defines an alignment or calibration mark. As used herein, the term "reticle" pertains not only to reticles and masks that define an actual pattern to be transferred to a substrate, but also to aperture plates and the like as used in, for example: variable-shaped-beam projection-exposure systems, character projection systems, and "divided" projection-exposure systems. In "divided" projection-exposure systems, the reticle is divided or segmented into multiple "exposure units" (e.g., subfields, stripes, or other subdivisions) that are exposed individually and sequentially onto the substrate on which the images of individual exposure units are "stitched" together contiguously to form the complete pattern on the substrate. BACKGROUND OF THE INVENTION Various methods and apparatus are under current research and development for transferring, using a charged particle beam, a pattern defined by a reticle or mask onto a sensitized substrate by microlithography. Representative charged particle beams used in such systems include electron beams and ion beams. Electron-beam systems have been the subject of most such effort; hence, the following summary is in the context of electron-beam systems. Charged-particle-beam (CPB) microlithography systems, such as electron-beam writing systems, offer tantalizing prospects of improved accuracy and resolution of pattern transfer, but exhibit disappointingly low throughput. Consequently, much contemporary research and development has focused on overcoming this disadvantage. Examples of various conventional approaches include "cell-projection," "character projection," and "block projection" (collectively termed "partial-block" pattern transfer). Partial-block pattern transfer is used especially whenever the pattern to be transferred to the substrate comprises a region in which a basic pattern unit is repeated many times. For example, partial-block pattern transfer is used generally for patterns having large memory circuits, such as DRAMs. In such patterns, the basic pattern unit is very small, having measurements on the substrate of, for example, (10 .mu.m).sup.2 (i.e., 10 .mu.m.times.10 .mu.m). The basic pattern unit is defined on one or several exposure units on the reticle and the exposure units are repeatedly exposed many times onto the substrate to form the pattern on the substrate. Unfortunately, partial-block pattern transfer tends to be employed only for repeated portions of the pattern. Portions of the pattern that are not repeated are transferred onto the substrate using a different method, such as the variable-shaped-beam method. Therefore, partial-block pattern-transfer has a throughput that is too low, especially for large-scale production of integrated circuits. A conventional approach that has been investigated in an effort to achieve a higher throughput than partial-block pattern-transfer methods is a projection microlithography method in which the entire reticle pattern for a complete die (or even multiple dies) is projection-exposed onto the substrate in a single "shot." In such a method, the reticle defines a complete pattern, such as for a particular layer in an entire integrated circuit. The image of the reticle pattern as formed on the substrate is "demagnified" by which is meant that the image is smaller than the pattern on the reticle by a "demagnification factor" (typically 4:1 or 5:1). To form the image on the substrate, a projection lens is situated between the reticle and the substrate. Whereas this approach offers prospects of excellent throughput, it to date has exhibited excessive aberrations and the like, especially of peripheral regions of the projected pattern. In addition, it is extremely difficult to manufacture a reticle defining an entire pattern that can be exposed in one shot. Yet another approach that is receiving much current attention is the "divided" or "partitioned" projection-exposure approach that utilizes a "divided," "partitioned," or "segmented" reticle. On such a reticle, the overall reticle pattern is subdivided into portions termed herein "exposure units." The exposure units can be of any of various types termed "subfields," "stripes," etc., as known in the art. Each exposure unit is exposed individually and sequentially in a separate "shot" or scan. The image of each exposure unit is projection-exposed (typically at a demagnification ratio such as 4:1 or 5:1) using a projection-optical system situated between the reticle and the substrate. Even though the projection-optical system typically has a large optical field, distortions, focal-point errors and other aberrations, and other faults in projected images of the exposure units are generally well controlled. Although divided projection-exposure systems provide lower throughput than systems that expose the entire reticle in one shot, divided projection-exposure systems exhibit better exposure accuracy and image resolution. In divided projection exposure, it is necessary to achieve very accurate alignment of the reticle with the substrate to ensure that the images of the exposure units are positioned at the respective locations on the reticle with extremely high accuracy. To such end, an operation termed "mark detection" is performed such as during calibration of the optical system and when aligning the substrate with the reticle before exposing an exposure unit onto the substrate. During mark detection, an image of one or more "upstream" marks provided on the reticle or other location on the reticle stage is projected onto a corresponding "downstream" mark provided on the substrate or other location on the substrate stage. The marks are scanned relative to each other to determine the relative positions of the marks. Systems designed for high-resolution pattern transfer, such as the divided projection-exposure system summarized above, employ very large acceleration voltages such as between the CPB source and the reticle. To achieve the requisite high accuracy of mark detection, either mark scanning must be performed relatively slowly or a large number of scans must be performed. Consequently, the cumulative beam energy that strikes the marks and their immediate surrounding area is very high. This energy usually is dissipated as localized heating which elevates the temperature and causes thermal deformation of the vicinity of the marks. Such deformation degrades the accuracy with which mark positions can be determined, reduces calibration and alignment accuracy, and reduces the accuracy with which images of exposure units on the substrate can be stitched together. The resulting devices manufactured under such conditions exhibit a higher incidence of defects such as shorts, opens, and non-uniform resistance values. SUMMARY OF THE INVENTION The present invention solves certain of the problems of conventional apparatus and methods summarized above and thereby provides more accurate transfer of a reticle pattern to a substrate. According to a first aspect of the invention, charged-particle-beam (CPB) microlithography (projection-exposure or projection-transfer) apparatus are provided. According to a representative embodiment, such an apparatus comprises an illumination optical system situated and configured to direct a charged-particle illumination beam along an optical axis from a source to a selected region on a reticle. The reticle is situated at a reticle plane orthogonal to the optical axis. The apparatus also comprises a projection-optical system situated and configured to direct a charged-particle imaging beam from the reticle to a sensitized substrate so as to transfer the pattern portion defined by the selected exposure unit to the substrate. An "upstream" mark is situated on the reticle plane so as to be irradiated selectively by the illumination beam. A shield is situated between the source and the upstream mark. The shield defines an aperture that transmits a portion of the illumination beam to the upstream mark while blocking other portions of the illumination beam from reaching the reticle plane. In the embodiment summarized above, the upstream mark can be situated on the reticle. In such an instance, the reticle can comprise multiple upstream marks distributed over the reticle. In such a configuration, the shield desirably defines multiple apertures each corresponding to a respective individual upstream mark on the reticle. Alternatively, the upstream mark can be situated on a mark member separate from the reticle. In such a configuration, the shield desirably extends over the mark member. This configuration usually is used for calibration of the optics of the CPB projection-exposure apparatus. The upstream mark can comprise multiple mark portions. In such an instance, the aperture defined by the shield can be sized, whenever the aperture is axially registered with the upstream mark, to circumscribe all the mark portions collectively. Alternatively, the shield can define multiple apertures each corresponding to a respective individual mark portion. According to another aspect of the invention, CPB microlithography methods are provided in which a charged-particle illumination beam is used to irradiate a portion of a pattern defined by a reticle situated on a reticle plane. A projection-optical system is used to direct a corresponding charged-particle imaging beam from the irradiated portion to a sensitized substrate situated on a substrate plane. An upstream mark is defined on the reticle plane and a "downstream" mark is defined on the substrate plane. The upstream mark is registrable selectively with the downstream mark to perform beam alignment. A shield is provided upstream of the upstream mark. The shield (a) serves to block downstream passage of the illumination beam, and (b) defines an aperture having a size and profile sufficient to pass therethrough only a portion of the illumination beam sufficient to irradiate the upstream mark. When irradiating the upstream mark with the illumination beam, the illumination beam is passed through the aperture of the shield before the illumination beam reaches the upstream mark. The upstream mark can be defined on the reticle, in which instance the shield desirably extends over the reticle. Alternatively, the upstream mark can be defined on a mark member (which can be separate from the reticle), in which instance the shield desirably extends over the mark member. In conventional CPB projection-exposure systems having utility for, e.g., performing "divided" projection exposure, the illumination beam as incident on the reticle can have a transverse profile that is relatively large (e.g., (100 .mu.m).sup.2 -(1000 .mu.m).sup.2). A typical upstream mark is much smaller, on the order of a few .mu.m square to about a hundred .mu.m square. Whenever such upstream marks are illuminated by the charged particle beam during calibration or alignment, the beam that strikes the upstream mark is much larger in transverse area than required for illuminating the upstream mark. As summarized above, the resulting large amount of energy being dissipated in an area surrounding the upstream mark can cause thermal deformation of the upstream marks. Whereas it might be possible to reduce the transverse area of the beam, such a method is impractical because it requires a very complex irradiation optical system. Apparatus and methods according to the invention, as summarized above, reduce the transverse area of the illumination beam actually irradiating an upstream mark, thereby largely eliminating thermal deformation of the mark(s). The foregoing and additional features and advantages of the invention will be more readily apparent from the following detailed description, which proceeds with reference to the accompanying drawings. |
claims | 1. A method for operating a pressurized water nuclear reactor comprising a core containing nuclear fuel assemblies comprising nuclear fuel rods, the method comprising:operating the nuclear reactor for successive cycles with, between each cycle, steps for replacing spent nuclear fuel assemblies with fresh nuclear fuel assemblies, including:operating the nuclear reactor for an initial cycle for starting the nuclear reactor, during which the core contains initial nuclear fuel assemblies, at least some of the initial nuclear fuel assemblies only comprising nuclear fuel rods, before irradiation, containing uranium oxide and not containing any plutonium oxide; thenoperating the nuclear reactor for transition cycles, at least some of the initial nuclear fuel assemblies being progressively replaced during the replacement steps preceding the transition cycles, with transition nuclear fuel assemblies or with plutonium-equilibrium nuclear fuel assemblies, the plutonium-equilibrium nuclear fuel assemblies only comprising, before irradiation, nuclear fuel rods exclusively based on uranium and plutonium mixed oxide, the nuclear fuel rods of each plutonium-equilibrium nuclear fuel assembly having a same nuclear fuel isotope composition and a same nominal total plutonium mass content; and thenoperating the nuclear reactor for at least one plutonium-equilibrium cycle during which the core only contains plutonium-equilibrium nuclear fuel assemblies. 2. The method as recited in claim 1 wherein at least some of the transition nuclear fuel assemblies comprise poisoned nuclear fuel rods, the poisoned nuclear fuel rods containing, before irradiation, at least one consumable neutron poison. 3. The method as recited in claim 2 wherein at least some of the poisoned nuclear fuel rods do not contain, before irradiation, any plutonium. 4. The method as recited in claim 1 wherein the nuclear fuel rods of at least some of the initial nuclear fuel assemblies have, before irradiation, nominal plutonium fissile isotope contents less than those of the nuclear fuel rods of the plutonium-equilibrium nuclear fuel assemblies. 5. The method as recited in claim 1 wherein at least some of the transition nuclear fuel assemblies comprise nuclear fuel rods, before irradiation, containing uranium oxide and not containing any plutonium oxide. 6. The method as recited in claim 1 wherein, at least during the replacement step preceding at least one of the transition cycles or the at least one plutonium-equilibrium cycle, transition nuclear fuel assemblies are replaced with plutonium-equilibrium nuclear fuel assemblies. 7. The method as recited in claim 1 wherein at least some initial nuclear fuel assemblies are zoned nuclear fuel assemblies, the nuclear fuel rods of the zoned nuclear fuel assemblies containing, before irradiation, uranium and plutonium mixed oxide, the zoned nuclear fuel assemblies each comprising several zones in which the nuclear fuel rods have, before irradiation, different nominal plutonium isotope contents. 8. The method as recited in claim 7 wherein at least some of the zoned nuclear fuel assemblies each comprise:a first central zone comprising nuclear fuel rods having a first nominal plutonium fissile isotope content; anda second zone extending along outer faces of the zoned nuclear fuel assembly and consisting of nuclear fuel rods having a second nominal plutonium fissile isotope content strictly less than the first nominal plutonium fissile isotope content. 9. The method as recited in claim 8 wherein at least some of the zoned nuclear fuel assemblies each further comprise:a third zone positioned in corners of the zoned nuclear fuel assembly and consisting of nuclear fuel rods having a third nominal plutonium fissile isotope content strictly less than the second nominal plutonium fissile isotope content. 10. The method as recited in claim 7 wherein, during the initial cycle for starting the nuclear reactor, the zoned nuclear fuel assemblies are located in an outer peripheral layer of the core. 11. The method as recited in claim 10 wherein at least some of the zoned nuclear fuel assemblies each comprise, during the replacement step preceding a first transition cycle, transition nuclear fuel assemblies loaded into the layer immediately adjacent to the outer peripheral layer of the core. 12. The method as recited in claim 11 wherein, during the replacement step preceding the first transition cycle, the zoned nuclear fuel assemblies are displaced immediately inside the layer located immediately inside the outer peripheral layer of the core. 13. The method as recited in claim 10 wherein transition nuclear fuel assemblies loaded during the replacement step preceding a first transition cycle comprise nuclear fuel rods, before irradiation, containing uranium oxide and not containing any plutonium oxide. 14. The method as recited in claim 13 wherein, during the replacement step preceding the first transition cycle, the zoned nuclear fuel assemblies are displaced immediately inside the layer located immediately inside the outer peripheral layer of the core. 15. The method as recited in claim 10 wherein, during the replacement steps preceding the first transition cycle, a second transition cycle, a third transition cycle and the equilibrium cycle, plutonium-equilibrium nuclear fuel assemblies are loaded. 16. The method as recited in claim 1 wherein, in the initial cycle for starting the nuclear reactor, all the nuclear fuel rods of the initial nuclear fuel assemblies, before irradiation, contain uranium oxide but do not contain any plutonium oxide. 17. The method as recited in claim 1 wherein in the plutonium-equilibrium cycle, the nuclear fuel rods of all the plutonium-equilibrium nuclear fuel assemblies have a same nuclear fuel isotope composition and a same nominal total plutonium mass content. |
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claims | 1. A method for assembling a secondary collimator including a first face plate having a first surface and an opposing second surface, said method comprising:positioning a lamella assembly on the first face plate, wherein the lamella assembly includes at least one radiation-absorbing material layer and at least one radiation-transmitting material layer such that a first surface of the lamella assembly is adjacent the second surface of the first face plate; andcoupling a second face plate to the first face plate and the lamella assembly such that a first surface of the second face plate is adjacent a second surface of the lamella assembly. 2. A method in accordance with claim 1, further comprising fabricating at least the second surface of the first face plate within about 10 μm of a predetermined profile. 3. A method in accordance with claim 1, wherein positioning a lamella assembly on the first face plate further comprises positioning the lamella assembly on a substantially planar first face plate. 4. A method in accordance with claim 1, wherein positioning a lamella assembly on the first face plate further comprises positioning the lamella assembly on a first face plate having an at least partially non-planar profile. 5. A method in accordance with claim 1, wherein positioning a lamella assembly on the first face plate further comprises positioning the lamella assembly on the first face plate such that the first surface and the second surface of the lamella assembly are substantially parallel to the second surface of the first face plate. 6. A method in accordance with claim 1, wherein positioning a lamella assembly on the first face plate further comprises positioning a first lamella on the second surface of the first face plate and positioning a second lamella on the first lamella, wherein each lamella includes at least one radiation-absorbing material layer and at least one radiation-transmitting material layer. 7. A method in accordance with claim 1, wherein coupling a second face plate to the first face plate and the lamella assembly further comprises coupling the second face plate to the first face plate and the lamella assembly using a mechanical fastening mechanism. 8. A secondary collimator, comprising:a first face plate;a second face plate; anda lamella assembly coupled between said first face plate and said second face plate, said lamella assembly comprising at least one lamella, each said lamella comprising at least one radiation-absorbing material layer and at least one radiation-transmitting material layer. 9. A secondary collimator in accordance with claim 8, wherein said lamella assembly comprises a plurality of lamellae, wherein said at least one radiation-absorbing material layer comprises a porous material and said at least one radiation-transmitting material layer comprises a metal layer. 10. A secondary collimator in accordance with claim 8, wherein said first face plate comprises a first surface and a second surface and said lamella assembly comprises a first surface and a second surface, said first surface of said lamella assembly substantially parallel to at least said second surface of said first face plate. 11. A secondary collimator in accordance with claim 8, wherein said first face plate is substantially planar. 12. A secondary collimator in accordance with claim 8, wherein said first face plate is at least partially non-planar. 13. A secondary collimator in accordance with claim 8, wherein said lamella assembly further comprises a plurality of substantially parallel aluminum composite panels. 14. An X-ray diffraction imaging (XDI) system, comprising:an X-ray source;a detector array comprising a plurality of detector elements; anda secondary collimator coupled between said X-ray source and said detector array, said secondary collimator comprising:a first face plate;a second face plate; anda lamella assembly coupled between said first face plate and said second face plate, said lamella assembly comprising at least one lamella, each said lamella comprising at least one radiation-absorbing material layer and at least one radiation-transmitting material layer. 15. An XDI system in accordance with claim 14, further comprising an examination area defined between said X-ray source and said secondary collimator. 16. An XDI system in accordance with claim 14, wherein said X-ray source comprises a multi-focus X-ray source. 17. An XDI system in accordance with claim 14, wherein said lamella assembly further comprises a plurality of lamellae, wherein said at least one radiation-absorbing material layer comprises a porous material and said at least one radiation-transmitting material layer comprises a metal layer. 18. An XDI system in accordance with claim 17, wherein a number of lamellae is equal to a number of channels of said detector array. 19. An XDI system in accordance with claim 14, wherein said secondary collimator has a substantially planar profile. 20. An XDI system in accordance with claim 14, wherein said secondary collimator has an at least partially non-planar profile. |
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052296164 | summary | Background of the Invention This invention relates to a lamp which is operable as an exposure light source of an exposure apparatus. An exposure apparatus of the type described is very often used in manufacturing a semiconductor device so as to project a mask pattern onto a semiconductor wafer and to expose the semiconductor wafer. To this end, the exposure apparatus comprises a lamp which is often called an exposure lamp and which is operable as an exposure light source in addition to an optical system which guides an optical beam from the lamp to the semiconductor wafer. It is noted that the lamp has major influence on the size of each of the products, namely, semiconductor devices, as known in the art. Specifically, a minimum pattern size of the semiconductor devices is dependent on the wavelength of the optical beam emitted from the exposure light source. Heretofore, a high pressure mercury lamp is used as such an exposure light source. In the high pressure mercury lamp, mercury which is enveloped in solid state form is vaporized and kept at a high pressure of several tens of atmospheres when it is excited by supply of an electric voltage. The high pressure mercury lamp can emit optical beams which have a spectrum distribution. The spectrum distribution includes spectra of specific wavelengths which are equal to 436 nm and 365 nm and which may be called g and i lines, respectively. The g line is suitable for manufacturing a dynamic random access memory of 4 megabits which may be referred to as 4 MDRAM and which has a minimum pattern size of 0.8 micron meters while the i line is used for manufacturing a dynamic random access memory of 16 megabits which may be referred to as 16 MDRAM and which has a minimum pattern size of 0.5 to 0.6 micron meters. Herein, it is mentioned that recent interest is mainly directed to a dynamic random access memory of 64 megabits which may be called a 64 MDRAM and which is considered as a very large scale integrated memory of the succeeding generation. In order to actually manufacture the 64 MDRAM, a line and space size, namely, a minimum pattern size should be equal to or smaller than 0.3 micron meter. This means that an exposure light source must emit a beam having an exposure wavelength of 250 nm or so within a far ultraviolet waveband. As a conventional exposure light source for emitting far ultraviolet light, an excimer laser of krypton-fluoride (KrF) has been known which emits a laser wavelength of 248 nm. It is mentioned here that the far ultraviolet light is projected onto a semiconductor wafer through an optical lens system of a reduction projection type. In such an optical lens system, use must be made of a lens material which exhibits a low absorptivity for the far ultraviolet light. The lens material may be, for example, fluorspar (CaF.sub.2) or quartz. However, fluorspar has difficulty as regards precise processing and temperature control. Under the circumstances, only synthetic quartz is practically used as the lens material of the optical lens system for the exposure light source. This shows that each lens of the optical lens system must be formed by a single material, namely, quartz which has an identical refraction coefficient. In general, an achromatic lens is formed by a combination of lenses which have different refraction coefficients. Therefore, the above-mentioned optical lens system in question can not include an achromatic lens or lenses and is structured as a monochromatic lens system. It is to be noted that quartz has a very large dispersion within the far ultraviolet region. Therefore, a spectrum bandwidth of an exposure wavelength must be adjusted in the optical lens system so that a half band width becomes equal to or less than 0.003 nm. When the excimer laser is used as the exposure light source to emit an optical beam, the optical beam emitted from a excimer laser has the spectrum bandwidth which is as wide as 0.3 nm. Such a wide spectrum bandwidth should be narrowed in some way so as to send the optical beam to the semiconductor wafer through the optical lens system, as mentioned above, and to expose the semiconductor wafer to the optical beam. Even if the optical beam has a narrow bandwidth, speckles are liable to occur on the semiconductor wafer, which makes it difficult to obtain a line and space of 0.3 micron meters. In addition, a halogen gas is inevitably used as a laser gas in the excimer laser. This necessitates large supplementary equipment for handling the halogen gas and evacuating it. Furthermore, running costs are very high as well as manufacturing costs. SUMMARY OF THE INVENTION It is an object of this invention to provide an exposure light source which is suitable for manufacturing a very large scale integrated memory without large-scale supplementary equipment. It is another object of this invention to provide an exposure light source of the type described, which is capable of producing a line and space necessary for the very large scale integrated memory. It is still another object of this invention to provide an exposure lamp which is capable of obtaining the high resolution necessary for a very large scale integrated memory without any supplementary equipment. It is yet another object of this invention to provide an exposure lamp of the type described which is comparatively inexpensive. According to this invention, a lamp is used in emitting far ultraviolet light to illuminate a substrate. The lamp envelopes a metal element which substantially consists of a single isotope such that the far ultraviolet light is emitted on vaporization of the single isotope. |
043409709 | abstract | Power wheel comprises a heat engine consisting of several expansion valves rigidly situated around the circumference of a stationary side gear and centralized to a axis which rotates through the work out put by expanding a fluid centralized inside a valve unit by a heat source operating in intervals introduced through the rotation of the expansion valves by moving a valve plunger in or out of the heat elements which in return will expand or contract a fluid, to move push rods stroke wise in both directions to activate a spindle drive by rotating a pinion gear with ratchet units to achieve rotation in one direction. |
claims | 1. A charged particle beam irradiation system comprising:an irradiation unit configured to irradiate an irradiation target with a charged particle beam;a radiation resistance state measuring section configured to measure a radiation resistance state of the irradiation target;a region dividing section configured to divide the irradiation target into a plurality of radiation resistance regions based on a measurement result of the radiation resistance state measuring section;a radiation dose computing section configured to compute a planned value of a radiation dose of the charged particle beam for each of the plurality of radiation resistance regions divided by the region dividing section; andan irradiation planning section configured to create an irradiation plan of the charged particle beam with respect to the irradiation target based on a computation result of the radiation dose computing section. 2. The charged particle beam irradiation system according to claim 1, further comprising:a pre-irradiation planning section configured to create a pre-irradiation plan of the charged particle beam with respect to the irradiation target based on a captured image of the irradiation target,wherein the irradiation planning section corrects the pre-irradiation plan based on the computation result of the radiation dose computing section to create the irradiation plan. 3. The charged particle beam irradiation system according to claim 1,wherein the irradiation planning section creates the irradiation plan based on a captured image of the irradiation target and the computation result of the radiation dose computing section. 4. The charged particle beam irradiation system according to claim 1,wherein the radiation resistance state measuring section includes a gamma-ray detector, andwherein the region dividing section divides the irradiation target into the plurality of radiation resistance regions based on a measurement result of the gamma-ray detector that uses FMISO as a tracer. 5. The charged particle beam irradiation system according to claim 1,wherein the radiation resistance state measuring section includes a gamma-ray detector, andwherein the region dividing section divides the irradiation target into the plurality of radiation resistance regions based on a measurement result of the gamma-ray detector that uses FAZA as a tracer. 6. The charged particle beam irradiation system according to claim 1,wherein the radiation resistance state measuring section includes a gamma-ray detector, andwherein the region dividing section divides the irradiation target into the plurality of radiation resistance regions based on a measurement result of the gamma-ray detector that uses IAZA as a tracer. 7. The charged particle beam irradiation system according to claim 1,wherein the radiation resistance state measuring section includes a gamma-ray detector, andwherein the region dividing section divides the irradiation target into the plurality of radiation resistance regions based on a measurement result of the gamma-ray detector that uses FETNIM as a tracer. 8. A charged particle beam irradiation planning method comprising:a radiation resistance state measuring step of measuring a radiation resistance state of an irradiation target;a region dividing step of dividing the irradiation target into a plurality of radiation resistance regions based on a measurement result of the radiation resistance state measuring step;a radiation dose computing step of computing a planned value of a radiation dose of a charged particle beam for each of the plurality of radiation resistance regions divided in the region dividing step; andan irradiation planning step of creating the irradiation plan of the charged particle beam with respect to the irradiation target based on a computation result of the radiation dose computing step. 9. The charged particle beam irradiation planning method according to claim 8, further comprising:a pre-irradiation planning step of creating a pre-irradiation plan of the charged particle beam with respect to the irradiation target based on a captured image of the irradiation target before the radiation resistance state measuring step,wherein in the irradiation planning step, the pre-irradiation plan is corrected based on the computation result of the radiation dose computing step to create the irradiation plan. 10. The charged particle beam irradiation planning method according to claim 8,wherein in the irradiation planning step, the irradiation plan is created based on a captured image of the irradiation target and the computation result of the radiation dose computing step. 11. The charged particle beam irradiation planning method according to claim 8,wherein in the radiation resistance state measuring step, the radiation resistance state of the irradiation target is measured based on a gamma-ray detection that uses FMISO as a tracer. 12. The charged particle beam irradiation planning method according to claim 8,wherein in the radiation resistance state measuring step, the radiation resistance state of the irradiation target is measured based on a gamma-ray detection that uses FAZA as a tracer. 13. The charged particle beam irradiation planning method according claim 8,wherein in the radiation resistance state measuring step, the radiation resistance state of the irradiation target is measured based on a gamma-ray detection that uses IAZA as a tracer. 14. The charged particle beam irradiation planning method according to claim 8,wherein in the radiation resistance state measuring step, the radiation resistance state of the irradiation target is measured based on a gamma-ray detection that uses FETNIM as a tracer. |
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claims | 1. Article for increasing Leidenfrost temperature comprising:a surface with a plurality of multiscale hierarchical structures including microscale and nanoscale features both having super wetting properties and forming protrusions, cavities, or combinations of both on the surface, the structures having a median size a, median aspect ratio h/a and median spacing ratio b/a selected to increase Leidenfrost temperature. 2. The article of claim 1 wherein the protrusions and cavities can have multiscale hierarchical features Wherein median size of an “nth” level hierarchy is an, aspect ratio (h/a)n, and spacing ratio (ba)n. 3. The article of claim 1 wherein a is less than 1000 microns. 4. The article of claim 3 wherein 0.01<b/a<30. 5. The article of claim 3 wherein 0.01<h/a<50. 6. The article of claim 3 wherein 0.1<b/a<20. 7. The article of claim 3 wherein 0.1<h/a<20. 8. The article of claim 3 wherein 0.1<b/a<10. 9. The article of claim 3 wherein 0.1<h/a<5. 10. The article of claim 1 where in a<100 microns. 11. The article of claim 10 wherein 0.01<b/a<30. 12. The article of claim 10 wherein 0.01<h/a<50. 13. The article of claim 10 wherein 0.1<b/a<20. 14. The article of claim 10 wherein 0.1<h/a<20. 15. The article of claim 10 wherein 0.1<b/a<10. 16. The article of claim 10 wherein 0.1<h/a<5. 17. The article of claim 1 wherein a<20 microns. 18. The article of claim 17 wherein 0.01<b/a<30. 19. The article of claim 17 wherein 0.01<h/a<50. 20. The article of claim 17 wherein 0.1<b/a<20. 21. The article of claim 17 wherein 0.01<h/a<20. 22. The article of claim 17 wherein 0.1<b/a<10. 23. The article of claim 17 wherein 0.1<h/a<5. 24. The article of claim 2 wherein an+1/an<0.1. 25. The article of claim 24 wherein 0.01<b/a<30. 26. The article of claim 24 wherein 0.01<h/a<50. 27. The article of claim 24 wherein 0.1<b/a<20. 28. The article of claim 24 wherein 0.1<h/a<20. 29. The article of claim 24 wherein 0.1<b/a<10. 30. The article of claim 24 wherein 0.1<h/a<5. 31. The article of claim 1 wherein the structures have shapes selected from the group consisting of a prism, sphere, polyhedron, cone and combinations thereof. 32. The article of claim 1 wherein the structures are made of materials with an intrinsic wetting angle less than 90 degrees. 33. The article of claim 1 wherein the structures are made of materials with an intrinsic wetting angle less than 50 degrees. 34. The article of claim 1 wherein the structures are made of materials with an intrinsic wetting angle less than 20 degrees. 35. The article of claim 1 selected from the group consisting of metal, ceramic, polymer, intermetallic, cermet, semimetal. 36. The article of claim 1 wherein the structures are multiscale structures formed from a combination of metal, ceramic, polymer, intermetallic, cermet, semimetal. 37. The article of claim 1 wherein a high-surface energy surface modification layer is further deposited on the surface to increase Leidenfrost temperature. 38. The article of claim 37 wherein the surface modification layer is a coating selected from the group consisting of a ceramic, polymer, metal, cermet, intermetallic. 39. The article of claim 38 wherein the coating comprises the surface energy modification coating layer, wherein the layer comprises a ceramic material, a hydrophilic polymer material, or a combination comprising at least one of the foregoing materials; wherein the ceramic material comprises titanium oxide, silicon oxide, copper oxide, aluminum oxide, 460 magnesium oxide, zirconium oxide, zinc oxide, iron oxide, yttrium stabilized zirconia, magnesium aluminate spinel, aluminum nitride, gallium nitride, silicon carbide, tungsten carbide cobalt chromium, or a combination comprising at least one of the foregoing. 40. The article of claim 37 wherein the surface modification layer is ion implanted. 41. The article of claim 37 wherein the surface modification layer is a diffusion layer. 42. The article of claim 37 wherein the surface modification layer is a self-assembled monolayer. 43. The article of claim 1 including multiscale structures wherein the multiscale structures are fabricated via heat treatment. 44. The article of claim 1 including multiscale structures wherein the multiscale structures are fabricated via deposition or growth of smaller length scale features onto larger length scale features. 45. The article of claim 1 or 2 wherein the surface is a boiler surface. 46. The article of claim 1 or 2 wherein the surface is an evaporator surface. 47. The article of claim 1 or 2 wherein the surface is a nuclear fuel rod and cladding surface. 48. A fuel rod comprising of preferential moderation sites that comprise of the article of claim 1 that are disposed on the surface of the fuel rod that is in contact with the fluid flowing or impinging on the surface to locally control fission at these preferred sites and maintain integrity of the fuel. 49. A fuel rod array wherein select fuel rod surfaces will be disposed with the article of claim 1 for preferential moderation. 50. The article of claim 1 wherein the surface is a heat transfer surface that is cooled by impingement of fluid. 51. The article of claim 50 wherein the fluid is selected from the group consisting of aqueous and non-aqueous liquids including water, organic liquids, mixtures, oils, emulsions, liquid metals, liquid nitrogen, liquid CO2, hydrocarbon liquids, liquefied hydrocarbons, liquid helium, and liquefied rare gases. 52. The article of claim 1 wherein the surface is on an electronic or photonic device. |
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summary | ||
abstract | A device for separating a chemical mixture into its constituents includes a central cathode that is aligned axially within a cylindrical plasma chamber. An anode, made of the chemical mixture requiring separation is positioned near the cylindrical wall of the plasma chamber. A working gas is introduced into the chamber to sputter the chemical mixture into the plasma chamber where it is dissociated and ionized. To reduce the unwanted loss of the central cathode due to sputtering by the working gas, the central cathode is formed with a plurality of radial projections that extend outwardly from the axis of the cylindrical plasma chamber. These radial projections act to capture sputtered cathode material before it is lost to the plasma. Once the chemical mixture has been ionized in the plasma chamber, the ions are separated, according to their respective mass to charge ratio, using crossed electric and magnetic fields. |
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abstract | An accelerator system is provided. According to an example, the accelerator system includes a ray source, a multi-leaf collimator including leaves, a multi-leaf collimator controller and a leaf position determining device. The multi-leaf collimator controller is configured to control each of the leaves to move according to a predetermined position. The leaf position determining device is configured to obtain a three-dimensional image of the multi-leaf collimator, determine a sub-field shape and a sub-field size of the multi-leaf collimator according to the three-dimensional image, determine an actual position of each of the leaves according to the sub-field shape and the sub-field size and obtain an error value for each of the leaves by comparing the actual position with the predetermined position for each of the leaves. In this way, the error value for each of the leaves may be used to control operation of the accelerator system. |
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054897378 | claims | 1. A processing system for handling radioactive waste after a part or all of a waste slurry, which is comprised of an aqueous slurry of used ion exchange resin generated from nuclear facilities and which is stored in storage tanks, is withdrawn to an adjusting tank and after radioactive concentration of respective nuclides in the waste slurry is measured, which comprises at least one of the following steps: 1) returning the waste slurry to the storage tank without introducing the slurry into a solidifying processing system when the waste slurry is determined to be improper for processing directly, based on the measured radioactive concentration; 2) subjecting the waste slurry to an incineration process when the waste slurry is determined, based on the measured radioactive concentration, to be proper for processing by incineration, and subsequently solidifying residual material remaining from the incineration process by a solidifying process; 3) treating the waste slurry by a thermal decomposition process for removing functional groups and radioactive nuclides when the waste slurry is determined, based on the measured radioactive concentration, to be proper for further processing, and subsequently solidifying residual material obtained from the thermal decomposition process by a solidifying process; 4) subjecting the waste slurry to a dissolution process by removing radioactive nuclides when the waste slurry is determined to be proper for processing to reduce its radioactivity, based on the measured radioactive concentration, and subsequently solidifying the processed waste slurry by a solidifying process; 5) introducing the waste slurry into a solidifying process wherein the waste slurry is solidified with inorganic solidifying materials or organic solidifying materials when the waste slurry is determined to be proper for processing directly, based on the measured radioactive concentration; 6) incinerating the waste slurry when the waste slurry is determined, based on the measured radioactive concentration, to be proper for processing directly, and introducing the gaseous waste generated from the incineration of the waste slurry into a gaseous waste treatment process; or 7) diluting the waste slurry with other waste having low radioactivity when the waste slurry is determined, based on the measured radioactive concentration, to be proper for processing after reducing its level of radioactivity by dilution, so that the radioactivity of a solidified waste body obtained by a subsequent solidifying process exhibits a value within a predetermined limit which is defined as an allowable value for handling of such radioactive waste. said measurement of radioactive concentration of respective nuclides is performed on at least either of Co-60 and Cs-137, and, when an observed value obtained by said measurement exceeds a predetermined value, said waste is returned to the storage tank without performing the measurement of radioactive concentration on another nuclide. said measurement of radioactive concentration of respective nuclides is performed on at least either of Co-60 and Cs-137, and when radioactivities of nuclides other than Co-60 and Cs-137, which are estimated by a scaling factor method for estimating radioactivity of a nuclide based on observed radioactivity of the at least either of Co-60 and Cs-137 with a preferable safety factor, do not exceed predetermined values, transferring said waste to a next processing system with measurement of radioactivity on only either of Co-60 and Cs-137. observed values in said measurement of radioactivity are recorded as data, and said data are stored corresponding to data on radioactivity of each package obtained by the solidifying process. 2. A processing system for radioactive waste as claimed in claim 1, wherein said measurement of radioactive concentration of respective nuclides is performed qualitatively and quantitatively by using chemical analysis technique. 3. A processing system for radioactive waste as claimed in claim 1, wherein said measurement of radioactive concentration of respective nuclides is performed, in aspect of the processing applicability evaluation, on at least one of; Co-60, Cs-137, Tc-99, Ni-59, Ni-63, Sr-90, I-129, Nb-94, C-14, H-3, and transuranium elements. 4. A processing system for radioactive waste as claimed in claim 3, wherein said measurement of radioactive concentration of respective nuclides is performed additionally on gamma ray nuclides in aspect of evaluation on compliance with transportation standard after the solidification. 5. A processing system for radioactive waste as claimed in claim 1, wherein 6. A processing system for radioactive waste as claimed in claim 14, wherein 7. A processing system for radioactive waste as claimed in claim 14, wherein 8. A processing system for radioactive waste as claimed in claim 14, wherein measurement of radioactivity is performed on said waste before a solidifying process and after a solidifying process. 9. A processing system for radioactive waste as claimed in claim 8, wherein measurement of radioactivity on said waste before solidifying process is performed by sampling of said waste directly from the adjusting tank or through a sampling port furnished to the adjusting tank. 10. A processing system for radioactive waste as claimed in claim 8, wherein at least a measuring point for the measurement of radioactivity on said waste is provided in the system. 11. A processing system for radioactive waste as claimed in claim 14, wherein measurement of radioactivity is performed before and after a solidifying process, and the waste includes condensed waste liquid, incinerated ashes, and non-burnable miscellaneous solid bodies. |
description | This application is a continuation-in-part of and claims the priority of U.S. non-provisional application Ser. No. 13/566,447, filed Aug. 3, 2012 which claims claims the priority of U.S. provisional application Ser. No. 61/514,534, filed Aug. 3, 2011, each of which is hereby incorporated by reference in its entirety. This application relates to methods of separating osmium from a mixture that includes the osmium and at least one additional metal. More specifically, this application relates to a method of oxidizing, selectively removing or distilling, trapping, and precipitating chemically pure osmium from a mixture that includes the osmium and at least one additional metal. Osmium is an element with numerous useful physical and chemical properties. Osmium is the densest known element and possesses an extremely high hardness comparable to diamond. Wear-resistant osmium alloys are used in the construction of many devices such as fountain pen tips and electrical contacts. Osmium is also used as a contrast agent for staining and fixing lipids prior to visualization using transmission electron microscopy. Osmium is also a useful material in the production of radioisotopes. Osmium-192 or osmium-189 may also be irradiated on a cyclotron to produce Re-186, a radioisotope with many potential applications to nuclear imaging methods and therapeutic compositions. Osmium-190 may be irradiated in a thermal neutron flux to produce the radioactive isotope Os-191. Os-191 decays to Ir-191m, another radioisotope that is an excellent source of x-rays. Os-191/Ir-191m radioisotope generation systems may be used to provide Ir-191m radioisotope for dynamic radiotracer studies such as angiography. Due to the toxicity of osmium, medical technologies that make use of osmium are limited in the amount of osmium they may use. In order to minimize the amount of osmium necessary to achieve the desired effect, a higher purity of osmium may be desirable. Existing techniques of producing chemically pure osmium typically involve nitric acid oxidation of a mixture containing the osmium and the fusion of the Os metal with KNOB/KOH at high temperatures. Unless the mixture is made up of finely divided particles, the nitric acid oxidation may be a lengthy process. During the purification of osmium radioisotopes, the lengthy process times may expose technicians to extensive dosages of hazardous ionizing radiation. A need in the art exists for a method of separating an amount of osmium from a mixture of the osmium and at least one other metal in a relatively short time compared to existing methods without need for high temperatures. Such a process may be used to produce chemically pure osmium samples in a shorter time using relatively simple chemical reactions and equipment. The shortened process times further limit the exposure of technicians to potentially hazardous conditions, particularly in the production of chemically pure osmium radioisotopes. In one aspect, a method of separating an amount of osmium from a mixture that includes the osmium and at least one additional metal is provided. This method includes contacting the mixture with an oxidizing solution to form a volatile OsO4 vapor and bubbling the OsO4 vapor through a KOH trapping solution to form an amount of K2[OsO4(OH)2] dissolved in the KOH trapping solution. This method further includes contacting the dissolved K2[OsO4(OH)2] with a reducing agent to form an Os precipitate and separating the Os precipitate from the KOH trapping solution. The mixture may be an irradiated osmium metal target that includes at least one osmium isotope selected from the group consisting of Os-184, Os-186, Os-187, Os-188, Os-189, Os-190, Os-192, and any combination thereof. The mixture may be an irradiated osmium metal target that includes at least one of Os-189 and/or Os-192, and the at least one additional metal may include at least one of Re-186, Ir-186, IR-187, Ir-188, and/or Ir-190. The oxidizing solution may include an aqueous solution of an oxidizing agent chosen from NaClO, LiClO, KClO, NalO4, Na2S4O8, XeO3, NaClO2, NaClO3, NaClO4, NaOH in contact with Cl2 gas, other alkali salts of ClO, ClO2, ClO3 and/or ClO4. The oxidizing solution may be an aqueous solution of NaClO at a concentration of about 12% available chlorine. The mixture may be contacted with the oxidizing solution in an impinger device. The mixture may be contacted with the oxidizing solution at a temperature of about 40° C. until the mixture is dissolved, and the dissolved mixture may be contacted with the oxidizing solution at a temperature of about 90° C. The KOH trapping solution may include an aqueous solution of KOH at a concentration of about 25% w/v. The KOH trapping solution may be maintained at a temperature of less than about 5° C. The reducing agent may be chosen from absolute ethanol, Zn shavings, Al shavings, Mg shavings, NaBH4, NaHS, H2S gas, Na2S2O3, UV light, phosphine ligands, hydrazine, hydroquinone, hydrophosphorous acid, formaldehyde, hydroxylamine, and citrate. The reducing agent may be absolute ethanol at a concentration of 5% v/v and the Os precipitate may be K2[OsO2(OH)4]. The reducing agent may be a mixture of Zn shavings and Al shavings, and the Os precipitate may be Os metal. The reducing agent is chosen from Zn shavings, Mg shavings, and Al shavings, the reducing agent is contacted with the dissolved K2[OsO4(OH)2] in combination with HCl, and the Os precipitate may be Os metal. The reducing agent may be NaBH4 and the Os precipitate may be Os metal. The reducing agent may be chosen from NaHS, H2S gas, Na2S2O3 and the Os precipitate may be OsS2. The remaining mixture in the oxidizing solution may be contacted with a reducing agent to form an osmium-free mixture that may include the at least one additional metal. In another aspect, a method of separating an amount of osmium from a mixture that includes the amount of osmium and at least one additional metal is provided. This method includes contacting the mixture with an aqueous solution of NaClO at a concentration of about 12% available chlorine to form a volatile OsO4 vapor and bubbling the OsO4 vapor through a trapping solution that includes an aqueous solution of KOH at a concentration of about 25% w/v to form an amount of dissolved K2[OsO4(OH)2]. This method further includes contacting the dissolved K2[OsO4(OH)2] with an aqueous solution of NaHS at a concentration of about 10% w/v to form an OsS2 precipitate. In addition, this method includes washing the OsS2 precipitate by agitating with water, separating the OsS2 precipitate from the KOH trapping solution by centrifuging, rinsing the OsS2 precipitate with acetone or other organic solvents to further remove the water from the precipitate, and drying the OsS2 precipitate. The mixture may be an irradiated osmium metal target that includes at least one osmium isotope selected from the group consisting of Os-184, Os-186, Os-187, Os-188, Os-189, Os-190, Os-192, and any combination thereof. The mixture may be an irradiated osmium metal target that includes at least one of Os-189 and/or Os-192, and the at least one additional metal may include at least one of Re-186, Ir-186, IR-187, Ir-188, and/or Ir-190. The mixture may be contacted with the aqueous solution of NaClO at a temperature of about 40° C. until completely dissolved, and the dissolved mixture may be contacted with the aqueous solution of NaClO at a temperature of about 90° C. The dissolved mixture may be contacted with the aqueous solution of NaClO until the aqueous solution of NaClO is colorless. The trapping solution may be situated within an ice bath while the dissolved K2[OsO4(OH)2] is forming. In one aspect, a method of separating an amount of osmium from a mixture that includes the osmium and at least one additional metal is provided. This method includes contacting the mixture with an oxidizing solution to form a volatile OsO4 vapor and bubbling the OsO4 vapor through a KOH trapping solution to form an amount of K2[OsO4(OH)2] dissolved in the KOH trapping solution. This method further includes contacting the dissolved K2[OsO4(OH)2] with a reducing agent to form an Os precipitate and separating the Os precipitate from the KOH trapping solution. The mixture may be an irradiated osmium metal target that includes at least one osmium isotope selected from the group consisting of Os-184, Os-186, Os-187, Os-188, Os-189, Os-190, Os-192, and any combination thereof. The mixture may be an irradiated osmium metal target that includes at least one of Os-189 and/or Os-192, and the at least one additional metal may include at least one of Re-186, Ir-186, IR-187, Ir-188, and/or Ir-190. The mixture may be contacted with the aqueous solution of NaClO at a temperature of about 40° C. until completely dissolved, and the dissolved mixture may be contacted with the aqueous solution of NaClO at a temperature of about 90° C. The dissolved mixture may be contacted with the aqueous solution of NaClO until the aqueous solution of NaClO is colorless. The trapping solution may be situated within an ice bath while the dissolved K2[OsO4(OH)2] is forming. In an additional aspect, a method of producing an amount of chemically pure Re-186 isotope is provided. This method includes irradiating a metal target that includes an amount of isotopically enriched osmium isotope consisting of Os-189, Os-192, or any combination thereof in a thermal proton flux to form a mixture that includes at least one osmium isotope and at least one additional metal that may include the Re-186 isotope. The method also includes contacting the mixture with an oxidizing solution to form a volatile OsO4 vapor comprising the at least one osmium isotope, distilling the OsO4 vapor out of the oxidizing solution to form a second solution comprising the Re-186 dissolved in the oxidizing solution, and separating the Re-186 isotope from the second solution. The Re-186 isotope may be separated from the second solution using a method chosen from: contacting the second solution with a reducing agent, contacting the second solution with a chromatographic column, and electroplating the at least one metal from the second solution. The Re-186 isotope may be separated from the second solution by contacting the second solution with an alumina chromatographic column and eluting the Re-186 using a saline solution In one aspect, a method of separating an amount of osmium from a mixture that includes the osmium and at least one additional metal is provided. This method includes contacting the mixture with an oxidizing solution to form a volatile OsO4 vapor and bubbling the OsO4 vapor through a KOH trapping solution to form an amount of K2[OsO4(OH)2] dissolved in the KOH trapping solution. This method further includes contacting the dissolved K2[OsO4(OH)2] with a reducing agent to form an Os precipitate and separating the Os precipitate from the KOH trapping solution. The mixture may be an irradiated osmium metal target that includes at least one osmium isotope selected from the group consisting of Os-184, Os-186, Os-187, Os-188, Os-189, Os-190, Os-192, and any combination thereof. The mixture may be an irradiated Os-190 metal target, the amount of osmium includes an amount of Os-191, and the at least one additional metal is chosen from Ir-192, Ir-193, Ir-194, Pt-192, and/or Pt-194. The oxidizing solution may include an aqueous solution of an oxidizing agent chosen from NaClO, LiClO, KClO, NalO4, Na2S4O8, XeO3, NaClO2, NaClO3, NaClO4, NaOH in contact with Cl2 gas, other alkali salts of ClO, ClO2, ClO3 and/or ClO4. The oxidizing solution may be an aqueous solution of NaClO at a concentration of about 12% available chlorine. The mixture may be contacted with the oxidizing solution in an impinger device. The mixture may be contacted with the oxidizing solution at a temperature of about 40° C. until the mixture is dissolved, and the dissolved mixture may be contacted with the oxidizing solution at a temperature of about 90° C. The KOH trapping solution may include an aqueous solution of KOH at a concentration of about 25% w/v. The KOH trapping solution may be maintained at a temperature of less than about 5° C. The reducing agent may be chosen from absolute ethanol, Zn shavings, Al shavings, Mg shavings, NaBH4, NaHS, H2S gas, Na2S2O3, UV light, phosphine ligands, hydrazine, hydroquinone, hydrophosphorous acid, formaldehyde, hydroxylamine, and citrate. The reducing agent may be absolute ethanol at a concentration of 5% v/v and the Os precipitate may be K2[OsO2(OH)4]. The reducing agent may be a mixture of Zn shavings and Al shavings, and the Os precipitate may be Os metal. The reducing agent is chosen from Zn shavings, Mg shavings, and Al shavings, the reducing agent is contacted with the dissolved K2[OsO4(OH)2] in combination with HCl, and the Os precipitate may be Os metal. The reducing agent may be NaBH4 and the Os precipitate may be Os metal. The reducing agent may be chosen from NaHS, H2S gas, Na2S2O3 and the Os precipitate may be OsS2. The remaining mixture in the oxidizing solution may be contacted with a reducing agent to form an osmium-free mixture that may include the at least one additional metal. In another aspect, a method of separating an amount of osmium from a mixture that includes the amount of osmium and at least one additional metal is provided. This method includes contacting the mixture with an aqueous solution of NaClO at a concentration of about 12% available chlorine to form a volatile OsO4 vapor and bubbling the OsO4 vapor through a trapping solution that includes an aqueous solution of KOH at a concentration of about 25% w/v to form an amount of dissolved K2[OsO4(OH)2]. This method further includes contacting the dissolved K2[OsO4(OH)2] with an aqueous solution of NaHS at a concentration of about 10% w/v to form an OsS2 precipitate. In addition, this method includes washing the OsS2 precipitate by agitating with water, separating the OsS2 precipitate from the KOH trapping solution by centrifuging, rinsing the OsS2 precipitate with acetone or other organic solvents to further remove the water from the precipitate, and drying the OsS2 precipitate. The mixture may be an irradiated osmium metal target that includes at least one osmium isotope selected from the group consisting of Os-184, Os-186, Os-187, Os-188, Os-189, Os-190, Os-192, and any combination thereof. The mixture may be an irradiated Os-190 metal target, the amount of osmium includes an amount of Os-191, and the at least one additional metal is chosen from Ir-192, Ir-193, Ir-194, Pt-192, and/or Pt-194. The mixture may be contacted with the aqueous solution of NaClO at a temperature of about 40° C. until completely dissolved, and the dissolved mixture may be contacted with the aqueous solution of NaClO at a temperature of about 90° C. The dissolved mixture may be contacted with the aqueous solution of NaClO until the aqueous solution of NaClO is colorless. The trapping solution may be situated within an ice bath while the dissolved K2[OsO4(OH)2] is forming. In an additional aspect, a method of producing an amount of chemically pure Os-191 isotope is provided. This method includes irradiating a metal target that includes an amount of isotopically enriched Os-190 metal in a thermal neutron flux to form a mixture that includes Os-191 isotope and at least one additional metal chosen from Ir-192, Ir-193, Ir-194, Pt-192, Pt-194, and combinations thereof. The method also includes contacting the mixture with an aqueous solution of NaClO at a concentration of about 12% available chlorine to form a volatile OsO4 vapor comprising Os-191, as well as bubbling the OsO4 vapor through a trapping solution that includes an aqueous solution of KOH at a concentration of about 25% w/v to form an amount of dissolved K2[OsO4(OH)2] that contains the Os-191. The method also includes contacting the dissolved K2[OsO4(OH)2] with an aqueous solution of NaHS at a concentration of about 10% w/v to form an OsS2 precipitate containing the Os-191. Further included in the method is washing the resulting OsS2 precipitate by agitating with water, separating the washed OsS2 precipitate from the KOH trapping solution by centrifuging, rinsing the separated OsS2 precipitate with acetone, and drying the rinsed OsS2 precipitate. In another additional aspect, a method of removing an osmium impurity from a mixture comprising the osmium impurity and at least one additional metal is provided. This method includes contacting the mixture with an oxidizing solution to form a volatile OsO4 vapor and distilling the OsO4 vapor out of the oxidizing solution to form a second solution that includes the at least one additional metal dissolved in the oxidizing solution. The method further includes separating the at least one additional metal from the second solution. The at least one additional metal may be separated from the second solution using a method chosen from contacting the second solution with a reducing agent, contacting the second solution with a chromatographic column, and electroplating the at least one metal from the second solution. Various aspects of the separation method overcome many of the limitations of existing osmium separation methods. The use of oxidizing solutions such as aqueous solutions of NaClO results in considerably shorter reaction times compared to existing methods. The KOH trapping solution does not require high temperature conditions, unlike existing osmium separation methods. Various aspects of this method of separating osmium may be performed in a relatively short time using simple equipment and readily available materials compared to existing methods. Further, the simple equipment may be housed within a shielded glove box or hot cell, minimizing the exposure of workers to radiation and limiting the possibility of inadvertent environmental release of osmium. Other aspects and iterations of the embodiments are described in detail below. Corresponding reference characters and labels indicate corresponding elements among the views of the drawings. The headings used in the figures should not be interpreted to limit the scope of the claims. Various aspects provide methods of separating osmium from a mixture of metals including the osmium as well as at least one other metal. These methods may be used to produce chemically pure osmium or to remove osmium impurities from a mixture that includes the osmium and at least one other metal. In an aspect, the chemically pure osmium may be a radioisotope. In another aspect, the chemically pure osmium may be used as an isotopically-enriched osmium target to produce radioisotopes by irradiation of the osmium target in an irradiation source including, but not limited to, a thermal neutron flux, a cyclotron, or a linear accelerator. The osmium target may be irradiated by any known type of irradiation, including, but not limited to: proton irradiation, neutron irradiation, deuteron irradiation, alpha particle irradiation, and any other known type of irradiation. A flowchart describing an aspect of a method 100 is provided in FIG. 1. In this aspect, the osmium within the mixture is oxidized in an oxidizing solution within an impinger to produce gaseous OsO4 vapor at step 102. The OsO4 vapor is bubbled through a KOH trapping solution at step 104, where the OsO4 reacts with the KOH in the trapping solution to form dissolved K2[OsO4(OH)2]. The K2[OsO4(OH)2] is then contacted with a reducing agent at step 106 to form an osmium-containing precipitate. Non-limiting examples of osmium-containing precipitate include osmium metal, OsO2, OsS2, K2[OsO2(OH)4], and any combination thereof. The mixture may include a variety of radioactive and non-radioactive isotopes. Non-limiting examples of metals that may be included in the mixture include lanthanide metals, transition metals, alkali metals, and metals from the platinum family. Non-limiting examples of specific metal elements that may be included with osmium in a mixture include Rh, Pd, Ir, Pt, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Y, Zr, Nb, Mo, Tc, Ag, Cd, La, Hf, Ta, W, Re, Au, Hg, Ac, Rf, Db, Sg, Bh, Hs, Mt, Ds, Rg, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Ac, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es, Fm, Md, No, Lr, and any combination thereof. In an aspect, the mixture may result from the irradiation of an enriched Os-190 target by a thermal neutron stream and may include Os-191, Os-191m, Ir-191m, Ir-191, Ir-192m, Ir-192, Ir-193, Ir-194, Pt-192, Pt-194, and combinations thereof. In another aspect, enriched Os-190 may be irradiated in a cyclotron to produce a mixture that may include Pt-188, Pt-189, Pt-191, Pt-193m, Pt-195m, and combinations thereof. In an additional aspect, enriched Os-190 may be irradiated in a cyclotron to produce a mixture that may include Re-186, an isotope with at least several potential applications including but not limited to nuclear imaging methods and therapeutic compositions. In other aspects, the mixture may result from the irradiation of any enriched Os isotope target without limitation. Non-limiting examples of Os isotopes suitable for inclusion in an enriched Os isotope target include Os-184, Os-186, Os-187, Os-188, Os-189, Os-190, and Os-192. In one aspect, the mixture may result from the proton irradiation of an enriched Os-189 and/or Os-192 target to produce a mixture that may include Re-186. Other aspects of methods of separating osmium from a mixture of metals are described in detail below. In various aspects, the osmium within a mixture may be separated from a mixture that includes the osmium and at least one other metal by contacting the mixture with an oxidizing agent to form a volatile OsO4 vapor. In an aspect, the oxidizing agent may be any compound capable of oxidizing the osmium into OsO4 in aqueous solution. In another aspect, a relatively strong oxidizing agent may be selected to reduce the overall time to separate the osmium from the mixture and to provide the capability to oxidize the osmium from mixtures in any form including, but not limited to: finely divided powders, shavings, pellets, slugs, and any combination thereof. Non-limiting examples of suitable oxidizing agents include NaClO, LiClO, KClO, NalO4, Na2S4O8, XeO3, NaClO2, NaClO3, NaClO4, NaOH in contact with Cl2 gas, other alkali salts of ClO, ClO2, ClO3 and ClO4, and combinations thereof. The concentration of the oxidizing agent in an aqueous solution may range from about 5% to about 30% available chlorine. The concentration of oxidizing agent may be selected based on any one or more of at least several factors including, but not limited to: the composition of the mixture, the availability of the oxidizing agent, the safety and ease of use of the oxidizing agent, the temperature and other reaction conditions of the mixture and the oxidizing agent, and the solubility of the oxidizing agent. The temperature at which the mixture is contacted with the oxidizing agent may range from about 20° C. to about 95° C. The temperature may be selected to result in a relatively rapid but controlled reaction rate without causing the aqueous solution containing the oxidizing agent to boil. In an aspect, the oxidizing agent is NaClO in aqueous solution at a concentration of about 12% available chlorine, and the NaClO solution is contacted with the mixture at a temperature of about 40° C. In another aspect, the oxidizing solution may be maintained at a temperature of about 40° C. until the mixture containing the osmium and other metals is completely dissolved, and the oxidizing solution may be maintained at a temperature of about 90° C. for the remainder of the reaction. As the mixture containing the osmium and other metals dissolves within the oxidizing solution, the solution may take on a yellowish colored appearance as the osmium is oxidized within the oxidizing solution. As the dissolved OsO4 is distilled out of the oxidizing solution, the oxidizing solution takes on a transparent white appearance. In an aspect, a colored oxidizing solution containing the dissolved mixture of metals may be maintained at a temperature of about 90° C. until the oxidizing solution again takes on a transparent appearance, indicating that essentially all osmium in the solution has been oxidized and distilled away. In various aspects, the mixture is contacted with the oxidizing agent in an impinger, shown schematically in FIG. 2. In this aspect, the impinger 200 includes a closed vessel 202 containing the oxidizing solution 204. In use, the mixture 216 and oxidizing solution 204 are placed into the impinger 200. A gas is introduced into the impinger 200 via an impinger inlet 208, which bubbles through the oxidizing solution 204, causing the mixing of the contents of the impinger 200. As the oxidizing solution 204 contacts the mixture 216, the osmium within the mixture may be converted into OsO4 vapor. This OsO4 vapor may form into bubbles 206 that may float to the surface 210 of the oxidizing solution 204, where the OsO4 vapor is released into the headspace 212 of the impinger 200. Driven by the building pressure of the gases introduced into the impinger 200, the gases within the headspace 212, which may include OsO4 vapor, exit the impinger 200 via the impinger exit 214. The gas that is introduced into the impinger 200 may be any gas that does not interfere with the reaction between the mixture and the oxidizing solution including, but not limited to: air, oxygen, nitrogen, noble gases, and combinations thereof. In an aspect, the gas is selected to be a non-toxic gas capable of being vented to the environment after the OsO4 has been extracted as described herein below. Non-limiting examples of gases suitable for introduction into the impinger 200 include nitrogen, argon, helium, oxygen, and combinations thereof. In another aspect, the gas introduced into the impinger 200 is nitrogen. The gas may be introduced at any suitable rate that results in the vigorous mixing of the mixture and the oxidizing solution, so long as the rate does force the venting of any oxidizing solution through the impinger exit 214. The source of the gas may be the atmosphere outside of the impinger 200, or the gas source may be a pressurized tank or any other existing gas source. In other aspects, the rate of introduction of the gas into the impinger 200 may be limited by the design and performance capabilities of the impinger 200, as well as the design and performance capability of other components downstream of the impinger 200, such as the trapping vessel, described in detail herein below. In various embodiments, the OsO4 vapor may be bubbled through a KOH trapping solution to form an amount of K2[OsO4(OH)2] dissolved within the trapping solution. In an aspect, the KOH trapping solution comprises an aqueous solution of KOH at a concentration ranging from about 10% to about 50% w/v. The concentration of the KOH may be selected based on any one or more of at least several factors including, but not limited to: the rate and concentration at which the OsO4 vapor is bubbled through the KOH trapping solution, the reaction conditions such as temperature of the KOH trapping solution, and the solubility of the KOH in the aqueous solvent. In an aspect, the KOH trapping solution is an aqueous solution of KOH at a concentration of about 25% w/v at a temperature of less than about 5° C. In another embodiment, a trap vessel containing the KOH trapping solution is situated within an ice bath. Without being bound to any particular theory, the reduction of the OsO4 vapor within the KOH trapping solution is an exothermic reaction. Cooling the KOH trapping solution to a lower temperature using an ice bath maintains the KOH trapping solution at a higher solubility for the OsO4 vapor. FIG. 3 is a schematic illustration of a trap vessel 300 in one aspect. In this aspect, the trap vessel 300 may be any closed vessel 302 containing an amount of KOH trapping solution 304. The gas exiting the impinger 200, which may contain OsO4 vapor, is directed into the trap inlet 308, which bubbles the gas through the KOH trapping solution 304. The gas bubbles impart mixing of the gas and the KOH trapping solution 304, as well as promote the intimate contact of the OsO4 vapor with the KOH within the trapping solution 304. The bubbles 306 exit the surface 310 of the trapping solution 304 into the trap headspace 312. Gas within the trap headspace 312 is forced from the trap vessel 300 by the continuous introduction of additional gases through the trap inlet 308. The headspace gas, which may contain a lower concentration of OsO4 vapor than the gas entering the trap inlet 308, may exit the trap vessel 300 through the trap exit 314. The concentration of OsO4 vapor exiting the trap vessel 300 through the trap exit 314 may be less than about 20% of the concentration of the OsO4 vapor entering the vessel 300 through the trap inlet 308. The degree of reduction of OsO4 vapor concentration may be governed by the effectiveness the reaction between the OsO4 and the KOH in the trapping solution 312. The effectiveness of the reaction may depend on any one or more of at least several factors including, but not limited, to: the concentration of KOH in the trapping solution 304, the reaction conditions such as temperature and pressure within the trap vessel 300, the rate of gas introduction into the trap vessel 300, and the design of the trap vessel 300. In other aspects, the concentration of OsO4 vapor exiting the trap vessel 300 through the trap exit 314 may be less than about 10%, less than about 5%, less than about 1%, or less than about 0.1% of the concentration of the OsO4 vapor entering the vessel 300 through the trap inlet 308. In another aspect, the trap vessel 300 may include a two-in-one pipette 400, shown schematically in FIG. 4. The two-in-one pipette 400 includes an inner pipette 402 situated within an outer pipette 404. The gas entering the trap vessel 300A via the pipette inlet 410 is released through the open tip 406 of the inner pipette 402. The released gas from the open tip 406 bubbles through a small amount of KOH trapping solution 304 contained within the outer pipette 404. Without being tied to any particular theory, the introduction of the gas through the two-in-one pipette 400 imparts more intimate and sustained contact between the bubbles 414 and the trapping solution 304. The bubbles 414 may be distorted into larger surface areas, shaped by capillary forces imparted by the outer surface of the inner pipette 402 and the inner surface of the outer pipette 404. Further, these capillary forces may impede the free movement of the bubbles to the surface of the trapping solution 304, resulting in a sustained time of contact between the gas bubbles 414 and the trapping solution 304. This combination of factors may result in more efficient and extensive conversion of the OsO4 vapor into K2[OsO4(OH)2] within the trapping solution 304. The two-in-one pipette 400 may be immersed in a liquid 408 to facilitate heat transfer from the pipette 400 to a heat sink such as an ice bath (not shown). The bubbles 414 may be released into the headspace 416 of the outer pipette 404 and may exit the two-in-one pipette 400 via a vapor outlet 412. The rate at which gas from the impinger 200 is introduced into the trapping vessel 300 may depend on any one or more of at least several factors including, but not limited to: the rate at which gases exit the impinger 200, as well as the sizing and design of the trap vessel 300. The rate at which gases exit the impinger 200 may be governed by the rate at which gas is introduced into the impinger 200 as well as the rate of production of OsO4 vapor within the oxidizing solution 204. The trap vessel 300 may be designed to have a volume that is larger relative to the impinger 200 in order to impart a lower flow velocity through the trap vessel 300. Alternatively, the gases exiting the impinger 200 may be directed into two or more trap vessels 300 attached in parallel, resulting in a larger overall trap vessel volume relative to the impinger 200. In another aspect, in order to trap a higher proportion of the OsO4 vapor released by the impinger 200, two or more trap vessels may be connected in series to the impinger exit 214. FIG. 5 is a schematic illustration showing a series of trap vessels 300A, 300B, and 300C connected in series to the impinger 200A. If two or more trap vessels 300 are connected to the impinger exit 510 in parallel as described herein above, additional trap vessels may be connected in series to each of the trap vessels connected directly to the impinger. In yet another aspect, a final trapping filter 534, including but not limited to an adsorbent filter such as an activated charcoal filter, may be connected to the trap exit 532 of each final trap vessel 300C in each series of trap vessels. In various embodiments, the osmium trapped within the dissolved K2[OsO4(OH)2] in the KOH trapping solution may be precipitated and/or encapsulated into a usable form by contacting the dissolved K2[OsO4(OH)2] with a reducing agent to form an Os precipitate. The particular Os precipitate formed depends upon the species of reducing agent contacted with the dissolved K2[OsO4(OH)2]. Non-limiting examples of Os precipitates include Os metal, OsS2, OsO2, and K2[OsO2(OH)4]. Non-limiting examples of suitable species of reducing agents include absolute ethanol, Zn shavings, Al shavings, Mg shavings, NaBH4 and other alkali salts of BH4, NaHS, H2S gas, Na2S2O3, UV light, phosphine ligands, hydrazine, hydroquinone, hydrophosphorous acid, formaldehyde, hydroxylamine, citrate, ascorbic acid, and hydrogen gas. In one aspect, the Os is recovered from the KOH trapping solution by reducing the dissolved K2[OsO4(OH)2] to K2[OsO2(OH)4] crystals by adding an amount of absolute ethanol to the KOH trapping solution. In this embodiment, the concentration of the absolute ethanol added may range from about 1% to about 20% v/v. In another aspect, ethanol is added at a concentration of about 5% v/v. The K2[OsO2(OH)4] crystals may then be harvested for encapsulation. In another aspect, Zn, Mg, or Al metal shavings may be added to the KOH trapping solution, and concentrated HCl may be added to the solution to lower the pH of the solution, resulting in the formation of Os metal. After removal of the shavings, the precipitate may be centrifuged and washed with water to isolate the Os metal. In an additional aspect, Zn and Al shavings may be added to the basic KOH trapping solution to form a precipitate that may include Os metal, OsO2, and combinations thereof. In another additional aspect, NaBH4 may be added to the KOH trapping solution to form an Os metal precipitate or other reduced species of Os such as OsO2. The precipitate may be centrifuged and washed with water to isolate the Os metal. In yet another aspect, NaHS may be added to the KOH trapping solution in order to form an OsS2 precipitate. The OsS2 precipitate may also be formed by bubbling H2S gas through the KOH trapping solution. In addition, Na2S2O3 may be added to either the basic or acidified KOH trapping solution to form an OsS2 precipitate. The OsS2 precipitate may be centrifuged and washed with water to isolate the OsS2 precipitate. In still yet another aspect, the osmium may be encapsulated by drawing an amount of the KOH trapping solution containing the K2[OsO4(OH)2] into a thin vial, followed by dipping the thin vial into an aqueous solution containing NaHS to form an encapsulated OsS2 precipitate within the thin vial. Various embodiments provide a system for the separation of osmium from a mixture including the osmium and at least one other metal. FIG. 5 is a schematic representation of an osmium separation system 500. The system 500 includes an impinger 200A and a series of trapping vessels 300A-300C that may include a first trapping vessel 300A in an ice bath 502, a second trapping vessel 300B, a third trapping vessel 300C, and an activated charcoal filter 534. The impinger exit 510 may be connected directly to the first trap inlet 512, the first trap exit 520 may be connected directly to the second trap inlet 522, the second trap exit 526 may be connected directly to the third trap inlet 528, and the third trap exit may be connected directly to the activated charcoal filter 534. The filter exit 536 may vent directly to the atmosphere. The elements of the system 500 form a continuous hydraulic path from the impinger inlet 504 to the filter exit 536, and the gases are impelled from the impinger 200A to the first trap vessel 300A due to the pressurization of the impinger 200A caused by the continuous introduction of gas into the impinger inlet 504. In use, a mixture 508 that includes an amount of osmium and at least one other metal may be placed into the impinger 200A along with an amount of oxidizing solution 506. The oxidizing solution 506 may be maintained at about 40° C., and a moderate flow of nitrogen may be introduced into the impinger inlet 504, causing the agitation of the oxidizing solution 506 as well as the fluids within the downstream trap vessels 300A-300C. As the mixture 508 dissolves into the oxidizing solution 506, the oxidizing solution 506 may take on a colored appearance. Once the mixture has completely dissolved within the oxidizing solution 506, the temperature of the oxidizing solution 506 solution may be maintained at about 90° C. until all of the osmium in the oxidizing solution 506 has been oxidized into OsO4 vapor. In an aspect, the color of the oxidizing solution 506 may change from colored to clear to indicate the oxidation of all dissolved osmium in the oxidizing solution 506. The OsO4 vapor formed in the impinger 200A may be carried along with the introduced nitrogen into the first trap vessel 300A. In an embodiment, the first trap vessel 300A includes a two-in-one pipette 516 containing an amount of KOH trapping solution 514. An amount of OsO4 vapor bubbling through the KOH trapping solution 514 may contact the dissolved KOH, forming dissolved K2[OsO4(OH)2]. The introduced nitrogen, along with any untrapped OsO4 vapor may pass into the second trap vessel 300B, where an amount of OsO4 may be captured within the second KOH trapping solution 524. Similarly, the introduced nitrogen, along with any further untrapped OsO4 vapor, may pass into the third trap vessel 300C, where an amount of OsO4 may be captured within the third KOH trapping solution 530. Any residual OsO4 vapor leaving the third trap vessel 300C may be captured within the activated charcoal filter 534, and essentially osmium-free nitrogen may exit the filter exit 536 to the atmosphere. Once essentially all of the dissolved osmium within the oxidizing solution 506 has been oxidized into OsO4 vapor and bubbled through the trap vessels 300A-300C, the flow of nitrogen gas may be stopped and the dissolved K2[OsO4(OH)2] within the KOH trapping solution 514 may be precipitated into a useable form using any of the methods described herein previously. In an aspect, only dissolved K2[OsO4(OH)2] within the KOH trapping solution 514 from the first trap vessel 300A is precipitated. In another aspect, the trapping solutions 514, 524, and 530 may be combined and the dissolved K2[OsO4(OH)2] within the combined KOH trapping solutions may be precipitated. In an aspect, the dissolved K2[OsO4(OH)2] within the combined KOH trapping solutions may be transferred into a centrifuge tube and combined with an amount of NaHS to form a black OsS2 precipitate. The OsS2 precipitate may be further treated after the addition of the NaHS, or the centrifuge tube may be left as long as about 6 hours to about 24 hours to ensure that the K2[OsO4(OH)2] has completely reacted with the NaHS. The OsS2 precipitate in the centrifuge may be agitated with water, centrifuged, and the water supernate may be discarded. This washing process may be repeated two or more times to ensure that any remaining impurities are rinsed from the OsS2 precipitate. The water-rinsed OsS2 precipitate may be additionally rinsed with a solvent such as acetone and dried at room temperature for about 1 hour. The osmium separation methods of various aspects may be applied in a variety of different contexts. As described herein above, one aspect of the method may be used to separate osmium isotopes or radioisotopes from a mixture including the osmium and at least one other metal. The resulting chemically pure osmium may be an enriched osmium target used for the production of radioisotopes, or the chemically pure osmium may be Os-191 or other Os radioisotopes used in a variety of applications including, but not limited to: a radiotracer composition, a radiotracer source, or as an ingredient in a therapeutic composition. In another aspect, the osmium separation method may be used to separate an osmium impurity from a mixture including the osmium impurity and at least one other metal. A flowchart of this method 600 is illustrated in FIG. 6. In this aspect, the material source that includes the osmium impurity is introduced into the impinger at step 602. The mixture is dissolved and oxidized in the impinger at step 604, as described herein previously. The volatile OsO4 vapor resulting from the oxidation of the osmium impurities may be trapped within the KOH trapping solution at step 606. This trapped osmium may be treated to reclaim the osmium or discarded, depending on the intended use of the osmium impurities. In this aspect, the mixture that includes at least one other metal remains dissolved in the oxidizing solution in the impinger. In an aspect, the at least one other metal may be precipitated out of the oxidizing solution by contacting the oxidizing solution with a reducing agent at step 608 to produce an osmium-free mixture. The selection of reducing agent may depend on one or more of at least several factors including, but not limited to, the particular species of dissolved metal within the oxidizing solution. In another aspect, if the oxidizing solution contains an amount of Re-186, the Re-186 may be separated from the oxidizing solution by contacting the oxidizing solution with an alumina chromatographic column, and eluting the Re-186 using a saline solution. In another aspect, a dissolved metal may be isolated from the oxidizing solution using electroplating methods. Having described the invention in detail, it will be apparent that modifications and variations are possible. Those of skill in the art should, in light of the present disclosure, appreciate that many changes could be made in the specific embodiments that are disclosed and still obtain a like or similar result without departing from the spirit and scope of the invention, therefore all matter set forth is to be interpreted as illustrative and not in a limiting sense. |
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044973497 | abstract | An apparatus for dispensing a gas contained in a sealed ampule, typically a radioactive gas dissolved in a saline solution. A housing with a cavity for receiving the ampule, the housing having a lead shield surrounded by a steel sleeve, and a cap for closing the cavity in the shield. First and second flow paths are provided in the housing from the exterior of the housing to a seat which engages the ampule output line. A syringe is attached on the outside of the housing and connects with the first flow path. A hollow needle is positioned in the second flow path with one end entering the ampule outlet line for breaking the ampule seal. A valve on the outer end of the needle is normally closed so that when the needle breaks the seal, liquid from the syringe is drawn into the ampule with the gas dissolving in the liquid. Then the gas-liquid solution is withdrawn through the needle and valve. |
abstract | A fuel assembly for a boiling water reactor having fuel rods, two or three water rods, a tie plate, spacers, a handle, and a joint arrangement. The joint arrangement is configured to transfer a vertical lifting force from the handle to the water rods. The joint arrangement includes a balancing element arranged between the water rods and the handle. The joint arrangement includes a first joint arranged between the balancing element and the handle and a set of second joints arranged between a respective one of said water rods and said balancing element. The first joint and the set of second joints are configured to allow a rotational movement of said balancing element in relation to said handle as well as in relation to said water rods in order to balance lifting forces in the water rods. |
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claims | 1. A method, comprising:identifying a target in a patient;determining a respiratory motion of the target in the patient;determining a pulsating motion of the target in the patient separately from the determining of the respiratory motion, wherein determining the respiratory motion comprises:detecting a first motion of the target in the patient caused by a combined respiratory and pulsating motion of the patient while the patient breathes during a first period using one or more radiographic images of the target generated during the first period;detecting a second motion of the target in the patient caused substantially only by the pulsating motion while the patient holds a breath during a second period using one or more radiographic images of the target generated during the second period; andcalculating the respiratory motion of the target in the patient using the detected first and second motions; anddirecting a radiosurgical beam, from a radiosurgical beam source, to the target in the patient based on the determining of the pulsating motion and based on the determining of the respiratory motion. 2. The method of claim 1, wherein the target is a heart and the pulsation motion is due to a heartbeat of the heart, and wherein directing the radiosurgical beam to the target comprises creating a lesion in the heart to inhibit atrial fibrillation. 3. The method of claim 1, wherein said directing the radiosurgical beam comprises:determining a relative position of the radiosurgical beam source and the target based on the determining of the respiratory motion and the determining of the pulsating motion; andcompensating for movement of the target, due to the respiratory motion and the pulsating motion of the patient, in the directing of the radiosurgical beam based on the determining of the respiratory motion and the determining of the pulsating motion. 4. The method of claim 3, wherein determining the respiratory motion comprises detecting a respiratory cycle and determining the pulsating motion comprises detecting a pulsating cycle, and wherein the pulsating cycle is detected separately from the detecting of the respiratory cycle. 5. The method of claim 3, wherein detecting the first motion of the target in the patient caused by the combined respiratory and pulsating motion comprises:generating the one or more radiographic images of the target during the first period; andcombining information of the pulsating cycle and the respiratory cycle with the one or more radiographic images of the target during the first period to form the first motion. 6. The method of claim 5, wherein detecting the second motion of the target in the patient caused substantially only by the pulsating motion comprises:generating the one or more radiographic images of the target while the patient holds the breath during the second period; andcombining information of the pulsating cycle with the one or more radiographic images of the target during the second period to form the second motion. 7. The method of claim 6, further comprising:computing a first correction factor from the respiratory motion; andseparately computing a second correction factor from the second motion. 8. The method of claim 7, further comprising generating a command signal to adjust the relative position of the radiosurgical beam and the target in compensating for the movement of the target, and wherein the compensating for the movement of the target comprises generating a correction to the command signal using the first and second correction factors. 9. The method of claim 8, wherein the directing of the radiosurgical beam, from the radiosurgical beam source, to the target in the patient comprises directing the radiosurgical beam using the using the command signal. 10. The method of claim 8, wherein the first correction factor, when applied to the command signal, is effective to adjust the relative position of the radiosurgical beam source and the target to account for the respiratory motion caused substantially only by the respiratory motion, and wherein the second factor, when applied to the command signal, is effective to adjust the relative position of the radiosurgical beam source and the target to account for the second motion caused substantially only by the pulsating motion. 11. The method of claim 8, wherein generating one or more radiographic images of the target during the first respiratory cycle comprises:generating a first operational scan of the target and an internal marker at a first position;generating a second operational scan of the target and the internal marker at a second position;generating a first model of internal motion of the internal marker using the first and second positions;generating a second model of external motion using the detected second motion and the calculated respiratory motion; andcorrelating the external motion and the internal motion using the first and second models. 12. The method of claim 11, wherein generating the first model of the internal motion comprises fitting a first curve to the first and second positions of the internal marker, wherein generating the second model of external motion comprises measuring external motion using external markers to generate a plurality of positions of the external markers and fitting a second curve to the plurality of positions of the external markers, and wherein correlating the external and internal motions comprises comparing the first and second curves. 13. The method of claim 8, wherein generating the command signal to adjust the relative position of the radiosurgical beam source and the target comprises adjusting the relative position of a radiosurgical beam source and the target using a robotic positioning system. 14. The method of claim 7, further comprising:generating a command signal to adjust a relative position of the radiosurgical beam and the target in compensating for the first motion; andgenerating a combined correction factor using the first and second correction factors, and wherein compensating for the first motion of the target comprises generating a correction to the command signal using the combined correction factor. 15. The method of claim 7, wherein computing the first and second correction factors comprises:generating a first signal representative of the respiratory motion;generating a second signal representative of the second motion;filtering the first and second signals to cancel out undesired frequency components;computing the first correction factor from the filtered first signal; andseparately computing the second correction factor from the filtered second signal. 16. The method of claim 1, wherein the pulsating motion is cardiac pumping motion, and wherein calculating the respiratory motion of the target comprises:establishing a look-up table of positional data for the first motion;establishing a look-up table of cardiac motion data for the second motion; andestablishing a look-up table of respiratory motion data for the calculated respiratory motion by subtracting the cardiac motion data from the positional data for the first motion. 17. The method of claim 16, further comprising:generating a first signal representative of the respiratory motion of the patient from the look-up table of respiratory motion data;generating a second signal representative of the second motion of the patient from the look-up table of cardiac motion data;computing a first correction factor from the first signal;separately computing a second correction factor from the second signal; andgenerating a command signal to adjust the relative position of the radiosurgical beam and the target to compensate for the movement of the target; andgenerating a correction to the command signal using the first and second correction factors. 18. The method of claim 7, wherein computing the first and second correction factors comprises digitally comparing the one or more radiographic images with a pre-operative scan. 19. The method of claim 18, wherein digitally comparing the one or more radiographic images with the pre-operative scan comprises:generating one or more digitally reconstructed radiographs (DRRs), using the pre-operative scan together with the one or more radiographic images; andcomputing an amount of movement of the target needed to register the one or more DRRs with the one or more radiographic images. 20. The method of claim 8, wherein generating the correction to the command signal comprises:extrapolating the detected first motion of the target into a complete cycle; andsynchronizing the command signal to adjust the relative position of the radiosurgical beam source and the target with the extrapolated motion of the target. 21. The method of claim 1, further comprising:computing a first correction factor from the second motion; andproviding a static correction factor representative of peaks of the respiratory cycle of the calculated respiratory motion caused substantially only by the respiratory motion of the patient. 22. The method of claim 21, further comprising:generating a command signal to adjust the relative position of the radiosurgical beam and the target to compensate for the movement of the target; andgenerating a correction to the command signal using the first correction factor and the static correction factor. 23. A system, comprising:an imaging device configured to acquire an image of a target in a patient;a pulsation measurement device configured to detect a pulsating cycle of the target in the patient;a radiosurgical beam source configured to direct a radiosurgical beam, from the radiosurgical beam source, to the target in the patient; anda controller coupled to the radiosurgical beam source and the pulsation measurement device, wherein the controller is configured to identify the target in the image, to receive a signal from the pulsation measurement device representative of the pulsating cycle, and to determine a pulsating motion of the target in the patient based on the pulsating cycle, wherein the controller is configured to detect a first motion of the target caused by a combined respiratory and pulsating motion while the patient breathes during a first period using a plurality of images of the target acquired during the first period, and a second motion of the target caused substantially only by the pulsating motion during a second period using a plurality of images of the target acquired during the second period, and wherein the controller is configured to calculate a respiratory motion of the target in the patient using the first and second motions, and wherein the controller is configured to position the radiosurgical beam source to direct the radiosurgical beam to compensate for the first motion of the target caused by the combined respiratory and pulsating motion using the second motion and the calculated respiratory motion. 24. The system of claim 23, further comprising a breathing sensor coupled to the controller, wherein the breathing sensor is configured to detect a respiratory cycle of the patient, and wherein the controller is configured to receive a signal from the breathing sensor representative of the respiratory cycle, wherein the controller is configured to detect the first motion based on the respiratory cycle. 25. The system of claim 24, wherein the target is a heart and the pulsation motion is due to a heartbeat of the heart, and wherein the controller is configured to position the radiosurgical beam source to direct the radiosurgical beam to the heart to create a lesion in the heart to inhibit atrial fibrillation. 26. The system of claim 24, further comprising a robotic positioning system coupled to the controller and the radiosurgical beam source, wherein the controller is further configured to compute a first correction factor from the calculated respiratory motion, and separately compute a second correction factor from the second motion, wherein the controller is configured to generate a command signal to adjust a relative position of the radiosurgical beam and the target to compensate for the first motion using the robotic positioning system, wherein the controller is configured to generate a correction to the command signal using the first and second correction factors, and wherein the robotic positioning system is configured to receive the command signal from the controller to adjust the relative position of the radiosurgical beam source and the target. 27. The system of claim 26, wherein the robotic positioning system comprises six degrees of freedom to adjust the relative position of the radiosurgical beam source and the target. 28. The system of claim 24, further comprising:an imaging system coupled to the controller, wherein the imaging system is configured to generate the plurality of images of the target during the first period, and to generate the plurality of images of the target during the second period, andwherein the controller is configured to record information of the respiratory cycle detected by the breathing sensor and information of the pulsating cycle detected by the pulsation measurement device, during the first period, to receive the plurality of images of the target of the first period, and to combine the recorded information of the respiratory cycle and of the pulsating cycle during the first respiratory cycle with the plurality of images of the target of the first period to generate the first motion, andwherein the controller is configured to record information of the pulsating cycle detected by the pulsation measurement device during the second period, to receive the plurality of images of the target of the second period, and to combine the recorded information of the pulsating cycle during the second period with the plurality of images of the target of the second respiratory cycle to generate the second motion. 29. The system of claim 28, further comprising a signal processor coupled to the controller, the signal processor to receive the detected respiratory cycle, the detected pulsating cycle, the plurality of images of the target during the first period, and the plurality of images during the second period,wherein the signal processor is configured to combine information of the pulsating cycle and the respiratory cycle during the first period with the plurality of images of the target during the first period to form the first motion,wherein the signal processor is configured to combine information of the pulsating cycle during the second period with the plurality of images of the target during the second period to form the second motion,wherein the signal processor is configured to calculate the respiratory motion of the target using the first and second motions, andwherein the signal processor is configured to filter the second motion and the respiratory motion to cancel out undesired frequency components. 30. The system of claim 29, wherein the controller further comprises a storage unit to store image data of the plurality of images of the target. 31. The system of claim 29, further comprising:a pre-operative scanner coupled to the signal processor, the pre-operative scanner to generate a pre-operative scan for one or more of the plurality of images of the target, andwherein the imaging system is configured to generate one or more operative scans for one or more of the plurality of images of the target, and wherein the controller is configured to digitally compare one or more operative scans with the pre-operative scan. 32. The system of claim 23, wherein the pulsation measurement device is configured to detect at least one of pulsating arteries or cardiac pumping motion of the patient. 33. The system of claim 24, wherein the breathing sensor is at least one of an infrared tracking system, a force sensor, an air flow meter, a strain gauge, or a laser range sensor. 34. The system of claim 23, wherein the pulsating measurement device is at least one of a strain gauge, electrocardiograph, or a heart beat monitor. 35. The system of claim 24, wherein the pulsating cycle is a heartbeat cycle, wherein the pulsation motion is due to cardiac pumping motion, and wherein the controller comprises:a look-up table of positional data for the first motion;a look-up table of cardiac motion data for the second motion; anda look-up table of respiratory motion data for the calculated respiratory motion, and wherein the controller is configured to calculate the respiratory motion data by subtracting the cardiac motion data from the positional data for the first motion. |
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abstract | The present invention relates to a control rod blade for a boiling water reactor. The control rod blade (2) comprises a free edge portion with a recess (7) which comprises a plurality of outlets, arranged in a row, for channels (3), which are arranged to receive an absorber material (10) and a cover element (4) arranged to be attached along at least a section of the edge portion. The cover element (4) comprises a cover portion (12) arranged to seal the opening of the recess (7) and a support portion (13) arranged to, in a mounted state, abut a bottom surface in the recess (7) and to allow the formation of at least a passage (16) between the outlets of the channel (3) in the recess (7). |
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047019409 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Embodiments of the present invention will be described hereinunder with reference to the accompanying drawings. EMBODIMENT 1 FIG. 1 illustrates Embodiment 1 of a patterning process according to the present invention; FIGS. 2a and 2b are sectional views of diffraction gratings, FIG. 2a being a sectional view of a conventional diffraction grating formed by an electron impact X-ray tube and FIG. 2b being a sectional view of a diffraction grating formed by an X-ray source according to the present invention; FIGS. 3a and 3b illustrate an angular dependence of pattern broadening width, FIG. 3a illustrating a relation between a line pattern and an electric vector and FIG. 3b illustrating a relation between an angle .alpha. of an electric vector relative to a line pattern and a broadening width; FIG. 4 illustrates a mask pattern in Embodiment 2 of the present invention; FIG. 5 is a sectional view of a line pattern in Embodiment 3 of the present invention; and FIGS. 6a and 6b illustrate an angular dependence of the quantity of photoelectrons emitted in the above embodiment, FIG. 6a illustrating a relation between a line pattern and an electric vector and FIG. 6b illustrating a relation between an angle .beta. of an electric vector relative to a line pattern and the quantity of photoelectrons emitted. In FIG. 1, a linearly polarized X-ray 6 was directed through a mask 8 to a resist 9 coated on a substrate 10. As the linearly polarized X-ray there was used a synchrotron radiation (hereinafter referred to as "SR"). The electron energy at a storage ring is 1 GeV. An electric vector direction 7 was fixed horizontally. In the mask 8, a membrance comprises boron nitride (BN) (3 .mu.m in film thickness) and polyimide (PIQ) (1 .mu.m in film thickness); an absorber is formed of gold (Au) (1 .mu.m in film thickness); and a protection layer is formed of PIQ (1 .mu.m in film thickness). The pattern is a diffraction grating of 1 .mu.m pitch, the pitch direction intersecting the electric vector direction 7 of the X-ray perpendicularly. The resist 9 has a 1 .mu.m thick coating of a micro resist for short wavelength (MRS) (trade name: RD20000N, a product of Hitachi Chemical Co., Ltd.). The substrate 10 is a silicon substrate. After irradiation of SR at a rate of 65 mJ/cm.sup.2, a pattern was formed by development. In the case of using an unpolarized X-ray source such as an electron impact X-ray tube, there is obtained such a sectional shape of line pattern as shown in FIG. 2a in which the bottom of a resist 11 broadens about 0.1 .mu.m. On the other hand, according to this embodiment of the present invention there is obtained a good sectional shape of a resist 12 as shown in FIG. 2b. FIG. 3 shows how the resist broadening width changes with change of the angle .alpha. between a line pattern 13 of the diffraction grating and the electric vector direction 7, from which it is seen that the broadening with decreases as the angle .alpha. becomes smaller. Also when magnetic materials (e.g. YIG (yhtrium iron garnet), GGG (gallium gadolinium iron garnet)) were used for the substrate, there were obtained the same results. EMBODIMENT 2 In this embodiment there was used the same optical system as in the above first embodiment except that such a shape of a mask pattern as shown in FIG. 4 was used. This pattern corresponds to one used in the production of a gate of MOS transistor. It is necessary that the pattern formation be effected accurately on both side portions of a projection 14 of the pattern. Therefore, the longitudinal direction of the pattern and the direction of the electric vector were brought into coincidence with each other and in this state there was performed exposure followed by development. The pattern thus formed was good on both side portions of the projection 14 although broadening occured at both end sides 15 and 16 of the pattern. EMBODIMENT 3 In the above two embodiments the influence of photoelectrons in pattern formation was observed, while in this Embodiment 3 photoelectrons per se created at the time of pattern formation are counted. More specifically, SR was applied to a line resist having such a sectional shape as shown in FIG. 5 and the quantity of photoelectrons emitted was measured. The line pattern was formed by laminating a 1 .mu.m thick resist MRS 18 onto a 0.2 .mu.m thick tungsten (W) 17 with a width of 2 .mu.m and a length of 2 mm formed on the silicon substrate 10. The quantity of photoelectrons emited from the tungsten of a high absorption coefficient to the right and left of the line pattern corresponds to the cause of the pattern deterioration in the foregoing Embodiments 1 and 2. FIG. 6 shows how the quantity of photoelectrons change with change in the electric vector direction 7 of SR. If the angle between the line pattern 19 and the electric vector direction 7 is .beta., the quantity of photoelectrons emitted decreases as the angle .beta. becomes smaller. According to the patterning process of the present invention, as set forth hereinabove, there is used a linearly polarized X-ray to restrict the emitting direction of photoelectrons created upon irradiation of X-ray and the direction of the electric vector of the linearly polarized X-ray is adjusted relative to the pattern shape to reduce the influence of photoelectrons, whereby a good pattern can be formed. Particularly, in a one-dimensional shape pattern (Embodiment 1), the foregoing influence of photoelectrons can be eliminated nearly completely, and in a two-dimensional shape pattern (Embodiment 2), an effective pattern can be formed by making the line direction and the electric vector direction intersect each other while taking note of only the portion where a highly accurate pattern formation is required. The present invention is suitable for production of a semiconductor device and a magnetic bubble memory device, for example. |
051280930 | summary | TECHNICAL FIELD The present invention relates generally to nuclear reactors, and, more specifically, to a hydraulic system for providing pressurized fluid to control rod drives thereof. BACKGROUND ART A conventional boiling water reactor includes a plurality of control rod drives (CRDs) which are effective for selectively inserting and withdrawing control rods into and from the core of the reactor. In one embodiment, the CRDs include a piston against which is provided a pressurized hydraulic fluid, such as water, for providing the force for inserting the control rods during a scram operation. The pressurized fluid is provided to the CRDs by conventional hydraulic control units (HCUs) for selectively controlling the operation of the CRDs. In order to ensure effective reactor shutdown during scram, the HCUs include conventional scram accumulators which are connected to a common charging water header and are charged with the pressurized fluid for storing the pressurized fluid for use in the scram operation. The charging water header is connected to a relatively high power, for example about 500 shaft horsepower, CRD pump which is run continuously at a high speed sufficient to generate a discharge pressure great enough to maintain the scram accumulators continuously charged to a pressure sufficient to scram the control rods at their required speeds. During normal plant operation, when the scram accumulators are full, there is no flow through the charging water header. However, the CRD pump discharge pressure must remain high enough to maintain the charging header at a pressure sufficient to keep the scram accumulators fully pressurized. This discharge pressure is significantly higher than that required for the CRD pump to perform its other normal operating function of delivering purge water on a continuous basis to a conventional fine motion control rod drive (FMCRD) found in the CRD. The purge water is provided from the CRD pump to the respective CRDs through a purge water header that is arranged in parallel with the charging water header. Thus, a significant economic penalty is incurred in this design by having to provide electric power to drive the high power CRD pump continuously at a speed and discharge pressure greater than those required for delivering the normal purge water flow to the CRDs. OBJECT OF THE INVENTION Accordingly, one object of the present invention is to provide a new and improved control rod drive hydraulic system in which the CRD pump may be operated for a substantial amount of time at a relatively low speed and low discharge pressure while maintaining the scram accumulators at an acceptable pressure. DISCLOSURE OF INVENTION A hydraulic system for a control rod drive (CRD) includes a variable output-pressure CRD pump operable in a charging mode for providing pressurized fluid at a charging pressure, and in a normal mode for providing the pressurized fluid at a purge pressure, less than the charging pressure. Charging and purge lines are disposed in parallel flow between the CRD pump and the CRD. A hydraulic control unit is disposed in flow communication in the charging line and includes a scram accumulator. An isolation valve is provided in the charging line between the CRD pump and the scram accumulator. Control means are operatively connected to the CRD pump and the isolation valve and are effective for opening the isolation valve and operating the CRD pump in a charging mode for providing the pressurized fluid at the charging pressure through the charging line to charge the scram accumulator, and closing the isolation valve and operating the CRD pump in a normal mode for providing to the CRD through the purge line the pressurized fluid at a purge pressure lower than the charging pressure. |
abstract | An X-ray diffraction apparatus having a solar slit, and a method for preventing the diffraction image on a detector from spreading in the in-plane direction even when an X-ray irradiation region spreads over the sample surface due to measurement by GIXD, thereby allowing for measurement with a short measurement time and a high resolution. The soller slit 100 includes a plurality of metallic thin plates 110, each being perpendicular to the bottom surface, which are arcuately arranged with a predetermined angular interval between each other so as to pass X-rays in a radiating direction from a particular focus, the soller slit being provided to be used at a position through which X-rays diffracted on a sample surface pass, the particular focus being the center of a goniometer circle, the X-rays being irradiated on a sample at an angle for GIXD. |
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claims | 1. A method of performing a medical procedure, said method comprising:providing a radiation-shielding cubicle having an interior defining a medical personnel region and including a first wall having an opening therein;locating the cubicle with respect to an x-ray table so a portion of the x-ray table extends through the opening into the interior of the cubicle; andseparating medical personnel from an x-ray emitter disposed outside of the cubicle using the first wall to shield the medical personnel from radiation emitted by the x-ray emitter. 2. A method in accordance with claim 1 further comprising joining the x-ray table to the cubicle using a radiation-shielding flexible interface. 3. A method in accordance with claim 2 wherein said joining the x-ray table to the cubicle using a radiation-shielding flexible interface comprises joining the x-ray table to the first wall using the radiation shielding flexible interface. 4. A method in accordance with claim 1 further comprising sealing the opening in the first wall using a flexible radiation-resistant skirt. 5. A method in accordance with claim 1 further comprising circumferentially joining the x-ray table to the cubicle using a flexible radiation-resistant skirt. 6. A method in accordance with claim 1 further comprising attaching a radiation-shielding screen to the x-ray table so the radiation-shielding screen covers a portion of a patient supported by the x-ray table and covers a portion of a top surface of the x-ray table. 7. A method in accordance with claim 6 further comprising joining the first wall to the radiation-shielding screen using a flexible radiation-resistant skirt. 8. A method in accordance with claim 6 wherein the radiation-shielding screen has at least one port for facilitating access to the patient, said method further comprising:inserting procedural equipment through the port to access the patient with the procedural equipment; andperforming a medical procedure on the patient using the procedural equipment. 9. A method in accordance with claim 8 further comprising positioning a cloak over the port to create a substantially radiation-resistant seal over the port and around the procedural equipment. 10. A method in accordance with claim 6 wherein the radiation-shielding screen has at least one port for facilitating at least one of connection and access to controls for at least one of the x-ray table, the x-ray emitter, and a catheterization system monitor, said method further comprising accessing the controls using the port to control at least one of the x-ray table, the x-ray emitter, and the catheterization system monitor. 11. A method in accordance with claim 1 further comprising:detecting radiation levels within the radiation-shielding cubicle; andterminating operation of the x-ray emitter when the detected radiation levels are above a predetermined level. 12. A method in accordance with claim 1 further comprising monitoring portions of the patient located outside the radiation-shielding cubicle from inside the radiation-shielding cubicle using a video camera. 13. A method in accordance with claim 1 further comprising communicating with the patient from inside the radiation-shielding cubicle using a two-way microphone system. |
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060692901 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The invention comprises a reactant alkaline metal alloy composition including significant quantities of magnesium and/or lithium to react with halogenated hydrocarbons, and particularly chlorinated hydrocarbons, to produce magnesium chloride and/or lithium chloride salt. One form of the invention comprises an alloy including between 10% to 70% magnesium, 10% to 70% aluminum, 1% to 25% zinc, 1% to 25% calcium, and 1% to 25% copper. Within this range, the preferred reactant alloy includes 50% magnesium, 40% aluminum, 4% zinc, 4% calcium, and 2% copper. As used in this disclosure and the appended claims, all percentages are by weight in the total alloy composition. Where the alloy is lithium based without magnesium, the reactant metal alloy composition according to the invention comprises between about 10% to 70% lithium, 10% to 70% aluminum, 1% to 25% zinc, 1% to 25% calcium, and 1% to 25% copper. Within this range, the preferred lithium alloy comprises 40% lithium, 45% aluminum, 1% zinc, 7% calcium, and 7% copper. Another preferred specific alloy compound including lithium comprises approximately 30% lithium, 46% aluminum, 2% zinc, 2% copper, and about 20% calcium. Another preferred form of the invention comprises a reactant metal alloy which includes both magnesium and lithium. This preferred reactant alloy composition comprises between about 1% to 60% magnesium, 10% to 70% lithium, 10% to 70% aluminum, 1% to 25% zinc, 1% to 25% calcium, and 1% to 25% copper. With each reactant metal alloy composition according to the invention, the alloy is heated to a molten state for chemical reaction with the waste material. The temperature of the molten alloy is maintained at least at 770 degrees Celsius to provide the desired reaction with the waste material. Higher temperatures for the molten alloy may also be used within the scope of the invention. A reaction temperature of at least 800 degrees Celsius is suitable for the preferred magnesium alloy composition containing 65% magnesium. The lithium based reactant alloy comprising 40% lithium is preferably maintained at least at 900 degrees Celsius for the reaction process. The preferred alloy comprising 30% lithium is preferably heated to at least 1,000 degrees Celsius for the reaction process. The reactant metal alloy composition and treatment process according to the invention may be used to treat many types of waste materials, including all halogenated hydrocarbons. However, the present alloy and process is particularly useful for chlorinated hydrocarbons such as PCBs (polychlorinated biphenyls). The magnesium alloy composition according to the invention is well suited for treating chlorinated hydrocarbons because the magnesium in the alloy chemically strips chlorine atoms from the waste material compound or compounds to produce magnesium chloride. The recovered magnesium chloride may be used as a feedstock in commercial magnesium refining processes. While the magnesium in the magnesium alloy according to the invention strips chlorine from the waste compounds to produce magnesium chloride salt, other elements in the waste material, such as phosphorous, sulphur, and nitrogen, are also stripped from the carbon atoms in the waste material. Much of this other stripped material forms salts (sulfates, nitrates, phosphates) which separate by gravity at the top of the molten reactant material for recovery by suitable means. The magnesium chloride salt and char both sublime to a gaseous state and are drawn off for separating and recovering the magnesium chloride salt and char. Metals such as chromium in the waste material remain in the molten alloy. The original metals which make up the alloy remain in the molten alloy unless consumed in the formation of salts and small quantities of oxides. The treatment process according to the invention includes charging a reactor container with the desired magnesium and/or lithium based alloy composition and isolating the alloy in the reactor from oxygen. The reactant alloy is then heated by a suitable heating arrangement to at least 770 degrees Celsius. Any remaining oxygen in the reactor vessel quickly reacts with the metal in the alloy to produce metal oxides which appear as slag at the surface of the molten material or sink to the bottom of the reaction container. In the preferred process a layer of pure carbon in the form of graphite is placed at the surface of the molten reactant metal alloy. The graphite layer may be from approximately one-quarter to several inches thick and helps further isolate the molten alloy from any oxygen which may be in the reactor container. Once the molten alloy reaches the desired reaction temperature, the process includes introducing the waste material into the reactant molten alloy. The waste material is preferably introduced below the surface of the molten alloy but may be introduced at the surface of the alloy. The temperature of the metal alloy is maintained at least at a temperature of about 770 degrees Celsius throughout the reaction process. Although heat will commonly need to be added continuously by the heating arrangement in order to maintain the desired reaction temperature, the chemical reaction between the waste material and the metal alloy may be sufficiently exothermal to help maintain the temperature of the molten alloy above the desired 770 degree Celsius level. FIG. 1 shows an apparatus for performing a treatment process embodying the principles of the invention. The apparatus includes a reactor container 2, a recovery/recirculation arrangement 40, a feed arrangement 41, and a heating arrangement 42. The reactor container 2 is preferably built from a suitable metal which will maintain structural integrity at the desired elevated temperatures. However, due to the highly reactive nature of the alloy 10, the reactor container 2 is lined with a ceramic material to prevent the metal of the container from reacting with the reactant alloy. An expendable hook 5 may be placed in the alloy 10 at the termination of the process and, after cooling, may be used to lift the solidified alloy ingot from the reactor container 2. The reactor container 2 also includes a cover 3 which contains a solids loading chute 7. The solids loading chute 7 preferably comprises a double reverse acting door so that when the outer door is open to charge solid waste into the chute, the inner door is closed, but the outer door is closed when the inner door is opened to drop the solid waste into the molten alloy 10. It is desirable to purge most of the air out of the charging chute 7 before admitting waste to minimize metal oxide formation in the anaerobic system of the invention. Of course, the aluminum and magnesium in the molten alloy 10 rapidly reacts with any oxygen under cover 3 to remove oxygen from the gas above the molten alloy bed 10. The heating arrangement 42 includes an induction heater, including an induction heater power supply 6 and induction coils 4 built into the reactor container 2. The coils 4 may be water-cooled and the water may be used to cool the reactant alloy 10 as desired, either during the reaction process or at the completion of the reaction process. The induction heater arrangement 42 includes a heater control 9 with a suitable sensor 9a inside the reactor container 2 for controlling the induction heater and maintaining the temperature of the metal alloy 10 at the desired reactive temperature. Although the induction heating arrangement is illustrated in FIG. 1, any suitable heating arrangement, including fossil fuel burning heater may be used to heat the alloy 10 to the desired temperature. The feed arrangement 41 includes feed tank 12 and feed coil 8. Feed tank 12 contains waste material to be processed. A feed pump 14 pumps the waste material from feed tank 12 to the reactor container 2 through a metering arrangement 15. The metering arrangement ensures that waste material is not transferred to the reactor container at a rate exceeding the capacity of the heater arrangement 42 to maintain the desired reaction temperature. Feed coil 8 is coated on its interior and exterior surfaces or formed from a ceramic material to prevent the coil from reacting with the molten alloy 10 in reactor 2. The outlet end of the coil is positioned well below the surface of the alloy 10. Passing the feed material through coil 8 serves to preheat the feed material prior to introduction to the molten alloy 10. The feed system 41 also preferably includes a gas purging arrangement including a gas storage cylinder 16 for containing a suitable purge gas such as nitrogen. The gas purging arrangement is operated to purge the feed lines and coil 8 of air prior to operation of the system. Gases other than nitrogen may be used to purge the system of oxygen, including flue gases such as gases from a fossil fuel burning heater arrangement. The recovery/recirculation system 40 includes an aqueous scrubber/separator 24, a char/water separator 30, a salt recovery arrangement 31, and a recirculation arrangement 32. Off-gas from the area above the molten alloy 10 in container 2 comprising gaseous magnesium or lithium chloride salt, char, and other gases are drawn off through line 18. Line 18 is preferably made of stainless steel and includes a relief valve 20 to maintain atmospheric pressure on line 18. A water spray nozzle 22 is associated with the scrubber/separator 24 and serves to spray water into the off-gas at the inlet to the scrubber/cyclone separator. The water sprayed into the off-gas causes the char to coalesce while the salt in the off-gas goes into solution in the water. The amount of water supplied through nozzle 22 is preferably controlled with temperature controller 23 to maintain the temperature below about 100 degrees Celsius in the scrubber/separator 24. A char slurry forms in the bottom of the scrubber/separator 24 and is drawn off through valve 26. The slurry comprises char and water with salt in solution. The char slurry is directed to char/water separator 30 which separates out the fine char particles from the water solution and passes the water solution through pump 33 on to salt recovery system 31. Salt recovery system 31 may comprise an evaporative system. Water from salt recovery system 31 may be recycled to nozzle 22. Any gas from separator/scrubber 24 is drawn off through recirculation fan 28 and reintroduced to the area above the molten alloy 10 for recycling through the system. The above described preferred embodiments are intended to illustrate the principles of the invention, but not to limit the scope of the invention. Various other embodiments and modifications to these preferred embodiments may be made by those skilled in the art without departing from the scope of the following claims. |
description | This application is a filing under 35 U.S.C. 371 of international application number PCT/GB02/05604, filed Dec. 11, 2002, which claims priority to application number 0208356.6 filed Apr. 11, 2002, in Great Britain the entire disclosure of which is hereby incorporated by reference The present invention relates to a radioisotope generator of the type commonly used to generate radioisotopes such as metastable technetium-99m (99mTc) and to a method of construction of the radioisotope generator. The diagnosis and/or treatment of disease in nuclear medicine constitute one of the major applications of short-lived radioisotopes. It is estimated that in nuclear medicine over 90% of the diagnostic procedures performed worldwide annually use 99mTc labelled radio-pharmaceuticals. Given the short half-life of radio-pharmaceuticals, it is helpful to have the facility to generate suitable radioisotopes on site. Accordingly, the adoption of portable hospital/clinic size 99mTc generators has greatly increased over the years. Portable radioisotope generators are used to obtain a shorter-lived daughter radioisotope which is the product of radioactive decay of a longer-lived parent radioisotope, usually adsorbed on a bed in an ion exchange column. Conventionally, the radioisotope generator includes shielding around the ion exchange column containing the parent radioisotope along with means for eluting the daughter radioisotope from the column with an eluate, such as saline solution. In use, the eluate is passed through the ion exchange column and the daughter radioisotope is collected in solution with the eluate, to be used as required. In the case of 99mTc, this radioisotope is the principle product of the radioactive decay of 99Mo. Within the generator, conventionally the 99Mo is adsorbed on a bed of aluminium oxide and decays to generate 99mTc. As the 99mTc has a relatively short half-life it establishes a transient equilibrium within the ion exchange column after approximately twenty-four hours. Accordingly, the 99mTc can be eluted daily from the ion exchange column by flushing a solution of chloride ions, i.e. sterile saline solution through the ion exchange column. This prompts an ion exchange reaction, in which the chloride ions displace 99mTc but not 99Mo. In the case of radio-pharmaceuticals, it is highly desirable for the radioisotope generator to be constructed and used under aseptic conditions i.e. there should be no ingress of bacteria into the generator. Moreover, due to the fact that the isotope used in the ion exchange column of the generator is radioactive, and is thereby extremely hazardous if not handled in the correct manner, the radioisotope generator also should be constructed and used under radiologically safe conditions. In trying to ensure adequate radiological protection, some known radioisotope generators have tended to be of a complicated construction incorporating a large number of components and requiring the ion exchange column to be introduced early on in the construction of the generator. This means that there is a lengthy period during construction when the radioisotope generator and those constructing the generator are unnecessarily exposed to radiation. Such complex structures also add to the cost of the generator. It is thus important that the actual construction of the generator is reliable and limits the extent to which the generator and those constructing the generator are exposed to radiation during construction. U.S. Pat. No. 3,946,238 describes a shielded radioisotope generator comprising a cylindrical shielded housing for a central repository. The repository is bound by a removable top cover and side walls and a base which are made from lead and which act as the shielding. Within the repository a bottle is provided which contains an ion exchange column on which 99Mo is absorbed. In this document the construction of the generator is almost completed before the ion exchange column is introduced to the repository. However, the eluate is introduced to/removed from the ion exchange column of the generator via apertures in the walls of the bottle. Thus, although the construction of the generator limits the exposure to radiation during construction, the eluate is introduced and extracted using only a pipette which is highly undesirable as it means that the users of the generator are exposed to radiation each time (i.e. once ever twenty-four hours) the radioisotope is extracted. Moreover, this arrangement provides no means for accurately controlling the flow of eluate. U.S. Pat. No. 3,564,256 describes a radioisotope generator in which the ion exchange column is in a cylindrical holder which is located within two box-shaped elements that are in turn located within appropriate radiation shielding. The holder is closed by rubber plugs at both ends, and the box-shaped elements have passages opposite each of the rubber plugs in which respective needles are located. At the outermost ends of the needles quick-coupling members are provided to enable a syringe vessel containing a saline solution to be connected to one of the needles and to enable a collection vessel to be connected to the other of the two needles. It is self-evident that the box-shaped elements and the radiation shielding must be constructed around the holder containing the ion exchange column. Therefore, throughout the construction of the generator all parts of the generator and those constructing the generator will, of necessity, be exposed to radiation. Furthermore, although reference is made to needles being used to pierce the rubber plugs at each end of the holder, this generator construction provides no means for controlling the penetration of the needles through the plugs. U.S. Pat. No. 4,387,303 describes a radioisotope generator comprising a column having an eluent inlet aperture and an eluate outlet aperture and containing an ion exchange bed with the parent radioisotope. Both the eluent inlet and eluate outlet are in communication with channels in the surrounding shielding for the introduction and removal of eluate to and from the ion exchange column. Although no information is provided with regard to the construction of the generator, it is evident that the shielding must be constructed around the ion exchange column as accurate alignment of the channels in the shielding with the inlet and outlet of the ion exchange column is essential. Thus, here too, during construction all parts of the generator and those constructing the generator will be exposed to radiation from the ion exchange column. U.S. Pat. No. 4,801,047 describes a dispensing device for a radioisotope generator in which the vial containing the saline solution that will be used to flush out the desired radioisotope from the ion exchange column, is mounted in a carrier that is moveable relative to the hollow needle used to pierce the seal of the vial and to extract the saline solution. This construction is described as providing control of the amount of saline solution removed from the vial. The present invention seeks to provide a radioisotope generator and a method of construction of the generator that is simple in construction but which ensures the necessary degree of sterility and radiological protection is provided during construction. In accordance with the present invention, there is provided a device for producing a fluid containing a radioactive constituent, the device comprising a shielded chamber with an opening for receiving an isotope container housing a radioactive isotope; a chamber closure adapted for cooperating with and closing the chamber opening; a first fluid port comprising a first hollow needle projecting into the shielded chamber from the chamber closure for fluid communication with the isotope container; a second fluid port comprising a second hollow needle projecting into the shielded chamber from the closed end of the chamber opposite the chamber closure for fluid communication with the isotope container; first and second compressible buffers mounted so as to surround at least partially the respective first and second hollow needles, each buffer providing an outer surface for contact with opposed ends of the isotope container; and a spacer of a predetermined thickness associated with one or each of the first and second compressible buffers for determining the positioning of the isotope container within the shielded chamber. Preferably, with the chamber closure in place in the chamber opening, the first and second hollow needles are fixed in position at each end of the shielded chamber and ideally the spacer is provided with the second compressible buffer at the closed end of the shielded chamber. In a preferred embodiment the material of the first and second compressible buffers is a semi-open cell foam whereas the material of the spacer is a closed cell foam. Furthermore, the isotope container is preferably an ion exchange column and each of its opposing ends preferably includes a frangible seal adapted to be pierced by and to seal around the respective first and second hollow needles. In the preferred embodiment the first and second hollow needles are each connected via associated fluid conduits with a fluid inlet and a fluid outlet respectively with the fluid inlet and the fluid outlet ideally consisting of hollow spikes. Also, the device preferably further includes an outer housing within which the shielded chamber is located wherein the fluid inlet and the fluid outlet are mounted in the outer housing to provide fluid connections external to the outer housing. The fluid conduits may each consist of flexible tubing which is greater in length than the distance between the hollow needles and their respective fluid inlet or outlet. In a further aspect the present invention provides a method of constructing a radioisotope generator comprising the steps of: providing a shielded chamber with an opening and a chamber closure adapted for cooperating with and closing the chamber opening; providing a first fluid port comprising a first hollow needle projecting into the shielded chamber from the chamber closure; providing a second fluid port comprising a second hollow needle projecting into the shielded chamber at the end of the chamber opposite the opening; mounting first and second compressible buffers so as to surround at least partially the respective first and second hollow needles, one or each of the compressible buffers including a spacer of predetermined thickness; thereafter introducing an isotope container housing a radioactive isotope through the chamber opening into the shielded chamber so as to contact with the second hollow needle and the second compressible buffer at the closed end of the chamber; and closing the shielded chamber by positioning the chamber closure in the opening and bringing the first hollow needle and the first compressible buffer into contact with the isotope container whereby the spacer determines the positioning of the isotope container within the shielded container. Preferably the method further comprises the steps of, prior to introduction of the isotope container into the shielded chamber, connecting the first hollow needle to a first fluid conduit; connecting the second hollow needle to a second fluid conduit; locating the shielded container within an outer housing and connecting the first fluid conduit to a fluid inlet in the outer housing and the second fluid conduit to a fluid outlet in the outer housing. Ideally, the first and second fluid conduits are each of flexible tubing which is greater in length than the distance between the first and second hollow needles and their respective fluid inlet and fluid outlet when the chamber closure is in place in the chamber opening and the shielded chamber is positioned within the outer housing whereby all fluid connections can be established prior to installation of the isotope container within the shielded chamber. An embodiment of the present invention will now be described, by way of example only, with reference to FIG. 1 which illustrates a radioisotope generator having fluid connections to the ion exchange column in accordance with the present invention. FIG. 1 illustrates a radioisotope generator 1 comprising an outer container 2, a top plate 3 which is sealingly secured to the outer container 2, and a separate top cover 4 which is secured to the outer container 2 over the top plate 3. Inside the outer container 2 an inner shielded container 5, providing shielding against radiation, is located which is preferably, but not exclusively, made from either lead or a depleted uranium core within a stainless steel shell. The shielded container 5 surrounds a tube 6 containing an ion exchange column 7. The molybdenum, in the form of its radioactive isotope 99Mo, is adsorbed on to the ion exchange column 7. The tube 6 containing the ion exchange column has frangible rubber seals 8 and 9 at opposing ends 10 and 11 which, as illustrated, when in use are pierced by respective hollow needles 12 and 13. Each of the hollow needles 12 and 13 is in fluid communication with a respective fluid conduit 14, 15 that are in turn in fluid communication respectively with an eluent inlet 16 and an eluate outlet 17. The fluid conduits 14, 15 are preferably flexible plastic tubing. The tubing 14, extending from the hollow needle 12, passes through a channel in a container plug 18, that closes the upper opening 19 to the shielded container 5, and then extends from the container plug 18 to the eluent inlet 16. The tubing 15, extending from the hollow needle 13, passes through a channel in the shielded container 5 to the eluate outlet 17. The inner shielded container 5 is smaller than the outer container 2 and so there is a free space 20 within the outer container 2 above the shielded container 5. This free space 20 accommodates part of the tubing 14, 15 extending from the hollow needles to the eluent inlet and eluate outlet as the lengths of the tubing 14, 15 are both much greater than the minimum length required to connect the hollow needles 12, 13 with the respective eluent inlet 16 and eluate outlet 17 and their length may be approximately twice the distance to the respective inlet and outlet. The top plate 5 of the radioisotope generator 1 has a pair of apertures 21 through which respective eluent inlet and outlet components project. The eluent inlet and eluate outlet components are each hollow spikes 22 though in the case of the inlet component the hollow spike has two holes, one for the passage of fluid and one that is connected to a filtered air inlet. The hollow spike 22 consists of an elongate generally cylindrical spike body 23 and an annular retaining plate 24 which is attached to or is moulded as a single part with one end of the spike body 23. The opposing end of the spike body 23 is shaped to a point and has an aperture communicating with the interior of the spike body adjacent the point. This pointed end of the spike body 23 is shaped so that it is capable of piercing a sealing membrane of the type commonly found with sample vials. The annular retaining plate 24 forms a skirt projecting outwardly from the spike body 23 and may be continuous around the spike body or discontinuous in the form of a plurality of discrete projections. The top cover 4 of the radioisotope generator 1 also includes a pair of apertures 25 arranged so as to align with the apertures 21 in the top plate 3 and shaped to allow through passage of the spike body 23. Thus, each of the hollow spikes 22 is arranged to be held and supported by its annular retaining plate 24 by component supports 26 provided on the inside of the top plate 3 whilst the hollow spike body 23 projects through the apertures in both the top plate 3 and the top cover 4 to the exterior of the outer container 2. Each one of the apertures 25 in the top cover 4 is located at the bottom of a well 27 that is shaped to receive and support either an isotope collection vial or a saline supply vial. Thus, both vials are housed outside of the outer container 2 and are not exposed to radiation from the ion exchange column 7. In order to supply the ion exchange column with the chloride ions required for elution of the radioisotope, saline solution is drawn through the ion exchange column 7, by establishing a pressure differential across the ion exchange column. This is accomplished by connecting a saline supply vial to the eluent inlet 16 which is in fluid communication with the top end 10 of the ion exchange column 7 via the tubing 14 and hollow needle 12 and connecting an evacuated collection vial to the eluate outlet 17 which is in fluid communication with the bottom end 11 of the ion exchange column 7 via the tubing 15 and hollow needle 13. The pressure differential is established by virtue of the fluid pressure of the saline in the supply vial and the extremely low pressure in the evacuated collection vial. This urges passage of the saline solution through the ion exchange column 7 to the collection vial carrying with it the daughter radioisotope. This process enables the radioactive isotope to be collected without either the outer container 2 or the inner shielded container 5 being opened. In this way radiological protection and aseptic conditions of the isotope generator 1 can be maintained during use. Of course, in the event of failure of the eluate path from the eluent inlet 16 to the eluate outlet 17 repairs would involve the opening of at least the outer container 2 and in all probability the inner shielded container 5 also. In practice such repairs are not undertaken because of the radiation exposure that would ensue. Therefore the reliability of the eluate path is extremely important. Existing radioisotope generators have sought to address this problem through complex designs in which the fluid path from the eluent inlet to the eluate outlet are ‘hard-wired’. That is to say the fluid connections are established during the actual construction of the generator. Such designs, though, have the disadvantage not only of complexity but also the exposure to radiation that results from the generator having to be built around the ion exchange column. The radioisotope generator illustrated in FIG. 1 has been designed to improve the reliability of the eluate path whilst minimising the radiation exposure during construction of the generator. In particular, the construction of the generator involves initially establishing the fluid connection between the hollow needle 13 and the tubing 15 that passes through the shielded container 5 and connecting the tubing 15 to the eluate outlet 17. The top plate 3 and the top cover 4 along with the hollow spikes 22 are connected together and are ready for closing the outer container 2. Similarly, with the container plug 18 free from the opening 19 of the shielded container 5, the fluid connections of the tubing 14 with the eluent inlet and the hollow needle 12 are established with the hollow needle 12 projecting outwardly from the inner end of the container plug 18. The need for the greater lengths of tubing 14, 15 is now apparent as the tubing must be sufficiently long to enable the top plate 3 to be kept clear of the opening to the outer container 2 even after the fluid path has been established. Of course, in addition or as an alternative the tubing could be formed of a resilient or elastic material which permits the tubing to be stretched when the top plate is held away from the opening of the outer container 2. During all of this construction the tube 6 containing the ion exchange to column 7 is not in place within the shielded container 5. Once all construction of the generator 1 is completed, and the only remaining steps are the closure of the inner shielded container 5 and the outer container 2, the tube 6 containing the ion exchange column 7 is inserted into the interior of the shielded container 5. This insertion of the tube may be performed using a robotic arm so as to minimise the extent of any radiation exposure. The opening 19 of the shielded container 2 to the interior space that is to accommodate the tube 6 includes a frusto-conical wall which assists in guiding and aligning the outlet end 11 of the tube 6 in position above the hollow needle 13 at the base of the substantially cylindrical interior space defined by the inner walls of the shielded container 5. Further lowering of the tube 6 down into the interior space results in the tip of the hollow needle 13 contacting and piercing the bottom seal 9 of the tube 6. Further lowering of the tube 6 ensures that the hollow needle 13 penetrates sufficiently into the interior of the tube 6 that the aperture in the tip of the needle 13 is positioned wholly within the tube 6. With the tube 6 now in position within the shielded container 5 the container plug 18 is inserted into the opening 19 of the shielded container 5 to close the shielded container. In the process of positioning the plug 18 into the opening 19 of the shielded container 5 the tip of the hollow needle 12 contacts and then pierces the seal 8 at the top end 10 of the tube 6 to penetrate the interior of the tube. Once the plug 18 is in position, sealing the opening 19 of the shielded container 5, the aperture in the tip of the hollow needle 12 is positioned wholly within the tube 6. There is a risk during this procedure of the hollow needles 12, 13 failing to penetrate far enough into the tube 6 to reliably ensure that the apertures in the tips of the needles are wholly within the tube. To prevent such an occurrence compressible disks 28, 29 are mounted about their respective needles 12, 13. The compressible disk 28 surrounding the upper hollow needle 12 is preferably made of a semi-open cell foam such as polyether and has a cross-section conforming to the cross-section of the interior space of the shielded container 5. The compressible disk 28 therefore acts to provide a protective sleeve to the hollow needle 12 before the needle is inserted into the tube 6 and also cushions the engagement of the container plug 1 8 with the top of the tube 6. The compressible disk 29, which also has a cross-section corresponding to the interior space of the shielded container 5, similarly acts as a protective sleeve about the hollow needle 13 in the base of the interior space into which the tube 6 is inserted. This compressible disk 29 is preferably formed of two separate layers, the first layer 30, adjacent the tip of the needle, is preferably of the same open cell foam as the compressible disk 28. The second layer 31, distant from the tip of the needle, is preferably of a closed cell foam such as polyethylene and is less compressible than the first layer 30. The thickness of this second layer is carefully selected with respect to the length of the needle 13 so that when the tube 6 is lowered over the needle, the needle penetrates a predetermined amount into the tube 6. By accurately controlling the extent of penetration of the needle 13 through the lower seal 9 of the tube, the extent of penetration of the needle 12 through the upper seal 8 can thereby also be controlled. Thus careful selection of the compressibility and the thickness of the two disks the fluid path ensures that the path from the eluent inlet to the eluate outlet can be reliably established in a construction process that minimises the extent of radiation exposure to which the generator is subjected. For example, both discs may consist of a 12.5 mm diameter cylinder comprising a 8 mm long cross-linked polyethylene closed cell foam of density 45 Kg/cubic meter laminated to a 16 mm long polyether semi-open cell foam of density 30 Kg/cubic meter. Thus, with the embodiment of the radioisotope generator described above, the constructional elements of the generator can each be rendered sterile and confined to a sterile environment during construction. Furthermore, during construction the radioactive material, which is confined within a sealed tube, is only introduced at the end of the construction process thereby minimising the radiation exposure during construction. Moreover, this construction process ensures the tube is introduced and is reliably connected to the fluid path of the generator. Further and alternative features of the radioisotope generator and of the process of construction of the generator are envisaged without departing from the scope of the present invention as claimed in the appended claims. |
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06295329& | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS A reactor-internal equipment handling apparatus according to an embodiment of the present invention will be explained in detail with reference to the accompanying drawings hereinafter. FIG. 1 is a longitudinal sectional view showing a reactor-internal equipment handling apparatus 18 according to an embodiment of the present invention. This reactor-internal equipment handling apparatus 18 can simultaneously load/unload all of the CR 7, the FS 8, and the CRGT 6 into/from a reactor by a remote manipulation. The reactor-internal equipment handling apparatus 18 has a main body frame 26. FIG. 1 shows a state where the main body frame 26 is properly positioned at a predetermined position in a reactor pressure vessel 1 (see FIG. 2). FIG. 2 is a longitudinal sectional view showing a state where the reactor-internal equipment handling apparatus 18 is installed inside the reactor pressure vessel 1 by a refueling machine 14. In this case, when the reactor-internal equipment handling apparatus 18 is to be utilized, four fuel assemblies 10 (see FIG. 10) which are located in an objective working area are pulled out from the core previously by the refueling machine 14. A top end of the main body frame 26 is fitted to a bottom end of a hoist rope 25 of the refueling machine 14 shown in FIG. 2 such that the main body frame 26 can be lifted up and down by the refueling machine 14. A guide member 27 is fitted to the hoist rope 25 to be faced to a top surface of the main body frame 26. To the main body frame 26 are fitted a control rod grapple (referred to as "CR grapple" hereinafter) 16 acting as a control rod holding means which releasably holds the CR 7 being placed inside the reactor pressure vessel 1, and fuel support/control rod guide tube grapples (referred to as "FS/CRGT grapples" hereinafter) 17 each acting as a fuel support/control rod guide tube holding means which releasably holds both the FS 8, which supports the bottom end of the fuel assembly 10, and the CRGT 6, on which the FS 8 is positioned at top end. FIG. 3 is a longitudinal sectional view showing the main body frame 26 and the FS/CRGT grapples 17 in an enlarged manner. A pair of FS/CRGT grapples 17 are fitted to the main body frame 26. Each of the FS/CRGT grapples 17 comprises an orifice engaging hook (orifice engaging member) 28 which can engage edge portions of both the FS orifices 33 shown in FIG. 12 and the CRGT orifices 32 shown in FIG. 11, an orifice engaging hook linking mechanism (orifice engaging member linking mechanism) 30 which is employed to operate the orifice engaging hook 28, and an orifice engaging hook driving cylinder (orifice engaging member driving means) 19 which is employed to drive the orifice engaging hook linking mechanism 30. Preferably, each of the orifice engaging hook driving cylinders 19 is composed of an air cylinder. Also, a clamping state detecting mechanism (holding state detecting mechanism) 20, which detects holding states of the FS/CRGT grapples 17 about the FS 8 and the CRGT 6, is provided to each of the FS/CRGT grapples 17. This clamping state detecting mechanism 20 is placed on a top portion of the orifice engaging hook driving cylinder 19. The clamping state detecting mechanism 20 has limit switches (detection switches) 44 whose on/off state can be switched depending on a change in clamping states. More particularly, the clamping state detecting mechanism 20 has a limit switch 44 which is directly on/off-operated by an output axis of the orifice engaging hook driving cylinder 19 when the output axis is moved back and forth simultaneously with a motion of the orifice engaging hook 28, and a limit switch 44 which is on/off-operated by transmitting a back-and-forth motion of an output axis via a lever mechanism 62. Further, each of the FS/CRGT grapples 17 has a guide portion 29. This guide portion 29 has a function of seating the main body frame 26 on a predetermined position without fail by guiding an inside of the fuel assembly sustaining hole 31 (see FIG. 12) of the FS 8. Both an FS stepped portion 34a which comes into contact with an edge portion of the FS orifice 33 (see FIG. 12) of the FS 8, and a CRGT stepped portion 34b which comes into contact with an edge portion of the CRGT orifice 32 (see FIG. 11) of the CRGT 6 are formed on the orifice engaging hook 28. With the use of the FS stepped portions 34a and the CRGT stepped portions 34b, both the FS 8 and the CRGT 6 can be handled simultaneously. The orifice engaging hook linking mechanism 30 is so constructed that an opening/closing motion of the orifice engaging hook 28 can be disabled in the situation that the FS stepped portion 34a and the CRGT stepped portion 34b are brought into contact with the edge portions of the FS orifice 33 and the CRGT orifice 32 respectively. In more detail, when the orifice engaging hook linking mechanism 30 is shifted from its clamping state (holding state) to its releasing state (non-holding state), the orifice engaging hook 28 once protrudes outwardly from the orifices 32, 33 and then withdraws inwardly. Thus, in the situation that both the FS 8 and the CRGT 6 are being hoisted or only the FS 8 is being hoisted, a mechanical lock can be made by its own weight of the hoisted substance and the FS stepped portions 34a and the CRGT stepped portions 34b of the orifice engaging hooks 28. Therefore, even when either an actuating pressure of the orifice engaging hook driving cylinder 19 is lost at the worst or the operator operates it erroneously, the hoisted substance is never released or unengaged. As shown in FIGS. 4 and 5, the reactor-internal equipment handling apparatus according to the present embodiment further comprises a stroke varying mechanism 35 which can change an operation stroke of the orifice engaging hook driving cylinder 19. The stroke varying mechanism 35 is composed of disk-like stoppers 36, stroke varying blocks 37, and an arm 38. The disk-like stoppers 36 are provided to output axes of two orifice engaging hook driving cylinders 19 of the FS/CRGT grapples 17 on the orifice engaging hook linking mechanism 30 side respectively. Each of the stroke varying blocks 37 is rotatably mounted between the disk-like stopper 36 and the orifice engaging hook driving cylinder 19 by pins 39 being provided to a cylinder case. The arm 38 can couple the stroke varying blocks 37 with each other. When the arm 38 is moved vertically, both stroke varying blocks 37 are put in and out simultaneously. A swing amount (amount of motion) of the orifice engaging hook 28 can be adjusted by changing an operating stroke of the orifice engaging hook driving cylinder 19 by the stroke varying mechanism 35. Therefore, the orifice engaging hook 28 can be set not to be engaged by the edge portion of the orifice 32 of the CRGT 6. As a result, the FS/CRGT grapple 17 cannot clamp the CRGT 6, but it can clamp only the FS 8. For example, in the case that a load applied in pulling out the CRGT 6 exceeds a limit load of the hoist of the refueling machine 14 because the core plate 3 and the CRGT 6 have stuck together, only the CR 7 and the FS 8 can be hoisted by operating the stroke varying mechanism 35 in order not to exceed the limit load of the refueling machine 14. As shown in FIG. 1, the CR grapple 16 is fitted to the main body frame 26 such that it can be slid by a predetermined width along the longitudinal direction of the CR 7. This predetermined sliding width is defined by the distance between the flange portion 70a of the movable member 70 and the inner upper surface 71a of the cap member 71. The movable member 70 is fixed to both the hoist rope 25 and the CR grapple 16. On the other hand, the cap member is fixed to the main body frame 26. In contrast, the FS/CRGT grapple 17 is secured to the main body frame 26. For this reason, the CR grapple 16 and the FS/CRGT grapple 17 can be relatively displaced mutually along the longitudinal direction of the CR 7. Therefore, when the CR 7, the FS 8, and the CRGT 6 are to be hoisted, first the CR 7 is hoisted slightly and then the FS8 and the CRGT 6 are hoisted. In this manner, a time difference can be introduced into an application of the load, which is equivalent to the head pressure caused by the air contained in the CRD housing 4, by hoisting the CR 7 previously, and as a result the simultaneously applied load can be reduced. Accordingly, the hoisting load to unload outside the reactor can be shared much more by the head pressure to lift the CRGT 6. As shown in FIG. 6, the CR grapple 16 has a hook (handle engaging member) 41 acting as an L-shaped swingable hooking member which can engage a hoisting handle 7a (see FIG. 1) secured to the top end of the CR 7. This hook 41 has a gaff 43. The hook 41 is connected to a hook driving cylinder (handle engaging member driving means) 40 via a linking mechanism 42, and is operated by the hook driving cylinder 40 to be swung. Then, in the situation that the CR 7 is hoisted via the hook 41, a mechanical lock using its own weight of the CR 7 can be made by the linking mechanism 42 and the gaff 43 of the hook 41. Such mechanical lock can act to hold a engaged state of the hoisting handle 7a by the hook 41. In addition, as shown in FIG. 6, a clamping state detecting mechanism (holding state detecting mechanism) 60 which detects a clamping state (holding state) of the CR 7 is provided to the CR grapple 16. This clamping state detecting mechanism 60 has a limit switch (detection switch) 61 whose on/off state can be changed depending upon a swing motion of the hook 41. More specifically, an on/off switching operation of the limit switch 61 is performed by a base end portion of the hook 41. Also, as shown in FIG. 6, the reactor-internal equipment handling apparatus 18 according to the present embodiment is connected to an external power supply 55 which is arranged apart from the reactor-internal equipment handling apparatus 18. A power supply for clamping state confirming indicator lamps (holding state confirming indicator lamps) 45 and a seating state confirming indicator lamp (positioning state confirming indicator lamp) 50, which are shown in FIG. 7, can be supplied from this external power supply 55. In place of the external power supply 55, a built-in battery (not shown) can be incorporated into the reactor-internal equipment handling apparatus 18. In this case, exchange of the battery must be performed by pulling up the main body frame 26 every run-down of the battery. If the work of unloading all the CRS 7, the FSs 8, and the CRGTs 6 must be carried out in the preventive maintenance work, etc., it is preferable to supply the power supply from the external power supply 55, which is placed on the refueling machine 14, etc., since an employment term of the reactor-internal equipment handling apparatus 18 is prolonged over a long term. As shown in FIG. 7, a plurality of clamping state confirming indicator lamps 45 whose lighting state can be changed depending upon a change in clamping states (holding states) are provided to both the clamping state detecting mechanism 20 (see FIG. 3) for the FS 8 and the CRGT 6 and the clamping state detecting mechanism 60 (see FIG. 6) for the CR 7. These clamping state confirming indicator lamps 45 are attached to the top surface of the main body frame 26. The clamping state confirming indicator lamps 45 can switch their lighting states according to on/off states of the limit switches 44, 61 (FIG. 3 and FIG. 6). In more detail, at least three clamping state confirming indicator lamps 45 are provided. A lighting state of a first clamping state confirming indicator lamp 45 can be switched by the limit switch 61 (see FIG. 6) which is switched depending upon a change in the clamping state of the CR 7. Also, a lighting state of a second clamping state confirming indicator lamp 45 can be switched by the limit switch 44 (see FIG. 3). When the orifice engaging hooks 28 shown in FIG. 3 is located at the position to clamp both the FS 8 and the CRGT 6, such limit switch 44 can be switched depending upon the change in the clamping state of the FS 8 and the CRGT 6 via the lever mechanism 62. In addition, a lighting state of a third clamping state confirming indicator lamp 45 can be switched by the limit switch 44. When the orifice engaging hooks 28 is located at the position to clamp only the FS 8, such limit switch 44 can be switched by the output axis of the orifice engaging hook driving cylinder 19 depending upon the change in the clamping state of the FS 8. In this way, by checking the lighting state of plural clamping state confirming indicator lamps 45 with the naked eye, the operator can know whether or not the reactor-internal equipment handling apparatus 18 has already clamped the CR 7, the FS 8, and/or the CRGT 6. As shown in FIG. 8 and FIG. 9, the reactor-internal equipment handling apparatus 18 is equipped with a first positioning state detecting mechanism 63 (FIG. 8) and a second positioning state detecting mechanism 64 (FIG. 9), which detect a positioning state of the main body frame 26 in the reactor pressure vessel respectively. As shown in FIG. 8, the first positioning state detecting mechanism 63 consists of a seating state detecting mechanism 21, a motion limiting mechanism 23, and a motion limiting mechanism locking device 24. While, as shown in FIG. 9, the second positioning state detecting mechanism 64 consists of the seating state detecting mechanism 21, and the motion limiting mechanism 23. The seating state detecting mechanism 21 is composed of seating detecting pins 46, 47, a cam mechanism 48, and a limit switch 49. The seating detecting pin 46 of the first positioning state detecting mechanism 63 is employed to detect the positioning pin 11 (see FIG. 1). The seating detecting pin 47 of the second positioning state detecting mechanism 64 is employed to detect the top surface of the FS 8. Two seating detecting pins 46, 47 are projected from the bottom surface of the main body frame 26. Thus, when the main body frame 26 is positioned or seated, the seating detecting pins 46, 47 are pushed upwardly by the top surfaces of the positioning pin 11 and the FS 8 on the core plate 3 (see FIG. 1). Such motions of the seating detecting pins 46, 47 are transmitted respectively via the cam mechanisms 48 to the limit switches 49, whereby the limit switches 49 can be operated. Then, when both limit switches 49 provided to the first positioning state detecting mechanism 63 and the second positioning state detecting mechanism 64 are operated, the seating state confirming indicator lamp 50 (see FIG. 7) provided on the top surface of the main body frame 26 can be lightened. In this fashion, based on the lighting state of the seating state confirming indicator lamp 50, the operator can visually check that the main body frame 26 has been seated on the position to properly clamp and unclamp the CR 7, the FS 8, and the CRGT 6. The motion limiting mechanism 23 has a cam 51 which is operated simultaneously with motions of the seating detecting pins 46, 47, and two valve switches 52, 52 whose on/off is switched by a vertical motion of the cam 51. Then, during the hoisting operation of both the FS 8 and the CRGT 6 or only the FS 8, motions of the orifice engaging hook driving cylinders 19 are limited by an air circuit (not shown) which is connected to the valve switches 52, 52. In other words, except the case where the main body frame 26 is seated on the proper position or where no load is applied the FS/CRGT grapple 17, i. e., the FS/CRGT grapple 17 is holding nothing, an actuating fluid is not supplied to the orifice engaging hook driving cylinders 19 so as to disable the orifice engaging hooks 28 of the FS/CRGT clamping mechanism 17. Accordingly, even when either an operating pressure of the orifice engaging hook driving cylinders 19 is lost or the operator performs the wrong operation in the course of the hoisting operation of both the FS 8 and the CRGT 6 or only the FS 8, the reactor-internal equipment handling apparatus 18 never releases the FS 8 and the CRGT 6. As shown in FIG. 8, the motion limiting mechanism locking device 24 is composed of a ball lock pin 54 and a stepped hole (not shown). The ball lock pin 54 is fitted to the upper portion in the main body frame 26. The stepped hole is formed over the motion limiting mechanism cam 51. The motion limiting mechanism locking device 24 is employed to use the Pin-FS 22 (see FIG. 13). More specifically, the Pin-FSs 22 which support a mimic fuel assembly (not shown) are provided to the peripheral portions of the core of the BWR. When the Pin-FSs 22 are employed, the main body frame 26 is rotated by 90 degree rightward or leftward rather than a normal orientation to avoid interference with the pin 53 (see FIG. 13), and then seated. The pin 53 is provided in the Pin-FS 22 to indicate the position of the mimic fuel assembly. In this case, detection of the seating state and restriction of the motion by virtue of the positioning pin 11 provided on the core plate 3 cannot be achieved. Therefore, the motion limiting mechanism 23 is operated and thus the motion of the orifice engaging hook 28 of the FS/CRGT grapple 17 is inhibited. As a result, the Pin-FS 22 and the CRGT 6 cannot be removed. For this reason, in such case, the ball lock pin 54 is inserted into a hole formed on the upper area of the cam 51 of the motion limiting mechanism 23 to lock the cam 51. At that time, since the seating detecting pin 46 of the first positioning state detecting mechanism 63 has already been in a seated condition, the seating on the normal position can be detected only by detecting the top surface of the Pin-FS 22 by the seating detecting pin 47 of the second positioning state detecting mechanism 64. As a result, clamping/unclamping of the Pin-FS 22 and the CRGT 6 can be achieved. In case the Pin-FS 22 is handled as described above, a function of the motion limiting mechanism 23 is lost. In this case, since the mechanical lock which has already been mentioned can be operated, the hoisted substance is in no way released even if, for example, the operating pressure of the orifice engaging hook driving cylinder 19 is lost or the operator performs the wrong operation. As described above, according to the reactor-internal equipment handling apparatus 18 of the embodiment of the present invention, in the event that the removing operation or the installing operation of the CR 7, the FS 8, and the CRGT 6 must be performed in the periodical inspection or the preventive maintenance work, all of the CR 7, the FS 8, and the CRGT 6 can be loaded/unloaded into/from the reactor simultaneously by the CR grapple 16 and the FS/CRGT grapple 17. Therefore, the number of steps can be reduced to half based on a simple calculation rather than the case where the CR 7 and the FS 8 are handled separately from the CRGT 6 in the related art, so that a term of work can be shortened considerably. In the reactor-internal equipment handling apparatus 18 according to the embodiment of the present invention, an operability can be assured to the same extent as the CR & FS grapple in the related art or more. In addition, if the CR & FS grapple and the CRGT grapple in the related art are employed, the CR 7/the FS 8 and the CRGT 6 must be stored separately based on the installing order in the reactor. Therefore, the wide storage space is needed as the fuel pool serving as the storage area. On the contrary, according to the reactor-internal equipment handling apparatus 18 of the embodiment of the present invention, since the CR 7, the FS 8, and the CRGT can be handled together, they can be stored collectively. Therefore, based on a simple calculation, the storage space can be reduced half of the storage space needed in the related art. As described above, according to the reactor-internal equipment handling apparatus and method of the present invention, since all of the control rod, the fuel support, and the control rod guide tube can be loaded/unloaded into/from the reactor simultaneously, both reduction in the term of work and reduction in their storage spaces can be achieved. |
055132294 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS A conventional position indicator probe has 53 position switches (see FIG. 5), usually designated S00 to S52. Forty-nine of the reed switches are spaced at equal 3-inch intervals on the switch support 103 (FIG. 4), providing an indication signal at each locking position and at the halfway point between each locking position. These switches are closed one at a time to transmit signals to the rod position information system (not shown), which in turn energizes the corresponding rod position display in the control room. The control room digital displays resulting from the closing of these 49 switches range from "00" (control rod fully inserted into the core) to "48" (control rod withdrawn to the backseat position). The switches in the position indicator probe are assigned corresponding number designations, S00 to S48. The even-numbered switches correspond with the locking positions and the backseat position of the index tube 26 (FIG. 1); the odd-numbered switches correspond with the intermediate positions on the index tube. Switches S00, S51 and S52, located at the top of position indicator probe 12a (FIG. 2A), provide the control room with a signal indicating "rod full in". Switch S49, located near the bottom of the probe and on the opposite side of switch support 103 from switch S48 (see FIG. 4), closes simultaneously with switch S48 to provide the control room with a signal of "rod full out". Switch S50 is installed 2 inches below the normal full-out position of the CRD, i.e., at the so-called "overtravel" position. Since the limit of drive piston down travel is provided by the backseat position of the control rod in the reactor vessel guide tube (not shown), switch S50 is closed only when the control rod and CRD are uncoupled and, when closed, provides the control room with an annunciation of this condition. The present invention utilizes switches S47, S49 and S50, during CRD removal, to monitor CRD uncoupling, as described in detail hereinafter. In accordance with a preferred embodiment of the invention depicted in FIG. 6, the means for supporting the CRD during removal comprise a generally circular cylindrical bucket adaptor 100 coupled to an extension tube 110 by a pin 113. The adaptor has a first circular cylindrical chamber 115 of diameter slightly greater than that of the ring flange 17 and a second circular cylindrical chamber 117 of diameter less than that of the ring flange 17 and greater than that of position indicator probe housing 14. These chambers are connected by way of a seat 101 on which ring flange mounting bolts 9 bear. Seat 101 is preferably an annular horizontal surface having an inner diameter which is less than the diameter of a pitch circle circumscribing the heads of ring flange mounting bolts 9. Extension tube 110 supports the CRD by way of bucket adaptor 100 after the mounting bolts 2 are unscrewed using a conventional de-torquing tool 112. Before the bucket adaptor/extension tube assembly is installed, an electronic monitoring tool 102 is connected to receptacle 14 of the position indicator probe assembly (see FIG. 6) via a plug 106 and shielded cables 104. Plug 106 fits snugly into receptacle 14, allowing tool 102 to be suspended via the shielded cables. Bucket adaptor 100 has a chamber for receiving the suspended monitoring tool 102. A window 111 is provided to allow observation of indicator lights 123-125 which are visible through a transparent enclosure (e.g., made of LEXAN.TM. pipe) of electronic monitoring tool 102. In accordance with the CRD removal method of the present invention, the position indicator probe 12a can be left in place inside the position indicator tube 61 and used to monitor the state of CRD coupling/uncoupling. A green LED 124 lights when ring magnet 67 is aligned with switch S50, indicating that the index tube is in the overtravel position, i.e., the CRD is uncoupled from the control rod. This condition indicates that the CRD can be safely removed from the CRD housing. A red LED 123 lights when ring magnet 67 (FIG. 2A) is aligned with either of switches S48 or S49, indicating that the index tube is in the backseat position or the adjacent intermediate position, respectively. In either case, the CRD should be pushed back into the housing and then uncoupled from the control rod. Switch S47 indicates that the index tube is extended just short of the first locking position (corresponding to switch S46) and also lights the red LED 123. Extension to the first locking position is to be avoided since the CRD cannot be simply pushed back in to retract the index tube. Finally, a flashing red LED 125 flashes when the ring magnet 67 is between adjacent switches, i.e., all position switches S00 to S52 are open. For example, green LED 124 might be lit, indicating that the CRD can be removed, but upon lowering of the CRD, the flashing red LED is activated, along with a pulsed horn 129 (see FIG. 7). The flashing red light and pulsed horn warn the crew that the CRD is not uncoupled and that CRD removal should be halted. The monitoring circuit in accordance with the preferred embodiment of the invention is shown in FIG. 7 connected to pins 3, 4, 6 and 8 of plug 106. Pin 8 is connected to ground; pins 3, 4 and 6 are respectively connected to switches S49 (or S48), S47 and S50 (see FIG. 5). A first circuit loop is formed by a closed ON/OFF switch 120, a dc voltage source 121 (e.g., a 9-volt battery), red LED 123, a closed switch S47 or S49 connected to pin 3, and a resistor 126 (470 .OMEGA.). Red LED 123 is activated when switch S47 or S49 is closed. Alternatively, green LED 124 is activated when switch S50, connected to pin 6, is closed. The light-emitting diodes 123 and 124 can be tested by pressing test pushbuttons 108 and 108', respectively. A switching transistor 122 (e.g., 2N1711 or its equivalent) has its base coupled to the positive terminal of the voltage source via resistors 126 and 127 (both 470 .OMEGA.) connected in series. The emitter of transistor 122 is connected to the positive terminal of the voltage source via a parallel network: one line consisting of a resistor 128 (1 k.OMEGA.) and flashing red LED 125 connected in series, and the other line consisting of a pulsed horn 129. The collector of transistor 122 is connected to the negative terminal of the voltage source via a series of diodes 130 (e.g., 1N914). When switches S47, S49 and S50 are all open, transistor 122 is biased to switch on the flashing red LED 125 and the pulsed horn 129. Using the monitoring circuit shown in FIG. 7, a CRD can be removed from a CRD housing with its position indicator probe in place. A signal is activated to indicate that the index tube occupies a position corresponding to CRD uncoupling from the control rod, i.e., the CRD is ready to be removed from its housing. In the event that the CRD is not uncoupled, another signal will be activated as the index tube, which is still coupled at its end to the control rod via the spud, is extended relative to the descending CRD. The preferred embodiments of the bucket adaptor and electronic monitoring tool have been disclosed for the purpose of illustration. Variations and modifications of the disclosed structure which do not depart from the concept of this invention will be readily apparent to mechanical engineers skilled in the art of monitoring devices. Also it should be borne in mind that different nuclear reactors may have switch or pin numbering schemes different than those disclosed herein. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter. |
052805050 | description | DETAILED DESCRIPTION Referring first to FIG. 1, a radioisotope generating apparatus or system which may be utilized in practicing the teachings of this invention is shown. The apparatus 10 consists of a sealed chamber 12 having a cryogenic dewer 14 positioned therein. A desired pressure, for example, vacuum pressure, may be maintained in chamber 12 by a suitable vacuum source, for example, a pump 16, connected to the chamber through tube 18 and sealed port 20 leading into the chamber. Alternatively, vacuum pressure may be obtained from the accelerator in a manner to be described later. Liquid nitrogen 21 or another suitable cooling agent such as liquid helium or liquid oxygen is applied to dewar 14 from a suitable source through tube 22 which tube passes through a port 24 in chamber 12. The cooling agent (coolant) may be removed from dewar 14 through a tube 26 attached to the dewar, which tube passes through a sealed port 28 in chamber 12. Chamber 12 also has a port 30 which is a spare port which may be used for taking measurements or other suitable purposes, and a port 32 having a tube 34 passing therethrough. The end of tube 34 in chamber 12 has a vapor jet nozzle 36 which is pointed in a generally horizontal direction. The end of tube 34 outside of chamber 12 is connected through a tube 38 and valve 40 to a target material reservoir 42. Tube 34 is mounted in a nozzle retraction assembly 44 which raises the nozzle to the position shown in FIG. 1 when the nozzle is to be utilized and otherwise retracts the nozzle to a position near the bottom of chamber 12 or in port 32. A funnel-shaped or cone-shaped target 46 is mounted in the lower portion of cryogenic dewar 14 with the axis of the cone oriented horizontally. The wide end of the cone is positioned opposite nozzle 36 and is sealed by a sealing ring 48 in the dewar. A plurality of cooling rings 50 are formed around the outer periphery of cone 46. The cone 46 and rings 50 are formed of a material having good heat transfer, and preferably also good electrical conduction, properties, for example a metal such as copper. The cone and rings may be integrally formed or may be separate elements which are pressure-fit, soldered or otherwise secured together. For a preferred embodiment, the cone is initially formed with a thick wall, and grooves are then machined into the walls to form the fins 50, which fins are thus integral with and concentric with the cone. As may be best seen in FIG. 2, there is a small opening 52 at the tip of cone 46 which leads into a channel 54 in a tube 56 extending from the cone tip. Tube 54 is connected by a fitting 58 (FIG. 1) to an extraction tube 60 which passes out of dewar 14 and chamber 12 through tube 22. Extraction tube 60 would be connected to a suitable collection vessel (not shown). The final port on chamber 12, port 62, is connected through a sealed joint 64 to a fast solenoid gate valve 66. Gate valve 66 can be used to seal port 62 under circumstances to be described later, but is normally open. The gate valve is connected through a sealed joint 68 to a rotating bellows assembly 70. Assembly 70 has a pivot 72 about which the entire assembly to the left thereof in FIG. 1 may rotate from the generally horizontal position shown in FIG. 1 to a vertical position 90.degree. counterclockwise from the position shown. The flexible metal bellows 74 flexes as the assembly is rotated to maintain an airtight seal during rotation. Assembly 70 is connected at an airtight sealed joint 76 with a high energy particle accelerator 78. The high energy particle accelerator may be, for example, a cyclotron particle accelerator, which provides higher yields, or a tandem cascade accelerator such as that shown in U.S. Pat. No. 4,812,775, issued Mar. 14, 1989. The tandem cascade accelerator, which is smaller and less expensive, utilizes a lower energy beam at higher current than accelerators such as a cyclotron. Other lower energy, high current accelerators which might be utilized as the accelerator 78 are shown in copending application Ser. No. 07/488,300, filed Mar. 2, 1990. Accelerator 78 may, depending on the isotope desired, be generating accelerated protons, deuterons, electrons, or other particles. For a preferred embodiment of the invention where the apparatus is being utilized to produce fluorine-18 (.sup.18 F), a tandem cascade accelerator is utilized to produce an up to 1 mA beam of 3.7 MeV protons which impinge on a target of enriched .sup.18 0-ice. One problem with prior art devices for generating radioisotopes is that when the high energy beam impinged on the target, which target was generally in liquid or gaseous form, the heat of the reaction would cause vaporization of the target substance. Further, the impingement of the high energy beam on the target material could also cause radiolysis as previously described, resulting in the release of gases such as hydrogen and oxygen. These released gases create a vapor pressure which varies with the target substance and beam energy, which vapor pressure, in conjunction with the normal target pressure of a liquid, degrades the vacuum required in accelerator 78. Therefore, it has been necessary to provide a window in junction 76, generally a thin metal foil, to separate the target chamber 12 from the accelerator 78. However, such windows, particularly for low energy, high current accelerators, tend to get hot as they absorb a small portion of the beam energy passing therethrough, and extensive cooling overhead may be required to prevent such windows from burning out. Further, if the total target pressure becomes substantial, the pressure differential across the window causes stresses in the window which may ultimately result in window failure. Window failure from pressure, heat or a combination thereof is, therefore, a significant maintenance problem in prior art radioisotope generators. It is, therefore, desirable to eliminate the need for such a window by reducing or eliminating the vapor pressure resulting from radioisotope generation so that either a window is not required, or the pressure gradients across the window are sufficiently small that window damaging stresses do not develop. Where a window is not employed in junction 76 and gate valve 66 is open, vacuum pressure in accelerator 78 is applied directly to chamber 12 so that pump 16 need only be used to pressurize the chamber, not to evacuate it. In accordance with the teachings of this invention, the objective of reducing pressure gradient across the junction 76, and thus permitting the window to be eliminated, is generally accomplished by employing a solid target, and in particular a frozen or cryogenic target, which is designed so as to minimize vaporization at the target surface. Since radiolysis is known to be substantially reduced in solids due, for example, to the lower mobility of free radicals, such a target also reduces the material losses due to radiolysis, and thus increases radioisotope yield for a given quantity of target substance and also reduces the vapor pressure causing release of the radiolysis gases. In particular, the parameter G, defined as the number of molecules radiolysed per 100 eV of incident particle energy, is roughly a factor of 10 lower for ice at 77.degree. K. than for room temperature water. This decrease in G with temperature may be due to trapping and subsequent recombination of radiolysis products in the solid lattice which reduces the number of chain reactions involved in radiolysis compared to a liquid target. In addition, the fraction of molecular products which actually escape the solid lattice should decrease with lowered temperature, thus further lowering the value of G. In particular, with the assembly oriented as shown in FIG. 1, pump 16, or preferably accelerator 78, applies vacuum to chamber 12 to evacuate this chamber. Liquid nitrogen 21 or other coolant is also applied through tube 22 to cryogenic dewar 14, reducing the temperature in the dewar to approximately 77.degree. K. The temperature of target cone 46 is also reduced to approximately 77.degree. K. Nozzle 36 is then raised by assembly 44 to the position shown in FIG. 1 directly adjacent cone 46 and valve 44 is opened for a selected time period. Since nozzle 36 is at vacuum pressure while reservoir 42 is at the vapor pressure of water, when valve 40 is opened, vapor will be drawn from reservoir 42 at a known rate through tube 38 and tube 34 to nozzle 36. Thus, by controlling the duration that valve 40 is open, a precisely controlled quantity of target material is permitted to pass to nozzle 36. The velocity of the fluid traveling through tube 34 and the construction of nozzle 36 causes a vapor jet of the target material to be directed toward cone 46. This vapor freezes on cone 46 to form a thin layer 80 (FIG. 3) of the target material on the interior surface 82 of cone 46. With the cone 46 maintained at 77.degree. K., the sticking fraction of the target material from nozzle 36 on cone 46 is greater than 90%. The vapor jet is a directional technique for depositing the target material in a specific location, the nozzle being designed generally to confine the target material to a selected expansion angle, for example 60.degree.. By varying the distance between the nozzle and cone 46, the coverage of frozen target material on the cone can be varied. Since the water vapor density is larger in the center of the jet than at the edges, depositing on the inverted cone may aid in creating a more uniform coating. While the desired coating on cone 46 may be achieved by merely introducing target material into chamber 12, this will result in a significantly lower percentage of the target material inputted into the chamber being deposited and frozen on the inside of cone 46. The additional target material in chamber 12 must ultimately be removed and is, therefore, undesirable. Further, the cost of the target material, for example $100/ml for .sup.18 0-water, makes it economically desirable that such target material not be wasted. While forming the target as a cryogenic ice layer has advantages as indicated above in providing both increased yield due to reduced radiolysis and reduced vapor pressure, the deposition of such a cryogenic target material on a cone shaped target provides additional advantages. First, in order to adequately cool the target ice layer 80, it is important that the ion beam be spread over as large an area as possible, preferably greater than 10 cm.sup.2. This could be done by expanding the ion beam from generator 78 using a magnetic lens. However, at the beam energy required for efficient production of radioisotopes such as .sup.18 F, the required magnetic lens is inconveniently bulky. A simpler method of spreading the beam over a large area is to have the target mounted at an oblique angle to the ion beam. This may be accomplished with an inclined plane, but is preferably accomplished with the cone-shaped target 46 oriented as shown in FIG. 1. The cone geometry has an additional advantage as illustrated in FIG. 3 in that the beam path through the frozen target layer 80 is larger than the perpendicular distance from the surface of the ice to the cooled surface 82 of cone 46 (i.e. t.sub.b >t.sub.i). Since the temperature of the ice increases with distance from surface 82, and since there is a minimum beam path length t.sub.b' which the beam must pass through the target material in order for a desired quantity or yield of radioisotope to be obtained from the target, the geometry shown in FIG. 3 allows the surface of the ice layer to be maintained at a lower temperature than would be possible with a flat target mounted perpendicular to the ion beam while still obtaining the desired yield. The lower surface temperature of ice layer 80 reduces the amount of evaporation from the surface and thus reduces vapor pressure and enhances yield. This geometry also reduces the amount of target material required to load the target, a thin layer of target material being usable, and thus reduces the cost for radioisotope production. To determine the thickness t.sub.i for ice layer 80 in order to obtain a beam length t.sub.b' for a given target material which is suitable for the formation of the desired quantity of radioisotope for a cone having a given cone angle .theta., the following equation applies: EQU t.sub.i =t.sub.b' sin .theta./2 (1) This equation may need to be modified by a factor d which is the density of the ice or other frozen target material in gm/cm.sup.3 such that Equation 1 becomes: ##EQU1## Where t' is the required target thickness in gm/cm.sup.2. For a preferred embodiment where .sup.18 F is being generated from .sup.18 O ice using a 3.7 MeV proton beam, t.sub.b' is approximately 136 micrometers. For this configuration, and a cone angle .theta. of 30.degree., the thickness of layer 80 is approximately 35 micrometers, for a total volume of target material of approximately 0.042 cm.sup.3. However, a thinner layer of .sup.18 0 ice may be utilized where optimum .sup.18 F yield is not required to reduce heating of the ice. When depositing of frozen target layer 80 is complete, gate valve 66 is opened, if it is not already opened to create the vacuum. Assembly 44 is also operated to retract nozzle 36 to a position at the bottom of chamber 12 or in port 32. Accelerator 78 is then operated to apply a proton or other suitable particle beam of suitable energy and current to target layer 80. The duration of target radiation will vary with the radioisotope desired and the reaction utilized to obtain it, but is normally related to the half life of the radioisotope. Thus, for example, for the .sup.18 F reaction previously discussed, the radiation time is approximately 110 minutes which is equal to the half life of .sup.18 F. Many of the radioisotope creating reactions have a threshhold energy. Thus, in order for the .sup.18 F reaction previously discussed to occur, a minimum energy of 2.5 MeV is required. Thus, if a 3.7 MeV proton beam is utilized, only 1.2 MeV of the beam energy need be deposited in ice layer 80, since anything beyond this will not result in .sup.18 F formation. This will yield 2.7 Ci/mA. The remaining 2.5 MeV of the protein beam energy is dissipated in cone 46. In order to avoid overheating of the ice, less than the 1.3 MeV may actually be deposited in the ice in practical applications so long as desired quantities of radioisotopes can be obtained with such lesser energy. Therefore, since a substantial amount of beam energy is dissipated in the cone, including both the energy initially deposited in the ice and that deposited in the cone, and in order to maintain cone 46 at a preferred temperature of approximately 77.degree. K., the coolant 21 in dewar 14 must be able to remove this quantity of heat from the cone. However, coolants have a burn out heat flux. Thus, if liquid nitrogen is used to remove more than approximately 10 W/cm.sup.2, a burn-out of heat flux occurs so that the liquid nitrogen loses its ability to cool and temperature rises quickly. This is because vapor film boiling at this point surrounds the entire object, and thus heat cannot be removed by convection. Sufficient heat must be dissipated across the barrier radiatively, resulting in the temperature rise. In order to avoid this burn out heat flux effect, fins 50 are provided on cone 46 to increase its surface area. While the total external surface in contact with the coolant for the cone alone is only 12 cm.sup.2, the fin assembly may be dimensioned to increase the total surface area to approximately 360 cm.sup.2 for a preferred embodiment, providing more than adequate surface area to avoid flux burn out. Some proton beam energy will also be dissipated in the ice layer 80. However, since the ice layer is very thin, this energy should not raise the temperature of the ice layer more than a few degrees and should result in minimum vaporization. When radiation of the target is complete, the desired yield of the radioisotope having been obtained, accelerator 78 is turned off and solenoid gate 66 is preferably closed to isolate the accelerator from chamber 12. The entire assembly 10 to the right of pivot point 72 is then rotated about pivot point 72 in a counterclockwise direction 90.degree. so that the axis of cone 46 is vertical with the tip of the cone pointing downward. The apparatus may be moved to this position manually with a suitable latch and release being provided in each detent position to assure proper orientations, or a suitable manually or automatically controlled mechanism may be provided for effecting such movement. With the apparatus oriented in the vertical position described above, coolant is pumped out of dewar 14 through tube 26, permitting the temperature in the dewar, and thus the temperature of cone 46, to rise rapidly to room temperature. This causes the frozen target material, which has been altered to contain the desired radioisotope, to melt and to flow down the sides of cone 46 to accumulate as a droplet at the tip of the cone. To the extent surface tension or the like may prevent all of the melted target material from flowing under the effect of gravity to the tip of the cone, a mechanism may be provided to, for example, vibrate the cone, or preferably the entire assembly, to break such surface tension bonds and to facilitate the flow of all of the target material to the tip. The vacuum in chamber 12 is preferably removed before the melting operation, for example, by the closing of gate valve 66. When the droplet of target material is formed in the tip of cone 46, a slight positive pressure is applied by pump 16 to chamber 12 to force the droplet out through opening 52 and channel 54 into extraction tube 60 and out through the extraction tube to the collection vessel (not shown). The apparatus may then be returned to the orientation shown in FIG. 1, again either manually or by use of a suitable motor or other mechanism, and the sequence of operations described above repeated to produce a new batch of radioactive material. If the material to be produced for a second batch is different than the material produced during the first batch, then it may be necessary to either replace cone 46 or to take other suitable steps to avoid potential contamination. While in the discussion above it has been assumed that there is no window at the junction 76, and this would be true for the .sup.18 F reaction discussed above which results in very low vapor pressure which can be dissipated by the vacuum, where the target material and reaction to generate a particular isotope results in a higher vapor pressure, a window may be required at juncture 76 to avoid contaminating the vacuum in accelerator 78. However, where a solid target is utilized, it is possible to maintain a vacuum or near vacuum in chamber 12 and thus to minimize the pressure differential across the window. Therefore, while the problem of dissipating heat from the window still exists with a solid target, the stresses on the window resulting from high pressure differentials thereacross are substantially eliminated, resulting in far less problems with window damage and thus far less maintenance overhead. While the discussion above has been primarily with reference to the generating of .sup.18 F radioisotopes, it is apparent that the teachings of this invention could be utilized to generate many other commonly used radioisotopes, including carbon-11, nitrogen-13 and oxygen-15. For example, oxygen 15 could be generated with a frozen nitrogen-14 target bombarded with deuterons, nitrogen-11 with a frozen carbon target such as frozen CO.sub.2, etc. The teachings of this invention might also be utilized, if desired, to generate certain stable isotopes such as .sup.15 N or .sup.5 Li. Further, while a cone has been shown as the target surface for a preferred embodiment, it is apparent that other angled surfaces, for example an angled flat surface, could be utilized. However, the cone shape is clearly advantageous in that it provides optimum surface area and also facilitates the collection of the melted radioisotope-containing target material. Also, while having an angled surface is advantageous in permitting the use of a thinner ice layer to achieve a given yield, an angled target surface is not an essential limitation on the invention and some of the advantage of having a cryogenic target for isoptope generation can be achieved with targets shaped and positioned such that all or a substantial part of the target are at angle perpendicular to the high energy particle beam. In addition, while melting the isotope containing ice target and extracting the resultant droplet is the preferred method of isotope extraction, other techniques might also be utilized to extract the isotope. For example, target 46 could be heated under conditions to cause sublimation of the ice, the ice evaporating or vaporizing to a gas which then may be removed from the chamber, for example through extra port 30. Where the isotope is to be mixed or dissolved in some other substance, it may also be possible to simply remove the cone with the ice layer adhering thereto and dipping the frozen cone in the higher temperature liquid or gas in which the isotope is to be utilized, the ice melting and simultaneously going into solution. The two techniques discussed above would be particularly advantageous where a target surface other than a cone was being utilized. Such techniques might also permit a simplification of the equipment shown in FIG. 1 in that rotating bellows assembly 70 would not be required, nor would rotation of the portion of the device to the right of pivot point 72 be requred during the extraction process. It may also be possible to eliminate the rotation step by initially orienting the cone vertically, and either also mounting the accelerator to be vertical or preferably bending the particle beam to properly impinge on the target. While several methods of extraction have been discussed above, it is apparent that such techniques are only illustrative of techniques available for extracting the ice target material from the target surface after the desired radioisotope or other isotope has been formed therein, and it is the intent that such other extraction techniques also be included within this invention. Other changes in the details of construction are also possible. Thus, while the invention has been particularly shown and described above with reference to a preferred embodiment, the foregoing and other changes in form and detail may be made therein by one skilled in the art while still remaining within the spirit and scope of the invention. |
H00005088 | abstract | Hybrid-drive implosion systems (20,40) for ICF targets (10,22,42) are described which permit a significant increase in target gain at fixed total driver energy. The ICF target is compressed in two phases, an initial compression phase and a final peak power phase, with each phase driven by a separate, optimized driver. The targets comprise a hollow spherical ablator (12) surroundingly disposed around fusion fuel (14). The ablator is first compressed to higher density by a laser system (24), or by an ion beam system (44), that in each case is optimized for this initial phase of compression of the target. Then, following compression of the ablator, energy is directly delivered into the compressed ablator by an ion beam driver system (30,48) that is optimized for this second phase of operation of the target. The fusion fuel (14) is driven, at high gain, to conditions wherein fusion reactions occur. This phase separation allows hydrodynamic efficiency and energy deposition uniformity to be individually optimized, thereby securing significant advantages in energy gain. In additional embodiments, the same or separate drivers supply energy for ICF target implosion. |
abstract | A passive cooling system of a nuclear power plant includes a steam generator, a cooling water storage tank, a water cooling heat exchanger, an air cooling heat exchanger, a divergence valve, and a cooling tower. The steam generator generates steam by heat exchange with a primary coolant system, and the cooling water storage tank stores cooling water therein. The water cooling heat exchanger is disposed in the cooling water storage tank, and the air cooling heat exchanger is connected to the steam generator. The divergence valve is controllable to divert steam from the steam generator into both the water cooling heat exchanger and the air cooling heat exchanger. Each of the cooling water storage tank, the water cooling heat exchanger, and the air cooling heat exchanger are located in the cooling tower. |
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claims | 1. Apparatus for thermally insulating a vessel during normal operating conditions and cooling the vessel in the event that an accident condition has a potential for raising the vessel temperature above a predefined limit, comprising:a vessel having a bottom head and a sidewall extending upwardly therefrom;thermal insulation surrounding the vessel and spaced from the bottom head and the sidewall; andat least one inlet door hingedly attached to the thermal insulation, the inlet door being sealed by gravity against a stop on the thermal insulation during normal operation to close an inlet to a space between the vessel and the thermal insulation and floatable with a given buoyancy when immersed in a liquid fluid to move to an open position to permit the liquid fluid to gain access to the space between the vessel and the thermal insulation under predetermined abnormal operating conditions to cool the vessel without the inlet door providing an obstruction to flow of the fluid in the open position. 2. The apparatus of claim 1 wherein the vessel has an axial centerline and the bottom head has an outside diameter and wherein the vessel bottom head and the thermal insulation define an annulus that increases in volume from the axial centerline to the outside diameter of the bottom head. 3. The apparatus of claim 1 wherein the inlet door is designed so that when the fluid enters the space between the thermal insulation and the vessel the inlet door will fail in an open position if it loses its buoyancy. 4. The apparatus of claim 1 wherein the vessel is a nuclear reactor pressure vessel that is supported in a reactor cavity suspended above a floor of the reactor cavity, the reactor cavity having upwardly extending side walls surrounding at least a lower portion of the vessel and spaced from the vessel, the thermal insulation being interposed between the side wall of the reactor cavity and the vessel and spaced from the side wall of the reactor cavity to define a cooling air path therebetween and the space between the vessel and the thermal insulation defining a liquid fluid flow path therebetween. 5. The apparatus of claim 4 wherein the fluid flow path and the cooling air path have separate exits from the reactor cavity. 6. The apparatus of claim 4 further including:a neutron shield for shielding at least some of a stream of neutrons from exiting the reactor cavity, the neutron shield surrounding and spaced from the vessel and spaced from the side wall of the reactor cavity so that the fluid flow path and the cooling air path respectively pass by the inner and outer surface of the neutron shield; anda fluid exit door hingedly attached above the neutron shield, the fluid exit door having a closed position and an open position, the closed position sealing the fluid flow path from an exit flow path from the reactor cavity and the open position providing the fluid flow path access to the exit flow path from the reactor cavity without forming an obstruction to air flow in the cooling air path. 7. The apparatus of claim 6 wherein the fluid exit door when open to permit the fluid cooling the vessel to exit remains in the open position until reset. 8. The apparatus of claim 6 wherein the vessel has nozzles disposed within a nozzle gallery that surrounds the vessel at a nozzle elevation wherein the fluid exit door permits the fluid to exit to the nozzle gallery. 9. The apparatus of claim 1 wherein the inlet door has vent holes to prevent pressure differentials between the interior and the outside of the door during vessel heat-up and cool-down. 10. The apparatus of claim 1 wherein the door is a metal shell filled with a hydrophobic material. 11. The apparatus of claim 4 further including:a neutron shield for shielding at least some of a stream of neutrons from exiting the reactor cavity, the neutron shield surrounding and spaced from the vessel to provide an opening for the fluid flow path;a second set of insulation surrounding an upper portion of the vessel above the fluid flow path and the neutron shield; anda fluid exit door hingedly attached to the second set of insulation at an upper end of the exit door, the fluid exit door sealing an exit to the fluid flow path when in a closed position. 12. The apparatus of claim 11 wherein the fluid exit door is a metal enclosure filled with a buoyant hydrophobic material. 13. The apparatus of claim 8 including a plurality of fluid exit doors hingedly attached to the neutron shield at a plurality of circumferential locations both at and between positions of the nozzles. 14. The apparatus of claim 13 wherein the fluid flow path and the cooling air path have separate exits from an upper portion of the reactor cavity and the cooling air path only exits the upper portion of the reactor cavity approximately at the nozzle positions. 15. The apparatus of claim 14 wherein the fluid flow path exits the reactor cavity approximately at and between the nozzle positions. 16. The apparatus of claim 13 wherein the plurality of fluid exit doors extend substantially continuously, circumferentially around the vessel. 17. The apparatus of claim 1 wherein the thermal insulation surrounding a lower portion of the vessel, has a rounded transition extending from below the bottom head to around the sidewall of the vessel, substantially following a contour of the vessel. |
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050911406 | abstract | A replacement heater sleeve for nuclear reactor coolant system pressurizer and method for replacing a damaged original heater nozzle. The original heater and heater nozzle are removed from the pressurizer and the original bore is enlarged. An outer sleeve is installed in the enlarged bore on the same center as the original heater nozzle and is substantially flush with the interior of the pressurizer. The outer sleeve is welded to the pressurizer on its interior and exterior surfaces. An inner sleeve is installed in the inner diameter of the outer sleeve and extends beyond the outer sleeve into the pressurizer. The inner sleeve is welded to the lower end of the outer sleeve and is provided with an inner diameter sized to receive a heater of the same size as that originally installed in the pressurizer. The inner sleeve also maintains the original heater alignment in the pressurizer. |
053430484 | abstract | A contour collimator for shaping a radiation beam for radiation therapy has two sets of radiation-impermeable lamellae disposed opposite each other, with each lamella being individually displaceable toward and away from its counterpart lamella in the other set. The sets of lamella are disposed between lateral walls of holder, at least one of which can be loosened and tightened. The lamellae are normally urged against each other by a spring force acting on each lamella. A user-manipulable mechanism is operable for displacing all of the lamellae in both sets away from each other, so that a form conforming to the desired contour can be placed between the lamellae. The mechanism is then moved to permit the spring force to act on each lamella, causing the lamellae to abut the form. The lamellae are then clamped in position. |
abstract | A scintillation crystal can include a sodium halide that is co-doped with thallium and another element. In an embodiment, the scintillation crystal can include NaX:Tl, Me, wherein X represents a halogen, and Me represents a Group 1 element, a Group 2 element, a rare earth element, or any combination thereof. In a particular embodiment, the scintillation crystal has a property including, for radiation in a range of 300 nm to 700 nm, an emission maximum at a wavelength no greater than 430 nm; or an energy resolution less than 6.4% when measured at 662 keV, 22° C., and an integration time of 1 microsecond. In another embodiment, the co-dopant can be Sr or Ca. The scintillation crystal can have lower energy resolution, better proportionality, a shorter pulse decay time, or any combination thereof as compared to the sodium halide that is doped with only thallium. |
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abstract | In an immersion lithographic apparatus, bubble formation in immersion liquid is reduced or prevented by reducing a gap size or area on a substrate table and/or covering the gap. |
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claims | 1. A fuel assembly (1) for a nuclear pressurized water reactor, said fuel assembly comprising:a lower end structure (6);an upper end structure including a top nozzle (5);a plurality of elongated fuel rods (2);a plurality of longitudinally extended guide thimbles (3) extending in a longitudinal direction from the lower end structure (6) to the top nozzle (5), said top nozzle (5) including a passageway extending through the top nozzle (5) and an annular groove (10) formed in said passageway;an elongated sleeve (11) formed by each of the plurality of longitudinally extended guide thimbles (3), the sleeves (11) each being configured for attaching a portion of the guide thimble (3) to the top nozzle (5) and each having at least three slots (12) in an upper end portion of said sleeve (11), each slot (12) extending downwardly from a top end of said sleeve (11) anda plurality of locking members (15), each locking member (15) being introduced into and positioned within a respective one of the elongated sleeves (11) from above the top nozzle (5),wherein:the upper end portion of the sleeve (11) includes at least three bulges (13), each of which said at least three bulges (13) seat in said annular groove (10) when said sleeve (11) is in an expanded locked position within said passageway;each of said at least three bulges (13) has two ends and extends circumferentially between a nearest two of the slots (12);wherein the two ends of each of said at least three bulges (13) are circumferentially spaced from each other; andat least one of the two ends of each of said at least three bulges (13) is located at a position which is at a spaced circumferential distance (d) from a nearest one of the two slots (12). 2. The fuel assembly according to claim 1, wherein each of said bulges (13) has an end portion at a respective end, wherein the end portion has a curved shape in a longitudinal section and in a transversal section along the bulges (13). 3. The fuel assembly according to claim 1, wherein both ends of each of said bulges (13) extend to a position at a distance (d) from the respective slot (12). 4. The fuel assembly according to claim 1, wherein said bulges (13) are provided so that there is at least one bulge (13) between each pair of adjacent slots (12). 5. The fuel assembly according to claim 1, wherein said bulges (13) are provided circumferentially after each other. 6. The fuel assembly according to claim 1, wherein:the sleeve (11) has a wall thickness seen in a radial direction with respect to the longitudinal direction; andthe distance (d) between the end of each of said at least three bulges (13) and the nearest slot (12) is at least equally long as the wall thickness of the sleeve (11). 7. The fuel assembly according to claim 1, wherein:the sleeve (11) has a wall thickness seen in a radial direction with respect to the longitudinal direction; andthe distance (d) between the end of each of said at least three bulges (13) and the nearest slot (12) is two or three times longer than the wall thickness of the sleeve (11). 8. The fuel assembly according to claim 1, wherein:the sleeve (11) has a wall thickness seen in a radial direction with respect to the longitudinal direction; andthe wall thickness of the sleeve (11) is in the range of from 0.20 to 0.50 mm. 9. The fuel assembly according to claim 1, wherein the sleeve (11) has at least one of 3, 4, 5 or 6 slots (12). 10. The fuel assembly according to claim 1, wherein at least one of the bulges (13) has a cylindrical profile seen in a longitudinal section. |
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claims | 1. A method for synthesizing a lithium-titanium oxide using a solid state method, comprising:mixing lithium oxide (Li2O) and titanium oxide (TiO2) in a solvent;separating a solid material which includes lithium oxide and titanium oxide from the solvent;drying the solid material separated from the solvent; andperforming a heat treatment on the solid material,wherein a molar ratio of lithium oxide to titanium oxide is 1:0.940 or more to less than 1:1. 2. The method of claim 1, wherein a molar ratio of lithium oxide to titanium oxide is in the range of 1:0.940 to 1:0.944. 3. The method of claim 2, wherein the titanium oxide has an anatase crystal structure. 4. The method of claim 1, wherein titanium oxide has a rutile crystal structure and a molar ratio of lithium oxide to titanium oxide is 1:0.942. 5. The method of claim 1, wherein the heat treatment is performed at 600° C. or more to less than 800° C. 6. The method of claim 1, wherein the heat treatment is performed at 670° C. or more to less than 800° C. 7. The method of claim 1, wherein the heat treatment is performed for 12 hours or more. 8. The method of claim 1, wherein the titanium oxide has an anatase crystal structure. 9. The method of claim 1, wherein lithium-titanium oxide prepared in the performing of the heat treatment has a Li2TiO3 structure. 10. The method of claim 1, wherein the solvent includes an alcohol. |
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claims | 1. A process automation system for determining, monitoring or influencing different process variables or state variables in at least one manufacturing or analytical process, comprising:at least one control station; anda plurality of field devices, which measure, monitor or influence the process variables or state variables; wherein:in each said field device, at least one sensor for ascertaining a measured value of a process variable or state variable, or one actuator for influencing a process variable or a state variable by means of an actuating value, is provided;each said field device makes available to every other field device of the process automation system cyclically or acyclically ascertained, measuring-device-specific, measured values or actuating values of the process variable or state variable, so that all ascertained measured values or actuating values of the manufacturing or analytical process are available as information to each said field device; andsaid current information for all ascertained measured values or actuating values of the process variables or state variables is available to each said field device as a current process-state-vector. 2. The process automation system as claimed in claim 1, further comprising:a digital fieldbus that said field devices use to communicate with said control station and one another, wherein:each of said field devices, continuously or cyclically, jointly reads measured values or actuating values provided from said other field devices of the manufacturing or analytical process on said digital fieldbus and stores the jointly read, measured values or jointly read, actuating values as information at predetermined positions in the process-state-vector. 3. The process automation system as claimed in claim 1, wherein:an individual one of said field devices transmits the field-device-specific, measured values and/or field-device-specific actuating values via a two-wire-connecting line or a fieldbus to said control station cyclically or upon request of said control station; andsaid control station transmits collected information cyclically or up request of said field devices to all field devices as the process-state-vector. 4. The process automation system as claimed in claim 1, further comprising:a control or evaluation unit in said field devices or said control station, wherein:said control or evaluation unit applies said information of the process-state-vector for reviewing plausibility of current measured values or current actuating values of the current process variable or state variable ascertained with the field device or for function-diagnosis of the field device. 5. The process automation system as claimed in claim 1, further comprising:a control or evaluation unit in said field devices or said control station, wherein:said control or evaluation unit ascertains, from the information of the process-state-vector, life expectancy of said field device or need for maintenance of said field device. 6. The process automation system as claimed in claim 1, further comprising:a control or evaluation unit in said field devices or said control station, wherein:said control or evaluation unit analytically or numerically derives from the information of the process-state-vector at least one other, indirect, measured value characterizing the process. 7. The process automation system as claimed in claim 1, wherein:in use of a process automation system in at least two manufacturing or analytical processes, a characteristic variable characterizing the manufacturing or analytical process is associated with said process-state-vector and serves for identification or for grouping of all ascertained measured values of said field devices from the same manufacturing or analytical process. 8. The process automation system as claimed in claim 1, wherein:in said process-state-vector, a time stamp is provided for characterizing a point in time of ascertaining the measured values. 9. The process automation system as claimed in claim 1, wherein:the individual measured values or actuating values of the process variables or state variables are arranged at predetermined locations in said process-state-vector. 10. The process automation system as claimed in claim 1, wherein:a priority designation of the measured values or the actuating values is provided, which fixes rank of a record of the measured value or the actuating value in said process-state-vector in the case of a plurality of measured values or actuating values for a particular process variable. 11. The process automation system as claimed in claim 1, further comprising:a control- and evaluation-unit encrypts the information of said process-state-vector. 12. The process automation system as claimed in claim 1, wherein:in each said field device or in said control station, a historical information of older measured values or older actuating values is stored with the current information for current measured values or current actuating values in a process-state-matrix. 13. The process automation system as claimed in claim 12, wherein:said control or evaluation unit ascertains, from the current information and the historical information, life expectancy or need for maintenance of said field device or reviews plausibility of current measured value or current actuating values of the current process variable or state variable ascertained with said field device or ascertains function-diagnosis of said field device or trend of the measured values or actuating values. |
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056420145 | summary | BACKGROUND OF THE INVENTION 1. Field of invention This invention relates to a beta voltaic power source integrated with a substrate as a power source for integrated circuits formed on the substrate. 2. Description of related art Radio isotopic power sources convert radiation from radioactive isotopes directly into electrical energy. Devices, such as artificial cardiac pacemakers, utilize the radio isotopic power sources for sustained long term power which allow the devices to function for many years without any other source of energy. Tritium is an isotope of hydrogen having a half life of 12.5 years. Because tritium emits only beta particles and the intensity of the beta particles is limited, tritium is an excellent source of radiation for radio isotopic power source applications. Beta voltaic power sources incorporate tritium together with a pn junction to directly convert the emitted beta particles into electrical energy. The beta particles emitted by the tritium is absorbed by the pn junction generating electron-hole pairs. The electron-hole pairs are separated by the built in electric field of the pn junction producing an electric current. Relatively high efficiencies are possible because each high energy beta particle produces many electron-hole pairs. Current applications of the beta voltaic power source are in the form of a battery component. The battery is connected to a separate device such as the artificial cardiac pacemaker. SUMMARY OF THE INVENTION An object of the invention is to provide a self-powered device integrating a radioactive power source with integrated circuits including an least one substrate, at least one radioactive power source formed over the at least one substrate generating electric current, and integrated circuits formed over the at least one substrate. The integrated circuits are adapted to receive power from the radioactive power source. The radioactive power source includes a first active layer having a first conductivity type formed over the substrate. An active layer is a semiconductor doped with an impurity to form either a p-type or n-type region. The substrate has a second conductivity type. A second active layer having the second conductivity type is formed over the first active layer forming a depletion region at the boundary between the first and second active layers. The interface between the first and second active layers forms either a pn or an np junction. A tritium containing layer is provided which supplies beta particles to the depletion region. A metal tritide layer is an example of the tritium containing layer. Another embodiment of the self-powered device includes an integrated circuit substrate and at least one cap substrate. The integrated circuit substrate includes a plurality of integrated circuits and at least one power source portion. Each of the power source portion includes a first active layer having a first conductivity type formed over the integrated circuit substrate and a second active layer having the second conductivity type formed over the first active layer. The cap substrate includes a fourth active layer having the first conductivity type formed over a bottom surface of the cap substrate. The cap substrate has the second conductivity type. A fifth active layer having the second conductivity type is formed over a top surface of the cap substrate. The cap substrate is placed over a corresponding power source portion on the integrated circuit substrate. A tritium containing layer is placed between the cap substrate and the power source portion. The cap substrate, the power source portion and the tritium containing layer together form a beta voltaic power source. When several of the beta voltaic power sources are connected either in series and/or in parallel, a wide range of voltage and current values can be obtained. The beta voltaic power source of the self-powered device is enhanced by trench structures formed by the first, second or fourth active layers. The trench structures allow the beta particles to be more efficiently converted into electric current. Another object of the invention is to provide a method for producing the self-powered device. The method includes providing at least one substrate, forming at least one radioactive power source over the substrate and forming integrated circuits over the substrate. The radioactive power source is provided by forming a metal layer and diffusing tritium into the metal layer. The metal layer is comprised of metal that forms stable metal tritides with tritium. |
claims | 1. A method for processing radioactive waste, comprising:a step of extracting a group of isotopes having a same atomic number from the radioactive waste, followed by no separation process of isotopes from the extracted group, the extracted group including a radionuclide and stable nuclides; anda step of generating a neutron (n) by an accelerator and irradiating the neutron (n) to the isotopes, so as to produce nuclear transmutation of a first nuclide as a long-lived radionuclide into a second nuclide as a stable nuclide, while suppressing nuclear transmutation of a third nuclide into the first nuclide by setting a value of irradiation energy of the neutron (n) within such a range that a (n, 2n) reaction cross section of the first nuclide is equal to or larger than 10 times as large as a (n, 2n) reaction cross section of the third nuclide,andwherein the isotopes, the first nuclide, the second nuclide, and the third nuclide are defined as below:selenium (Se) isotopes, Se-79, Se-78, and Se-80, respectively;palladium (Pd) isotopes, Pd-107, Pd-106, and Pd-108, respectively;zirconium (Zr) isotopes, Zr-93, Zr-92, and Zr-94, respectively;krypton (Kr) isotopes, Kr-85, Kr-84, and Kr-86, respectively; orsamarium (Sm) isotopes, Sm-151, Sm-150, and Sm-152, respectively. 2. The radioactive waste processing method according to claim 1, further comprising:after the extracting step and before the irradiating step, a step of increasing or decreasing a ratio of isotopes having an odd number of neutrons with respect to isotopes having an even number of neutrons in the extracted group of isotopes, by inducing an isotopic shift phenomenon in the extracted group of isotopes. |
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040381358 | abstract | The fuel element comprises a core having a base of ceramic fuel material enclosed between two metallic cladding plates and constituted by the juxtaposed array of a plurality of ceramic fuel wafers. At least a number of wafers are provided with individual metallic protection which performs a contributory function in the cladding of the wafers and in the division of the fuel element into compartments. |
050892215 | claims | 1. A spacer for use in a nuclear fuel bundle having a plurality of fuel rods, said spacer comprising in combination: a plurality of spring metal spacer cells, each cell having; at least one spring leg, said spring leg inwardly deflected at the medial portion thereof for spring contact with a fuel rod within said spacer cell; at least two rod encircling arms affixed at remote ends of said spring legs; each rod encircling arm defining stops for abutting a fuel rod whereby said spring leg can bias an encircled fuel rod within said cells into said stops; said rod encircling arms having differential length including a first portion of said arms having a relatively longer length and a second portion of said arms having a relatively shorter length; said cells confronted into cell pairs with said spring legs remote from one another; said rod encircling arms affixed to one another adjacent the end of said rod encircling arms; each said arm of one cell at one end thereof fastens to a rod arm of an adjacent cell with said relatively longer rod arm of one cell of said cell pair fastening to a relatively shorter arm of the other cell of said cell pair whereby said cell pair forms a unitary rigid substructure; a unitary grid formed from said cell pairs, each cell pair joined to adjoining cell pairs by welding at said rod encircling arms; a continuous Zircaloy band surrounding said unitary grid, said Zircaloy band including tabs for deflection to said rod encircling arms of said cell units for inhibiting vertical movement of said band with respect to said unitary grid whereby said band is keyed to said grid against vertical movement with respect to said unitary grid. and means configured in said band for engagement to said unitary grid at said corners. a plurality of spring metal spacer cells, each cell having; at least one spring leg, said spring leg inwardly deflected at the medial portion thereof for spring contact with a fuel rod within said spacer cell; at least two rod encircling arms affixed at remote ends of said spring legs; each rod encircling arm defining stops for abutting a fuel rod whereby said spring leg can bias an encircled fuel rod within said cells into said stops; said rod encircling arms having differential length including a first portion of said arms having a relatively longer length and a second portion of said arms having a relatively shorter length; said cells confronted into cell pairs with said spring legs remote from one another; said rod encircling arms affixed to one another adjacent the end of said rod encircling arms; each said arm of one cell at one end thereof fastened to a rod arm of an adjacent cell with said relatively longer rod arm of one cell of said cell pair fastening to a relatively shorter arm of the other cell of said cell pair whereby said cell pair forms a unitary rigid substructure; a unitary grid formed from said cell pairs, each cell pair joined to adjoining cell pairs by welding at said rod encircling arms. each rod encircling arm affixed to said spring leg having an upper rod encircling arm of short length overlying a lower rod encircling arm of short length and an upper rod encircling arm of long length overlying a lower rod encircling arm of long length. 2. The invention of claim 1 and including means for stiffening said band configured to the sides of said band. 3. The invention of claim 2 and wherein said unitary grid is formed of square section with four corners; 4. The invention of claim 2 and wherein said stiffening means includes raised portions in said band for imparting stiffness to said band. 5. The invention of claim 4 and wherein said raised portions are disposed vertically on said band. 6. The invention of claim 4 and wherein said raised portions are disposed horizontally on said band. 7. The invention of claim 1 and including a metal band surrounding said Zircaloy band, said metal band welded to spring metal spacer cells. 8. The invention of claim 7 and wherein said metal band surrounding said Zircaloy band is spring metal. 9. The invention of claim 8 and wherein said metal band surrounding said Zircaloy band welds to said spring metal spacers through apertures defined by said Zircaloy band. 10. A spacer for use in a nuclear fuel bundle having a plurality of fuel rods, each spacer comprising in combination: 11. The invention of claim 10 and wherein: |
047568682 | description | DETAILED DESCRIPTION OF THE INVENTION Referring to FIG. 1, an operating and securing mechanism is housed in a tubular enclosure 10 made from several parts assembled together. The enclosure projects through the cover 12 of the tank of the reactor, constructed so as to withstand the pressure of the primary coolant of the reactor. The enclosure accomodates a rod 14 for driving the control bar, connected to the latter by a releasable coupling, (not shown). The rod has, in the vicinity of its top end, an enlargement 15 forming a piston, having sealing rings 16. A linear 18 is fitted inside enclosure 10 and is fixed, at its upper part, to a second liner 24. It has an inner diameter such that rings 16 slide over the liner substantially sealingly. The outer diameter of liner 18 is such that an annular space 20, opening into the tank, exists between liner 18 and enclosure 10. The liner 18 extends sufficiently high for the segments to remain in contact therewith, even when the rod occupies the high position, shown in FIG. 1. When in that high position, the drive rod 14 is in abutment against a transverse cam-forming pin 22 placed across the enclosure. The second liner 24 is fitted inside the enclosure 10, in the extension of liner 18. With the enclosure it defines a second annular space 26 which communicates with space 20 through radial grooves 28. The upper part of the enclosure 10 is closed by a sealing plug 30 in which is formed a drain hole 32 closed by a screw 33. During filling of the primary circuit of the reactor, it may happen that air is trapped in the pressurized enclosure of the mechanism. A collecting chamber 35, having a volume considerably greater than that occupied by the air at normal service pressure, is provided under the plug. The mechanism of the invention comprises, first, a disengageable device for securing the control rod in the high position and, second, a step-by-step lifting device. The two devices will be described successively. SECURING DEVICE The securing device is located within liner 24. It comprises electromagnetically actuated means having a fixed unit and a movable unit. The fixed unit comprises an operating coil 36 placed outside the enclosure and coaxially with this latter and a fixed pole piece 38 fixed inside liner 24. The movable unit includes a pole piece 40 movable between a low position (FIG. 1) and a high position in which it is in abutment against the fixed pole piece 38. The two pole pieces have axial passages 42 and 44 formed therein for communicating the volume defined by the rod and liner 24, which forms a decompression chamber 46, with the volume situated above the fixed pole piece 38. A return spring 47 tends to bring and maintain the fixed pole piece 40 in the low position in which it is shown in FIG. 1. The movable pole piece 40 forms both a distributing valve and a releasable gripping mechanism for securing the rod, it has two sealing rings 48 axially spaced apart from each other and sealingly slidable on the internal face of liner 24. Holes 50 are formed through liner 24 at a location such that they open between rings 48 when the pole piece 40 is in its low position and so that they communicate the annular space 26 with the decompression chamber 46 when the pole piece 40 is raised, i.e. when the operating coil 36 is energized. So that the pole piece 40 may operate as a gripping mechanism, it has flexible downwardly directed blades 52 formed with end gripping catches. These catches have a shape such that the blades may be moved apart by a bulge 54 on rod 14 when the latter arrives at its upper end of travel, but such that they cannot be spread apart by the weight of the rod. Means are provided for forcibly opening the blades by raising the movable pole piece 40. As shown in FIG. 1, these include pin 22 and inward 14 directed bosses 56 on the blades. The bosses 56 are placed so as to be immediately below pin 22 when the movable pole piece 40 is in the low position, and to come in alignment with the pin when the movable pole piece 40 is raised. MECHANISM FOR DISPLACING THE ROD The mechanism for displacing the rod comprises essentially an electromagnetically controlled piston pump having a fixed part and a mobile part. The fixed part of the pump comprises a coil 58 similar to coil 36 and a fixed pole piece 60 inside liner 24. Sealing between liner 24 and pole piece 60 is provided by one or more rings. The movable part of the pump is formed by a pole piece 62 movable between a rest position (FIG. 1) and a work position into which the fixed pole piece 60 piece 62 is moved into the work position by coil 50a when the latter is energized. A return spring 64 compressed between the pole pieces biases piece 62 to its low position. Passages 66 are formed in the movable pole piece 62 for balancing the fluid pressures exerted on the surfaces thereof. The movable pole piece 62 has an extension or reduced diameter, forming a plunger 68 having sealing rings which projects into a recess in the fixed pole piece 60 for defining a pumping chamber 70. The hydraulic circuit of the pump comprises an axial passage 72 in the plunger, provided with an intake non-return check valve 74, connecting the pumping chamber 72 with the volume situated below the movable pole piece 62 and so to the decompression chamber. This circuit further comprises, starting from the pumping chamber 70, a duct 76 provided with a delivery check valve 78. This duct 76 is extended by one or more holes formed in liner 24, so that duct 76 is in permanent communication with the inside of the tank of the reactor. The intake and delivery valves may be of conventional construction and formed for instance by a ball urged by a return spring onto a seat. It will be appreciated that the pump thus formed is wholly integrated within liner 24, except for the operating coil 58. OPERATION Operation of the mechanism will now be described with reference to FIGS. 2A-2B, in which only the main elements of the mechanisms have been shown, the scale not being respected for the scale of increased clarity. LIFT OF THE DRIVING ROD The driving rod is moved up step-by-step, each step corresponding to an energization-de-energization cycle of coil 58. Each energization of coil 58 moves the pole piece 62 from the position shown in FIG. 2A to the position in abutment against the fixed pole piece 60. The contents of the pumping chamber 70 is discharged through duct 76 and the annular spaces 26 and 20 into the tank. The lifting movement causes an upwardly directed force on rod 14. If the leaks are negligible, the force is equal to the product of the force exerted on the movable pole piece 62 multiplied by the ratio S.sub.1 / S.sub.0, S.sub.1 being the cross-sectional area defined by the rings of piston 15 and S.sub.0 the cross-sectional area of plunger 68. The stroke of rod 14 is equal to the stroke of the movable pole piece 62 multiplied by the ratio S.sub.0 /S.sub.1. When coil 58 is de-energized, the movable pole piece 62 moves back to its rest position. Liquid flows to the pumping chamber 70 from the decompression chamber 46 through the intake valve 74. Since downward movement of the mobile pole piece 62 takes place without variation of the volume offered to the liquid above rod 14, the latter remains in place. Consequently, each energization-de-energization cycle of coil 58 causes upward movement of the drive rod 14 by a height equal to the product of the stroke of the movable pole piece by the ratio S.sub.0 /S.sub.1. SECURING During the final portion of the lifting stroke of the drive rod 14, that rod penetrates between blades 62. The prestress of spring 47 is selected at a value greater than the vertical force which bulge 54 must exert to spread blades 52 apart. Consequently, bulge 34 passes beyond the blades without driving the movable pole piece 40. As soon as the lower shoulder of bulge 54 has moved beyond the end catches of the flexible blades 52, the latter close again and prevent the drive rod from moving down, as shown with continuous line in FIG. 1. A sensor is advantageously provided for supplying a signal when the drive rod has passed beyond the level of the catches. The sensor is, for example, a flexible blade switch (not shown) placed in a sealed bulb outside enclosure 10 for actuation by a permanent magnet mounted at the upper part of rod 14. The signal supplied by the sensor may be used for stopping the repetition of the energization-de-energization cycle of coil 58. If there is no leak at the sealing rings or ball valves, the drive rod 14 remains in its high position, in abutment against pin 22, as shown in dot dash lines in FIG. 2A. If there are leaks, rod 14 moves slowly down until it bears on the catches of the flexible blades 52 whose profile is such that the weight of the drive rod and control bar tends to close the blades further, as shown in dot-dash lines in FIG. 1. The control bar is thus held in a position where it is outside the core. RELEASE The release operation will generally take place while the parts are in the initial position shown in dot-dash lines in FIG. 1. Then rod 14 is first of all raised by operating the lift mechanism until the sensor (not shown) indicates that the rod is clear of the catches of blades 52. Then the control coil 36 is energized to raise the movable pole piece 40. During the upward movement, the flexible blades 52 are moved apart by pin 22 and the decompression chamber 46 is connected to the tank through the annular spaces 20 and 26. The drive rod and the control bar then move slowly down under the action of their weight, the falling speed being determined by the head loss which is adjustable by metering the flow area of holes 50 appropriately. Numerous modifications are possible, in respect of the securing means and the rod displacement means. FIGS. 3A and 3B show schematically a modification of the securing means allowing the pump to be used for accelerating downward movement of the rod and the control bar. In FIGS. 3A and 3B, the parts corresponding to those in FIG. 1 are designated by the same reference numbers. The displacement means are unchanged. The securing means again comprises a movable pole piece 40. On the other hand, it comprises, in addition to the sealing rings 48, two rings 80 each placed close to a ring 48, but at a smaller distance from the middle of the pole piece. In the pole piece 40 are formed two internal ducts 82 and 84. Duct 82 connects the space between the upper pair of rings 48 and 80 and the space between rings 80. Duct 84 connects the central hole 42 of pole piece 40 (which then is a blind hole) to the space defined by the lower pair of rings 80 and 48. As shown in FIGS. 3A and 3B, the annular space 26 is separated by a seal 8 into a top part and a bottom part. Holes 50 connect the low part to a zone of the bore in liner 24 which is located between rings 80 when the pole piece is at rest (FIG. 3A), facing duct 84 when this pole piece comes into abutment against the fixed pole piece 38 following energization of the control coil (FIG. 3B). Transfer holes 88 connect the top part 26A to a zone of the bore in liner 24 situated between rings 80 whatever the position of the mobile pole piece 40. Finally, an internal transfer channel 90 formed in the body of liner 24 places the decompression chamber 46 in communication with a zone of the bore of liner 24 which is situated above rings 48 and 80 when the mobile pole piece 40 is at rest (FIG. 3A), between rings 48 and 80 of the upper pair when the control coil 36 is energized (FIG. 3B). In the modification shown in FIGS. 3A and 3B, lifting of the bar is provided by means of a pump while maintaining the mobile pole piece 40 at rest (FIG. 3A). The downward movement of the bar is caused by bringing the mobile pole piece 40 into the position shown in FIG. 3B. It can be seen that the downward movement occurs, as in the case shown in FIG. 1, if the pump is maintained at rest, but through valves 74 and 78. When it is desired to accelerate the downward movement of the bar, it is sufficient to activate the pump. In both cases, the water flow takes place along the path indicated by arrows f1 in FIG. 3B, which may be compared with the lifting path indicated by arrows f0 in FIG. 3A. Whatever the embodiment used, it must be initially filled with pressurized liquid coolant after it has been assembled. For that purpose, the control coil 36 is energized for communicating the decompression chamber 46 with the annular space 26. Filling of the pumping chamber 70 then takes place when the pressure of the primary coolant is increased. Valve 74 opens to allow this filling. Then the collecting chamber 36 is filled through the annular space 26 and ducts 76 and may be scavenged by opening the screw 33. |
description | FIG. 1 shows diagrammatically an X-ray examination apparatus 1 in accordance with the invention. The X-ray source 2 emits an X-ray beam 15 for irradiating an object 16. Differences in X-ray absorption within the object 16, for example a patient to be radiologically examined, give rise to an X-ray image formed on an X-ray sensitive surface 17 of the X-ray detector 3, which is arranged opposite the X-ray source. The X-ray detector 3 of the present embodiment is formed by an image intensifier pick-up chain which includes an X-ray image intensifier 18 for converting the X-ray image into an optical image on an exit window 19 and a video camera 23 for picking up the optical image. The entrance screen 20 acts as the X-ray sensitive surface of the X-ray image intensifier which converts X-rays into an electron beam which is imaged on the exit window by means of an electron optical system 21. The incident electrons generate the optical image on a phosphor layer 22 of the exit window 19. The video camera 23 is coupled to the X-ray image intensifier 18 by way of an optical coupling 24, for example a lens system or a fibre-optical coupling. The video camera 23 extracts an electronic image signal from the optical image, which signal is applied to a monitor 25 for the display of the image information in the X-ray image. The electronic image signal may also be applied to an image processing unit 26 for further processing. Between the X-ray source 2 and the object 16 there is arranged the X-ray filter 4 for local attenuation of the X-ray beam. The X-ray filter 4 comprises a large number of filter elements 5 in the form of capillary tubes whose X-ray absorptance can be adjusted by application of an electric voltage, referred to hereinafter as adjusting voltage, to the inner side of the capillary tubes by means of the adjusting unit 7. The adhesion of the X-ray absorbing liquid to the inner side of the capillary tubes is adjusted by means of this electric voltage. One end of the capillary tubes communicates with a reservoir for an X-ray absorbing liquid. The capillary tubes are filled with a given quantity of X-ray absorbing liquid as a function of the electric voltage applied to the individual tubes. Because the capillary tubes extend approximately parallel to the X-ray beam, the X-ray absorptance of the individual capillary tubes is dependent on the relative quantity of X-ray absorbing liquid in such a capillary tube. The electric adjusting voltage applied to the individual filter elements is adjusted by means of the adjusting unit 7, for example on the basis of brightness values in the X-ray image and/or the setting of the X-ray source 2. For this purpose, the adjusting unit 7 is coupled to the output terminal 40 of the video camera and to the power supply 11 of the X-ray source 2. The construction of an X-ray filter 4 of this kind and the composition of the X-ray absorbing liquid are described in detail in International Patent Application WO 96/13040. FIG. 2 is a side elevation of an X-ray filter 4 of the X-ray examination apparatus of FIG. 1. The Figure shows seven capillary tubes by way of example, but a practical embodiment of an X-ray filter 4 of an X-ray examination apparatus in accordance with the invention may comprise a large number of capillary tubes, for example 40,000 tubes in a 200xc3x97200 matrix arrangement. Each of the capillary tubes 5 communicates with the X-ray absorbing liquid 6 at one end 31. The inner side of the capillary tubes is covered by an electrically conductive layer 37, for example of gold, platinum or aluminium, which layer 37 is coupled to a voltage line 34 via a switching element 33. For application of the electric adjusting voltage to the electrically conductive layer 37 of a capillary tube, the relevant switching element 33 is closed while the voltage line 34 is supplied with the desired electric adjusting voltage. The switching elements are driven by a control line 35. When brief voltage pulses having a length of a few tens of microseconds are used, adjusting voltages in a range of from 0 V to 400 V can be used. In this voltage range, switches in the form of xcex1-Si thin-film transistors can be used. Preferably, an adjusting voltage in the range of from 0 V to 100 V is used. Because the voltage pulses are so brief, the application of the adjusting voltage does not cause any, or hardly any, electrolysis of the X-ray absorbing salt solution used as the X-ray absorbing liquid. The salt solution may, for example, comprise a Lead salt or Caesium Chloride salt solution. The X-ray absorptance of the individual capillary tubes can be controlled on the basis of the level of the electric adjusting voltage applied to the capillary tubes. On the electrically conductive layer there is preferably provided a dielectric layer of a thickness sufficient to ensure that the electric capacitance of the capillary tubes remains low enough to enable fast response to the application of the electric voltage. However, the shorter the switch on time, the more significant becomes the electrical response time of the capillaries. A coating layer having suitable hydrophobic properties may also be provided on the dielectric layer. FIG. 3 is a plan view of an X-ray filter 4 of the X-ray examination apparatus shown in FIG. 1. An X-ray filter 4 comprising 16 capillary tubes in a 4xc3x974 matrix arrangement is shown by way of example. However, in practice the X-ray filter 4 may comprise a much larger number of capillary tubes, for example 200xc3x97200 tubes. Each of the capillary tubes is coupled, by way of the electrically conductive layer 37, to the drain contact 40 of a field effect transistor 33 which acts as a switching device and whose source contact 41 is coupled to a voltage line which supplies the adjust voltage. The transistors 33 together define an array of switching devices on a common substrate, for example manufactured using thin film technology. However, the capillaries 5 can not be formed on this substrate and are therefore formed as a separate array of filter elements. The invention is concerned with the connection between these two arrays. The array of transistors is coupled using edge connectors to the adjusting unit 7, which may comprise integrated circuit driver chips, and which therefore cannot be formed on the same substrate as the array of transistors 33. For each row 9 of capillary tubes there is provided a control line 35 which is coupled to the gate contacts of the field effect transistors in the relevant row in order to control the field effect transistors in this row. The control line 35 of the relevant row is energized by an electric control voltage pulse, in order to enable an adjusting voltage to be applied to the electrically conductive inner side of the capillary tubes in the row. The field effect transistors in the relevant row are electrically turned on during the control voltage pulse. The control signals are provided by the adjusting unit 7, which comprises a voltage generator 27 for applying an electric voltage to the timer unit 8 which applies the control voltage pulses having the desired duration to the individual control lines of the rows of capillary tubes. While the relevant field effect transistors are turned on, i.e. the switching elements are closed, the electric adjusting voltage of the relevant control lines 34 is applied to the capillary tubes. The level of the adjusting voltage applied to individual capillary tubes in a row can be differentiated by application of different electric adjusting voltages to the respective voltage lines 34 of individual columns. To this end, the adjusting unit 7 comprises a column driver 36 which controls the application of the electric adjusting voltage generated by the voltage generator 27 to the individual voltage lines. Each of the voltage lines 34 is coupled to a respective switching element, for example a transistor 44. When the transistor 44 of the voltage line 34 is turned on by energizing the gate contact of the relevant transistor by means of a gate voltage, the adjusting voltage is applied to the voltage line. The gate contacts of the transistors 44 are coupled, via a control unit 45, to the voltage generator 27 which supplies the gate voltage. The adjusting voltages are also supplied by the control unit 45. FIG. 4 shows the array 50 of switching devices, for example thin film transistors. The array is provided on a glass substrate 52, and each thin film transistor on the substrate is provided with an external connection portion 54 which is in electrical contact with the drain 40 of the associated thin film transistor 33 (shown in FIG. 3). The connecting portions 54 thereby define an array which overlies the array of thin film transistors 33. The edges 56 of the substrate 52 are provided with edge connections 58 which fan out from the transistor array. The array 50 of transistors 33 is arranged in rows and columns, and a small number of rows and columns are represented in FIG. 4 for the purposes of clarity. Each connection portion 54 is arranged as a metal node or bump over the drain pad of each of the thin film transistors 33. These nodes may be formed using a wire bonding machine which ultrasonically bonds a wire ball bond onto the drain pad, for example a gold wire. The wire is then broken off above the ball bond, to form a gold bump over the drain pad of each transistor. All of these bumps are then planarized to give a flat-topped structure. Additional height may be applied to these external connection portions by bonding additional bumps onto the top of the first layer. The array of external connection portions provide an external interface to the array 50 which does not require the use of conducting tracks around the edges of the substrate 52. The use of a gold wire bond is preferred as a result of the resistance of gold to oxidation. However, an aluminium wire bond may be used, preferably with the wire bonding taking place in an inert (e.g. nitrogen) local atmosphere. It may also be possible to use a solderable material, for example by depositing a low melting point solder on to the drain pads of the TFT array. This could be reflowed to form solder bumps, with a subsequent reflow process giving rise to the conductive mechanical contact between the TFT array and the capillary array. FIG. 5 shows a cross section through one of the thin film transistors 33. Transistor 33 comprises a top-gate TFT having lower source 41 and drain 42 patterns. The transistor body 44 spans the gap between the source and drain 41, 42 and preferably comprises an amorphous silicon semiconductor layer. In the example shown in FIG. 5, the source 41 and drain 42 comprise interlocking spiral portions, which enables a TFT to be formed with a very high width to length ratio of the gate. Thus, in the example of FIG. 5, the drain comprises a central drain pad 42a, from which a spiralling limb 42b extends. The source 41 comprises an interleaved spiral pattern. A gate insulator layer 48, for example silicon nitride, overlies the transistor body 44 and a gate contact layer 46 is provided over the gate insulator layer 48 to define the top-gate structure. A well 49 is provided in the gate insulator layer 48 over the drain pad 42a to enable an external contact to be made with the drain pad 42a. The well 49 is metallized with a region 46a of the gate conductor layer, and a further drain contact 40 is provided in the well 49 to which the external connection portion 54 may be bonded. FIG. 6 shows one example of the array of filter elements for use in the filter of the invention. The capillary tubes 5 are arranged in a honeycomb structure 60 which defines a network of the capillaries. The honeycomb network 60 is formed from a series of parallel membranes 62. Preferably, each membrane 62 includes two layers, and the honeycomb network is formed by selectively separating the two layers which form each membrane 62 to define the honeycomb configuration. Each membrane 62 is effectively associated with a row of capillaries 5, and conducting lines 64 are provided on the surface of the membranes 62 and which lead to individual capillaries 5 and which then form the electrically conductive layer 37 on the inner surface of each capillary. In FIG. 5, conducting lines 64 provided on one face of one membrane are shown and which define the conductive layers 37 for three sides of the hexagons of a partial row of capillaries. Each capillary has two conductive layers 37 which together provide coverage on all six faces of the internal surface of the capillaries. Thus, two conducting lines are associated with each capillary, and these terminate at an end block 66. For the example shown in FIG. 6, it is necessary, at least for some of the capillaries, for the two conducting lines to be provided on separate membranes 62, so that two contacts are required to address those capillaries. Conducting lines may be provided on opposite sides of the membranes 62, and even between the two layers of the individual membranes, to enable all the required conductive layers 37 to be provided. The membranes 62 may comprise flexible foils, for example PETP (polyethylene terephthalate) plastics foils, which provide some flexibility to the overall structure. The end block 66 is defined by heat sealing the ends of the foils to provide a rigid connection interface. The end surface of this block is metallographically ground and polished to reveal an array of aluminium track cross sections, for example as shown in FIG. 7. In the schematic example shown, each membrane 62 is provided with tracks 64 only on opposite sides of the membrane 62. A heat sealing member 68 is provided between the membranes 62 in order to form the end block 66. The member 68 may comprise extra blank membranes 62 inserted into the gaps. A metal deposition layer may be provided over the end faces of the aluminium control lines 64 in order to provide a larger uniform metal bond pad over each individual track in the end block 66. This deposition may be performed by a mask to provide discrete bond pads, or may be performed by deposition over the entire face of the end block followed by laser ablation patterning. To ensure good electrical contact, it may be necessary to selectively etch back the foil material of the end block to expose more of the underlying aluminium tracks. This can be achieved either by chemical etching or by gas plasma etching, followed by aluminium oxide removal. Instead of heat sealing the foils into a rigid end block, the loose foils may be interleaved with thin glass layers 70 as shown in FIG. 8, or fibre reinforced epoxy polyimide. These glass layers are patterned with corresponding tracks 72 to those on the foils, but significantly thicker. The glass plates are then stacked face to face with the tracks on the foil surfaces, as shown in FIG. 8. The whole foil-glass assembly is then clamped into one block consisting of a number of foil-glass couples, where each foil and glass plate contain a number of tracks. The end of this clamped assembly is then metallographically ground and polished. The cross sections of the tracks on the glass are then patterned with bond pads as explained with reference to FIG. 7. FIG. 9 illustrates the manner in which the prepared end block of the array of filter elements is joined to the array of external connection portions 54 (the nodes) of the control circuit array 50. The array 50 of transistors 33 is inverted and the nodes 54 are dipped into an isotropic conductive adhesive. The control circuit array is then aligned and the adhesive-coated nodes are bonded to the array of bond pads on the end block 66 of the array of filter elements. The gap between the surface of the control circuit array and the surface of the end block can be filled with a reinforcing epoxy underfill material, to provide additional mechanical and environmental protection. Instead of an isotropic conductive adhesive, an isotropic conducting film may be laid between the two surfaces to be bonded, with pressure being applied under temperature causing reflow and curing of the film. The film comprises a dispersion of conductive particles suspended in a plastic adhesive matrix. The conductive particles provide the electrical connection between the TFT array and the capillary end block at the positions where localised pressure has been applied, namely at the gold bumps. This arrangement provides electrical and mechanical connection between the array of transistors of the control circuit and the array of filter elements. The transistor array may be defined so that the spacing between transistors corresponds to the spacing between the contact pads in the end block of the array of filter elements. Thus, as shown in FIG. 9, each external connection portion 54 is aligned with a contact pad 65 formed over the end of a control line 64. Matching the spacing between transistors 33 in the control circuit array and the spacing between membrane 62 emerging from the capillary array avoids the need for complicated reshaping of the membranes 62 from the capillary array. The transistor array 50 will therefore have dimensions corresponding to the part of the end block 66 carrying the control lines 64. The transistor array 50 may be segmented into a discrete number of sub-arrays, sharing the glass substrate 52. This may be desirable to enable the end block 66 to be divided into the same discrete number of end block portions. The reason for adopting this approach is that the effective thermal cycle strain of the interface between the end block and the transistor array is proportional to the diagonal length of the end block. Therefore, dividing the end block into a number of portions results in the interface suffering less cyclic strain than for the single connection described above. The thermal strain of the interface between the two arrays may also be reduced by attempting to match the temperature coefficients of expansion of the two arrays. The glass inserts 70 described with reference to FIG. 8 assist in this respect. In the example of the array of filter elements shown in FIG. 6, the membranes 62 which form the end block 66 also define the capillaries 5 themselves. However, it is alternatively possible for the membrane 62 of the end block 66 to be interleaved with further membranes 63 which define the capillary array. The membranes of the end block may still comprise flexible foils, or they may comprise alternative structures. For example, in the arrangement shown schematically in FIG. 10, the membranes 62 comprise thin metallized glass plates. The plates are patterned with metallic contact pads 71. The glass plates are inserted into the open slots at the side of the array of filter elements, with their contact pads aligned with contact pads from the control line 64 leading to the individual capillaries. Each membrane 62 has tracks on opposite sides but which are connected to the same contact pad 73 on the top surface of the glass plate. When all the plates are in place and aligned, the whole assembly is clamped to be held in place. The contact pads 73 form a regular array of pads of the same size and pitch as the gold bumps on the transistor array. The transistor array is then bonded on top of the clamped glass plate assembly in the same manner as described previously. Since each glass sheet 62 bridges the gap between an associated pair of the additional membranes 63, it is possible to provide an arrangement in which only an individual contact pad 73 is required for each capillary. In this arrangement the end block and the transistor array are predominantly glass. Consequently, the coefficient of thermal expansion of the two surfaces is almost identical. Resultant thermal cycle strain experienced by the interconnections between the two surfaces will be of a very low magnitude. Although a wirebonding or soldering technique has been described for forming the external connection portions, other techniques may be employed, provided a contact is made available at the location of the switching devices. The switching device need not be transistors, arrays of other switching devices may equally be appropriate. All references cited herein, as well as the priority document Great Britain Patent Application 9902252.7 filed Feb. 3, 1999, are incorporated herein by reference in their entirety and for all purposes to the same extent as if each individual publication or patent or patent application was specifically and individually indicated to be incorporated by reference in its entirety for all purposes. |
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041749995 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings wherein identical reference numerals have been used in the several views to identify like elements, FIG. 1 shows an exploded view of a nuclear reactor vessel 10. While the vessel 10 may be fabricated in differing ways, the overall cylinder which results from welding together the several smaller cylinders 12 is used herein as an illustrative example for purposes of this descrption. The several welds, A-K, shown in FIG. 1 are typically those which are to be inspected together with the stud holes 34 and the ligament areas 35 therebetween of the vessel top flange 13. It will be understood by those familiar with the inspection code requirements for nuclear reactor vessels, that not all of the welds A-K or the top flange 13 are necessarily inspected at the end of one time period, but that any inspection apparatus therefor must be capable of efficiently and accurately determining the integrity of the vessel welds A-K and its top flange 13, at one time or in predetermined code specified groupings. Further, the inspection apparatus must be accurately positioned to accomplish vessel interrogation without harming the top flange 13 and, in particular, its ability to form a proper seal with the vessel header (not shown). FIG. 2 depicts an illustrative example of an inspection site. The inspection apparatus 14 is shown therein seated in the reactor vessel 10. Prior to inspection, the apparatus 14 is assembled using an erection rig 5 partially shown. After assembly, the inspection apparatus 14 is lowered by the site work crane 7 into the reactor vessel pool 11 and into the vessel 10. The work bridge 9, also partially shown in FIG. 2, can be utilized as necessary. The inspection apparatus 14 is illustrated in FIG. 3. It comprises a quick-disconnect lifting assembly 16, a support ring 18 having an annular key 19 attached thereto, three support legs 20A, 20B and 20C, a head support assembly 22, a main column 24, a manipulator arm 26, a transducer array 28 and an overall control system 30 which includes an assortment of motors, resolvers and cabling, and is mainly resident in a console 31. These main elements cooperate, in a manner to be more specifically described hereinafter, to permit inspection of the reactor vessel 10 in accordance with code requirements. The inspection apparatus 14 is adapted to be lowered into the reactor vessel 10 and is shown in two of its many possible inspection positions in FIGS. 9 and 10. Prior to insertion of the inspection apparatus 14, the reactor vessel header is removed and tapered guide studs 32, having chamfered heads 33, are inserted into three of the stud holes 34 which have been designated for that purpose. With the guide studs 32 in place, the inspection apparatus 14 is fully lowered into the reactor vessel 10 and positively seated therewithin, as shall be hereinafter explained. The guide studs 32 are engaged by guide stud bushings 36 which are movably mounted to the support ring 18. Accurate circumferential positioning of the inspection apparatus 14 is accomplished through the employment of the guide studs 32 and guide bushings 36 in conjunction with a specially adapted support leg show 88, to be more fully described below. The clearance between the guide stud bushings 36 and the guide studs 32 is typically a maximum of only 3/8". Therefore, it is of critical importance that the inspection apparatus 14 be lowered into the reactor vessel 12 with the guide studs 32 and bushings 36 in near perfect alignment with each other. Alternatively stated, the inspection apparatus 14, which is a relatively heavy piece of equipment, must be closely aligned with respect to the vertical and horizontal axis or the guide bushings 36 will be cocked with respect to the guide studs 32 causing the inspection apparatus 14 to hang up thereon, which might result in damage to the inspection apparatus 14, the guide studs 32 or the reactor vessel 10. Thus, the lifting assembly 16 must be adjustable to accommodate the cantilevered weighting effect of manipulator arm 26 and/or any weight distribution disparity in the inspection apparatus 14 which would cause it to tilt from a level attitude as it is being lowered. In addition, with the inspection apparatus 14 in place, the lifting assembly 16 must be quickly and readily removable to allow use of the inspection site work bridge 9 should that be necessary. The lifting assembly 16 is shown in greater detail in FIGS. 6, 7 and 8. When secured to the inspection apparatus 14, it is engaged by the site crane 7 which connects to the "U" bolt assembly 42 coupled to its uppermost portion. The "U" bolt assembly 42 is shown in FIG. 3. A cylindrical collar 44 having a stepped, star-shaped or cloverleaf bore 46 is bolted to the top of the head support assembly 22. The crane 7 now lowers the lifting assembly 16 until a spider 48 enters bore 46, as shown in FIG. 6. The lifting assembly is then manually rotated about 45.degree., so that the splines of spider 48 are positioned to engage the hidden or dotted line portion of bore 46 as is shown in FIG. 7. The crane 7 now raises the lifting assembly 16 until the spider 48 abuts the upper surface of the stepped portion of bore 46 at which point it is engaged by and in the collar 44, as is shown in FIG. 8. At this point in the procedure of connecting the lifting assembly 16 to the inspection apparatus 14, the feet 60 of the ball and socket assemblies 50 are held about 3/16" above the leveling pads 52. The leveling pads 52, as shown in FIG. 3, are connected to the upper portions of the support legs 20. A hydraulic cylinder 54 is now actuated, causing its internal piston 56 to push against a fixed surface 58 forcing the socket feet 60 into tight engagement with the leveling pads 52. Three non-adjustable base struts 65 are utilized to enhance the structural rigidity of the lifting assembly 16 and are connected between the three ball and socket assemblies 50, as is shown in FIG. 3. As connected, the base struts form a triangle, the center of which is coincidental with the central axis of the lifting assembly 16 and the main column 24. The inspection apparatus 14 is thereby fully secured to the lifting assembly 16 and is now suspended from the crane 7 for alignment procedures prior to being seated in the reactor vessel 10. Such alignment procedures are necessary due to probable repositioning of the movable guide stud bushings 36 from site-to-site to accommodate differing locations of the guide studs 32. In addition, the position and extension of manipulator arm 26 may be different from an inspection start at one site than at another. Further, the vessel locating key 62, shown in FIGS. 3 and 14, may or may not be in use. Consequently, the net effect of these and other possible causes will be to present the inspection team with a different weight distribution at each inspection site, thereby necessitating the alignment procedure. Finally, even if the same weight distribution was expected, proper inspection technique would demand alignment verification. The alignment procedure is carried out by turning one or more of the turnbuckle struts 64 which are rotatably adjustable and fixedly connected between the three ball and socket assemblies 50 and the slidable sleeve 66 of the hydraulic cylinder 54. Adjustment of the turnbuckle struts 64 has the effect of gimbaling the inspection apparatus 14 about the center axis of the triangle formed by the base struts 65 or the lower end of the lifting assembly 16. This enables the inspection team to plumb the main column 24 of the inspection apparatus 14 and verify its vertical alignment. In addition, each of the three ball and socket assemblies 50 can be individually adjusted to shift the position of the end of the turnbuckle strut 64 connected thereto to effect adjustment of the inspection apparatus 14 with respect to both the vertical and horizontal axes. Horizontal alignment is verified by checking the level on any one of the three leveling pads 52. The lifting assembly 16 is capable of being quickly disconnected by reversing the order specified above. First, the hydraulic cylinder 54 is deactivated causing its outer sleeve 66 to move upwards lifting the socket feet 60 from the leveling pads 52. The crane 7 now lowers the lifting assembly 16 by an amount sufficient to allow the spider 48 to fall out of engagement with the upper portion of the bore 46 of collar 44. Spider 48 can now be rotated and withdrawn from collar 44. After this is done, the entire lifting assembly can be removed by the crane 7, freeing it for other work, and leaving the inspection apparatus 14 seated in the reactor vessel 10. Alternatively, the lifting assembly 16 can be so disconnected after it has been used to remove the inspection apparatus 14 from the reactor vessel 10, leaving the inspection apparatus 14 on resting pads (not shown) or on the erection rig 5 preparatory to shipment. By removing the lifting assembly 16 with the inspection apparatus 14 still seated in the reactor vessel 12, the work bridge 9 can be moved across the vessel pool 11 allowing for the performance of other maintenance or inspection procedures or to assist in the vessel inspection itself. Referring again to FIG. 3, there is shown three support legs 20A, 20B and 20C. Each of these is joined to the head support assembly 22 by a spacer 68 which is of a length appropriate to the diameter of the vessel to be inspected. It should be noted that for differing diameter reactor vessels, the spacers 68 and the support ring 18 are selected and sized so that the guide stud bushings 36 extend radially to a point where they will be aligned with and then engage the guide studs 32. Very small variations in radial dimensions are accommodated by loosening the guide stud bushing clamps 70 and inserting shims of an appropriate thickness which would have the effect of moving the center of the guide stud bushings 36 radially outward of support ring 18 as desired. It should also be noted that the guide stud bushing clamps 70, when loosened, permit movement of the bushings 36 along the support ring 18 to accommodate variations in the placement of the guide studs 32 in the vessel top flange 13 at different inspection sites. As previously noted, the guide stud bushings 36 are movably connected to the support ring 18 by the bushing clamps 70. As was also previously noted, the support ring 18 carries an annular key 19 about its outer surface. A keyway 71, see FIG. 4, cut in the surface of clamp 70, which mates with support ring 18, accommodates key 19 and aligns the guide stud bushings 70 on support ring 18 with respect to the remainder of the inspection device 14. In addition, support legs 20A, 20B and 20C are connected to support ring 18 respectively by a bracket 90, see FIG. 5, having a keyed hole 92 therethrough. Thus, the bracket 90 engages the key 19 and positively locates and locks the support legs 20A, 20B and 20C to support ring 18 which enhances the structural stability of inspection apparatus 14. The leveling pads 52 are bolted or welded to the upper segments of the support legs 20A, 20B and 20C in horizontal alignment and are utilized in the manner described above. When seated within the reactor vessel 10, the three legs 20A, 20B and 20C support the entire weight of the inspection apparatus 14. Stainless steel shoes 84 are bolted respectively to the bottom of support legs 20B and 20C. These shoes reset either on the circumferential vessel flange 15 or on the core barrel flange (not shown), depending on whether the core barrel has been removed. A special "A" shaped shoe 88 is bolted to the end of support leg 20A and is adapted to accurately position inspection apparatus 14 as it is being seated within the reactor vessel 10. With the core barrel remaining in the vessel 12, a plate 94 having a keyway 96 cut therein is bolted to shoe 88 as shown in FIGS. 12 and 13. As it is being seated, keyway 96 engages a head-to-vessel alignment pin, the position of which is known, and positively locates the inspection apparatus 14 within the vessel 10. As mentioned above, the clearance between the guide studs 32 and the guide stud bushings 36 is about 3/8" and their engagement yields a coarse circumferential alignment. The subsequent engagement by keyway 96 of the head-to-vessel alignment pin yields a fine circumferential alignment which provides for an absolutely certain placement of the inspection apparatus 14 within vessel 10. With the core barrel removed for inspection, the plate 94 is removed from shoe 88 and a vessel locating key 62, as shown in FIGS. 3 and 14, is bolted to shoe 88 in its place. The vessel locating key 62 fits into a notch 17 cut in the circumferential vessel flange 15, see FIG. 11, which notch is otherwise covered by the core barrel flange. This engagement of notch 17 by the vessel locating key 62 provides the same fine circumferential alignment means, with the core barrel removed, as was yielded by the use of plate 94. It should be noted that plate 94 can be built up with appropriately configured shims to accommodate the different sized head-to-vessel alignment pins that may be encountered from one vessel to another. Thus, by choice of the shoe 88 configuration, in conjunction with the engagement of the guide studs 32 by the guide stud bushings 36, the exact circumferential location of support leg 20A, and derivatively that of manipulator arm 24, is known and assured. In addition, this positive location or seating of the inspection apparatus is accomplished without touching or threatening the sealing surface of the vessel top flange 13. Connected immediately below the head assembly 22, as shown in FIG. 3, is a gear box and motor assembly 72 which drives manipulator arm 26 vertically along the main column 24 utilizing a pulley system 75. The main column itself consists of several sections of flanged pipe bolted together. Sections may be readily removed or added to accommodate the depth of reactor vessel inspection requirements. Each section is individually encased so that water cannot enter therein. Between the flanges 85 thereof, the sections of main column 24 carry a track 78 which is used, in conjunction with a sensor to be hereinafter described, to determine the extent of vertical travel or, alternatively stated, to fix the vertical position of manipulator arm 26. The main column sections also include "U" shaped grooves 80 which accommodate bearings carried by manipulator arm 26. The grooves 80 and flanges 85 combine functionally to restrain the manipulator arm 26 from making any unwanted or undesirable rotary movements about the main column 24 as it travels therealong. The manipulator arm 26, which is more clearly shown in FIG. 15, includes a carriage assembly 82 which rides on the main column 24 in the "U" shaped grooves 80. The carriage assembly 82 and the remainder of manipulator arm 26 would typically be fabricated from a low weight material which can withstand the hostile operating environment. The carriage assembly 82 is fitted with internally mounted and sealed ball bearings which ride in and are engaged by the "U" grooves 80 and facilitate vertical movements by manipulator arm 26 on the main column 24. When the vertical drive motor (not shown) in the vertical motor assembly 72 is actuated, it rotates the drive pulleys 74 and 77 as is shown in FIG. 16. A pulley cable 79 is looped about the carriage idler pulleys 76 and 81 and the head assembly idler pulleys 83 and 87. When the vertical motor shaft 89 is rotated counterclockwise, the pulley cable 79 is released respectively by both drive pulleys 74 and 77 from their take-up spools 91 and 93, lowering the manipulator arm 26 with equal force on both sides of the carriage assembly 82. This equalization of the release force applied to both carriage idler pulleys 76 and 81 insures that the carriage assembly will not be cocked and therefore hang-up or unduly wear its bearings as it travels down the main column 24. Likewise, when the vertical motor drive shaft 89 is rotated clockwise, the upward or lifting forces applied to the carriage idler pulleys 76 and 81 is equalized and the carriage assembly 82, as well as the remaining elements of manipulator arm 26, is lifted smoothly, at the proper attitude, up the main column 24. The head assembly idler pulleys 83 and 87 serve to define the upper portion of the pulley cable loop. This upper portion of the cable loop is utilized to equalize any cable slippage or unbalance in the cable 79 which might otherwise unequally tend to pull up on or release idler pulleys 76 and 81. Thus, except for any movement to effect compensation due to an unbalance, the pulley cable 79 is in motion during vertical travel of the manipulator arm 26 only between the drive pulleys 74 and 77 and the carriage idler pulleys 76 and 81 respectively. An emergency cable clip 99 is secured to the cable 79 between the head assembly idler pulleys 83 and 87. If the pulley cable 79 should happen to snap, the clip 99 will become wedged between one of the idler pulleys 83 or 87 and its respective support bracket 95 or 97, thereby restraining further vertical movement of manipulator arm 26. An emergency braking system 100 is shown in FIGS. 17, 18 and 19. It serves to halt vertical movement of the manipulator arm 26 whenever its vertical speed of travel exceeds a predetermined velocity, typically a speed greater than five inches per second. A vertical velocity rate error signal is developed utilizing a signal generated by the Z axis resolver 102 which engages the vertical track 78 and thereby follows and helps to determine the vertical position and rate of change therein of the manipulator arm 26. When an overspeed condition is sensed by the control system 30, an emergency brake signal is forwarded to three pneumatic cylinders 104 mounted beneath the carriage assembly 82. The pneumatically operated piston 106 of each cylinder 104 is connected via a header 108 to a brake shoe 110. The brake shoe 110 is fitted with spring loaded roller bearings 112 which ride in bearing slots 114 in the brake shoe 110 and are normally urged against the "U" grooves 80 of the main column 24. In the rest position illustrated in FIG. 17, the emergency braking system 100 is disabled and the bearings 112 are spring loaded against the " U" groove 80 holding the brake shoe 110 in its rest position and avoiding unnecessary wear. A cross-sectional view of the brake shoe 100 and brake lining 116 is shown in FIG. 19. When the emergency brake signal is received by the pneumatic cylinders 104, the pistons 106 thereof are thrust upwardly at a speed significantly in excess of that exhibited by the manipulator arm 26, even in its overspeed condition. This rapid piston movement forces the wedge shaped brake shoe 110 upwardly into contact with the brake lining 116 which is bolted to the bottom of the carriage assembly 82. As the brake shoe 110 fully contacts the positionally fixed brake lining 116, as is shown in FIG. 18, the roller bearings 112 are forced inwardly in slots 114 and the brake shoe 110 becomes jammed against the "U" groove 80 halting further vertical movement of the manipulator arm 26. As noted above, the speed of piston 106 is significantly greater than the overspeed limit of the manipulator arm 26. It is therefore fast enough, when actuated, to overtake the manipulator arm 26 and cause braking action to occur even when the overspeed condition of manipulator arm 26 results from upward movement thereof. Thus, the described emergency braking system 100 functions to halt vertical movement of manipulator arm 26 when an overspeed condition occurs regardless of the direction of vertical or Z axis travel at that time. To insure absolute downward restraint of manipulator arm 26, an emergency stop plate 117, as depicted in FIGS. 3 and 15, is bolted to the bottom section of main column 24. Plate 117 serves to halt downward movement of manipulator arm 26 should the emergency braking system 100 fail to function properly. The manipulator arm 26 is thereby prevented, by either the emergency braking system 100 or the stop plate 117, from striking the bottom of the reactor vessel 10 or any portion thereof as it is vertically driven in the vessel 10. An axis motion or rotation of the manipulator arm 26 about the main column 24 is shown in FIG. 15. As illustrated therein, actuation of the A axis motor 118 drives the carriage rotary gears 122 and 124 causing the entire manipulator arm to swing about the main column 24. The position of manipulator arm 26 in the A axis is verified by a signal which is generated by the rotary resolver 120. It should be noted with respect to all of the drive motors described herein, whether shown or not, that a resolver or position determining sensor is coupled thereto to provide a signal which is then employed to indicate the position of manipulator arm 26 or any portion thereof, in or about the particular axis of movement associated with the motor being described. Y axis movement, which is also indicated in FIG. 15, is achieved by driving a set of telescoping arms 126 and 128, which are movable mounted within the carriage channels 130, toward and away from the carriage assembly 82. As is more clearly illustrated in the end view shown in FIG. 17, the Y axis motor 132 is coupled by its shaft 134 to a drive gear 136. When the Y axis motor 132 is actuated, it causes drive gear 136 to be rotated, driving a rack 138 engaged thereby, which rack is bolted to the telescoping arm 126. This causes arm 126 to be driven towards or away from the carriage assembly 82, depending on the direction of rotation of the Y axis motor 132. When the outer telescoping arm 126 is moved, it carries with it an idler gear 140 which is meshingly engaged between rack 142, which is attached to the inner telescoping arm 128, and rack 144 which is coupled to the carriage channel 130. For purposes of clarity, the illustration in FIG. 17 depicts only one half of the telescoping arrangement of the Y axis drive, but it will be understood that the Y axis motor 132 causes, through the action of another drive gear (not shown), both sets of telescoping arms 126 and 128 to be driven in a desired direction along the Y axis. Movement of the manipulator arm 26 along the Y axis is required, in particular, to position the transducer array 28 within any one of the reactor vessel nozzles 38 for inspection thereof, as is shown in FIG. 9. In the event of total power failure or an inability to withdraw the transducer array 28 from within a nozzle 38, an emergency retraction assembly 140 is provided. As is depicted in FIG. 21, the emergency retraction assembly 140 includes a retraction cable 142 arranged within the carriage assembly 82 and extending therefrom to be looped about an idler pulley 144 which is rotatably mounted within and to the telescoping arm 126. Cable 142 also is guided by the half-pulley 149. One end of the retraction cable 142 is fixedly secured to the carriage assembly 82 by a clamp 146. The other end of the retraction cable 142 is formed into a ring 150 which is detachably secured to the carriage assembly 82 at an initial position 152 by a removable clamp 148. The ring 150 is mounted so as to be accessible from above. When an emergency retraction of the transducer array 28 becomes necessary, a hook (not shown) is lowered into the reactor vessel 10 to engage the cable ring 150. Once engaged, the ring 150 is pulled up, which action frees the detechable clamp 148 from the carriage assembly 82. Upward force is maintained, moving the cable ring 150 from its initial position 152 towards its final position 154. As the cable ring 150 is pulled upwards toward its final position 154, the retraction cable 142 forces the pulley 144 from its initial position 156 to its final position 158. Since the pulley 144 is secured to the outer telescoping arm 126, it forces it back into the carriage channel 130 as it moves towards its final position 158. Simultaneously, the outer telescoping arm 126 causes the inner telescoping arm 128 to be moved inwardly, through manual operation of the Y axis drive described above, thereby forcibly withdrawing the manipulator arm 26 and the transducer array 28 from within a vessel nozzle 38. B axis motion is obtained by actuating the B axis motor (not shown) which is mounted within the B axis drive housing 160 and connected to mounting bracket 178. As is more clearly illustrated in FIG. 22, the B axis drive housing 160 is secured in the following manner. A mounting bracket 162 is bolted to each of the inner telescoping arms 128. Attached to the upper end portion of bracket 162 is an apertured dog ear 164. Attached to the upper portion of the B axis drive housing 160 is a movable linkage assembly 166 which is actuated by a locked-over-center lever 168. The linkage assembly 166 terminates in a dog 170 which engages the aperture in dog ear 164 when lever 168 is moved to its locked position 172 and holds the B axis drive housing in a normal position with respect to the telescoping arm 128. The bottom portion of the B axis drive housing is movably secured by engagement with a hinge pin 174. As noted above, the transducer array 28 and manipulator arm 26 can be withdrawn from a vessel nozzle 38 in an emergency situation. However, it may not yet be safe to lift the inspection apparatus 14 from the vessel 10 since the forward portion of the manipulator arm may strike the reactor vessel 10. Accordingly, after the manipulator arm 26 has been manually retracted, the hook is again lowered and engages the linkage lever 168. As the hook and lever 168 are pulled upwardly, the linkage assembly 166 extracts the dog 170 from engagement with the dog ear 164, allowing the B axis drive housing to rotate about hinge pin 174 as is shown in phantom in FIG. 22. With the B axis drive housing in its final position 176, the entire inspection apparatus 14 can be withdrawn from the vessel 10 without any fear of striking the vessel walls. Further movement of the transducer array 28 is possible along or about five additional axes of movement. In addition to movement of the manipulator arm 26, and derivatively movement of the transducer array 28, along or about the A, B, Y and Z axis, movement can be effected about the C, D, E, F and G axes. The B axis motor shaft is connected to a mounting bracket 178 and, when driven, rotates bracket 178 and all elements connected forwardly thereof about the B axis. Two additional mounting brackets 180 and 182 are secured to the B axis motor bracket 178, as is shown in FIGS. 3 and 15. The C axis motor housing 184 is coupled between and secured to the brackets 180 and 182 with the C axis motor shaft 186 extending through and being drivingly engaged by the brackets 180 and 182. When actuated, the C axis motor drives its shaft 186 and the brackets 180 and 182, as well as all of the manipulator elements connected forwardly thereof, about the shaft 186. Motion in the D axis is achieved in a similar manner. The D axis motor housing 188 is also coupled between and secured to the brackets 180 and 182 with the D axis motor shaft 190 extending through and being drivingly engaged by the brackets 180 and 182. When the D axis motor is actuated to drive its shaft 190, motor shaft 190 and all of the manipulator arm elements connected forwardly thereof are rotated in the D axis. The E axis motor housing 192 is connected to the C axis motor housing 188 with the E axis motor shaft (not shown) being connected to mounting bracket 194. When actuated, the E axis motor shaft drives bracket 194 about the E axis, as well as all of the manipulator arm elements connected forwardly thereof. The F axis motor housing 196 is secured by mounting bracket 194 and by mounting brackets 198. The shaft 200 of the F axis motor (not shown) extends through and drivingly engages the mounting bracket 198. When actuated, the F axis motor drives its shaft 200 and the remainder of the manipulator arm elements connected forwardly thereof through F axis motion. The G axis motor housing 202 is secured to the end of mounting bracket 198. The G axis motor shaft 204 extends outwardly of housing 202 and is clamped into the transducer plate collar 206 which, in turn, is clamped to the transducer array plate 40. When actuated, the G axis motor drives its shaft 204 and the transducer array plate about the G axis. Thus, the transducer array plate 40 and the transducer array 28 mounted thereon, with reference to any point in the reactor vessel 10, can be driven in nine planes of movement or about nine axes of rotation. This highly mobile and segmented articulating drive train can be employed to accurately position the transducer array 28 at any point within the reactor vessel 10. Ordinarily, electrical connection to and from the different motors, resolvers and the transducer array 28 would be accomplished by means of components particularly suited for use in an underwater operating environment. To avoid the use of such special components, which are more expensive and require longer delivery times, it was decided to pressurize the electrical cabling system allowing for the use of ordinary components. For example, the junction box 208, shown only in FIGS. 15 and 23 for purposes of clarity, can be pressurized to a degree which would prevent water seepage therein and thusly allow the use of standard electrical connectors. In order to conserve on cabling, the air supply and electrical supply was combined in the cabling assembly 210, shown in FIG. 23. The illustration in FIG. 23 is merely representative of the cabling assembly 210 and only one cable 212 and one dual cable 214 has been shown, although more are used. The electrical cable 212 carries a plurality of electrical conductors to and from the console 31 which would typically include the control system 30. These conductors would be utilized to energize the different motors and transducers and carry signals which would report on transducer and resolver responses, among other things. The cable 212 is routed to the air supply junction box 216 which is sealed at its entry point therewith by a seal 218 to prevent air leakage from junction box 216. An air supply hose 220 is also routed to the air supply junction box 216 and carries air at a pressure significantly higher than atmospheric thereto. The air supply hose 220 is sealingly connected by clamp 224 about an air receiving nozzle 222 extending from the junction box 216. The cable 212 can either be through-routed through the junction box 216 or terminated at a connector 225 provided for that purpose. In either event, the cable 212 is routed from junction box 216 into the larger cable or hose 214. Hose 214, with cable 212 disposed therein, includes a generally annular space 215 along its length to carry the pressurized air where needed. Cable 214 is clamped over nozzle 226 by clamp 228 to provide an air-tight fit between the air supply junction box 216 and the dual hose 214. From junction box 216, the hose 214 is routed to the underwater junction box 208. It is secured thereto by water-tight seals 230 and 232 at the points where it enters junction box 208. From the junction box 208, the hose 214 can be branched by internal connectors (not shown) to any one or more of the motors, resolvers, transducers, etc., used in the inspection apparatus 14. Further, since cable 214 can depart the junction box 208 carrying pressurized air, various motor and resolver housings can also be pressurized where desired. As previously noted, the transducer array 28 is employed as the examination means by which the integrity of the vessel welds 13 or any appropriate portion of the vessel 10 can be inspected. A typical plan view of the transducer array 28 disposed on the mounting plate 40 is shown in FIG. 24. It should be noted with respect to the individual transducers themselves, that they are grouped or arrayed in a manner which permits the manipulator arm 26 to optimally position the plate 40 so that the greatest inspection flexibility results. For example, the three transducers 240, 242, and 244 can be positioned, as illustrated in FIG. 25, to direct their ultrasonic beams to impinge at point 246 on the vessel 10. Transducer 242 can be oriented to impinge perpendicularly to the vessel wall at point 246 to verify the water path distance or to check for vessel flaws. Transducers 240 and 244 can be used to direct angled beams at point 246 which may be a weld point or material adjacent thereto. Further, transducers 240 and 244 may be coupled to pitch-catch or merely echo their respective beams. The individual transducers are secured to plate 40 by a transducer mounting assembly, generally designated 250, shown in FIGS. 26 and 27 in its normal orientation. The transducer mounting assembly includes a hollow, generally rectangularly shaped bar 252 having a slot 254 cut longitudinally therein. The bar 252 is bolted to the transducer plate 40 by bolts 256 one of which is shown in FIG. 27. A circular bar 258 is captured at either end thereof by holders 260 and fastened securely therein by set screws 262. The holders 260 are secured to the transducer plate 40 by bolts 264, also shown in FIG. 27, parallel to and spaced apart from bar 252. A transducer 244 is held in a retaining block 266 having a circular bore 268 therein sized to accommodate the transducer 244. The top portion of bore 268 is countersunk or cut away to accept and support the flange 245 of transducer 244 in the circular shelf 270. Plates 272, which are fitted over and about the transducer flange 245 and secured to the top of retaining block 266 by bolts 274, tightly capture and retain the transducer 244 in the block 266. If necessary, the transducer 244 can be rotated in the retaining block by loosening the bolts 274. The retaining block 266 includes upstanding flanges 276 and 278 having circular bores 280 and 282 cut therethrough for respectively accepting a hinge pin 284 therein. The retaining block 266 is, in turn, secured to a yoke 286 which also includes two upstanding flanges 288 and 290, each having a circular bore 292 and 294 cut respectively therein. The hinge pin 284 extends through the bores 280 and 292 to pivotally fasten one side each of the block 266 and the yoke 286 to each other. A set screw 296, extending from the top of flange 276 through a bore 298 therein is used to clamp the hinge pin 284 to the retaining block 266. The other end of hinge pin 284 remains free to rotate in bore 292 of flange 288. The other side of retaining block 266 is also pivotally secured to the yoke 286 by a hinge pin 300, which is "T" shaped in cross-section. The leg of hinge pin 300 extends through the bores 282 and 294 of flanges 278 and 290. It is secured within bore 282 and clamped to flange 278 by a set screw 302. The head portion of hinge pin 300 abuts the flange 290 and is captured by a "U" shaped clamp 304 which is bolted to flange 290. The leg portions 306 and 308 of clamp 304 are held together by a bolt 310 which is threaded through bores 312 and 314 cut respectively in leg portions 306 and 308. When the bolt 310 is tightened down, leg portions 306 and 308 are drawn tightly together about the head portion of hinge pin 300 preventing it from turning in clamp 304. When bolt 310 is loosened, however, the transducer 244 and the retaining block 266 can be pivoted about the hinge pins 284 and 300. A side view of a pivoted restraining block 266, with transducer 244 having been tilted forwardly, is shown in FIG. 31. An exploded isometric view of the transducer 244, restraining block 266 and yoke 286 coupling is illustrated in FIG. 33. As shown in FIGS. 26 and 31, two circular sleeves 320 and 322 are fit over and slid along the circular bar 258 prior to its being clamped into the holders 260. The sleeves 320 and 322 are bolted to one side of the yoke 286 by bolts 321. An angle bracket 324 is secured to the other side of yoke 286 by bolts 326. The perpendicular portion of bracket 324 is bolted to the rectangular bar 252 by the end bolts 328. If bolts 328 are loosened, the yoke 286 and therefore the transducer 244 held therein can be moved transversely along the bars 252 and 258. The bolts 328 pass through a bore 325 in the perpendicular portion of bracket 324 as illustrated in FIGS. 27, 30 and, most clearly, 32. After passing through the bore 325, the bolts extend through plate 340 and the slot 254 into the bar 252. The legs 330 of bolts 328 are threaded through the barrel nuts or pivots 342 and are pierced by cotter pins 344 at their terminal point to prevent their being worked out of the barrel nuts 342. A centered bolt 346 is threaded through the perpendicular portion of bracket 324 and abuts the plate 340 which acts as a stop therefor. When a locknut 348 is loosened, the bolt 346 can be tightened down, increasing the distance between plate 340 and the bracket 324, thereby pivoting the yoke 286 about the circular bar 258. An example of a pivoted yoke 286 is shown in FIG. 30. When the bolt 346 is tightened, the barrel nuts 342 pivot in the slot 254 permitting the yoke 286 to move to its canted position. It should be noted that the end bolts 328 are not loosened to effect or aid in this pivoting motion of the yoke 286. A number of bolt head flanges 350 are used to cover and retain various bolts should they loosen and work out of engagement. As the transducer array 28 is disposed about the vessel 10, particularly in or near one of the nozzles 38, it becomes difficult because of the curved vessel surfaces, to maintain one of the transducers perpendicular to the vessel wall and simultaneously insure proper clearances. For that reason, at least two transducers 370 and 372 are mounted on upstanding brackets 374 and 376 rather than on the bars 252 and 258. An example of this mounting arrangement is depicted in FIGS. 28 and 29. The restraining block 266 is removed from the yoke 286 and is bolted to the brackets 374 and 376. It is then pivoted at an appropriate angle by loosening bolt 310 of the "U" clamp 304 as previously described. In this case, however, clamp 304 is bolted to the block 266 rather than the yoke 286. As shown in FIG. 29, the transducer beam 380 can be directed against the curved vessel wall 382, generally normal thereto, and the same transducer can be employed to receive the echo. Thus, the perpendicular distance between the transducer plate 40 and the vessel wall 382 can be continuously monitored. Utilizing such information, the manipulator arm 26 can be moved accordingly to prevent collisions. Thus, there has been described a versatile transducer mounting assembly which tightly retains a transducer therein, but which can be adjusted to permit translational, pivotal and rotary motion of the transducer relative to the mounting plate. While the invention has been shown and described herein in considerable detail, such disclosure is to be considered as only illustrative or exemplary in character and not restrictive, as within the broad scope of the invention, modifications of or alternatives thereto may readily suggest themselves to persons skilled in this art. |
description | The present application is related to and claims the benefit of the earliest available effective filing date(s) from the following listed application(s) (the “Related Applications”) (e.g., claims earliest available priority dates for other than provisional patent applications or claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Related Application(s)). For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of United States Patent Application entitled METHOD, SYSTEM, AND APPARATUS FOR THE THERMAL STORAGE OF ENERGY GENERATED BY MULTIPLE NUCLEAR REACTOR SYSTEMS, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, CLARENCE T. TEGREENE, JOSHUA C. WALTER, LOWELL L. WOOD, JR., AND VICTORIA Y. H. WOOD as inventors, filed Jul. 30, 2010, application Ser. No. 12/804,894, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of United States Patent Application entitled METHOD, SYSTEM, AND APPARATUS FOR THE THERMAL STORAGE OF NUCLEAR REACTOR GENERATED ENERGY, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, CLARENCE T. TEGREENE, JOSHUA C. WALTER, LOWELL L. WOOD, JR., AND VICTORIA Y. H. WOOD as inventors, filed Feb. 18, 2010, application Ser. No. 12/660,025, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of United States Patent Application entitled METHOD, SYSTEM, AND APPARATUS FOR THE THERMAL STORAGE OF NUCLEAR REACTOR GENERATED ENERGY, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, CLARENCE T. TEGREENE, JOSHUA C. WALTER, LOWELL L. WOOD, JR., AND VICTORIA Y. H. WOOD as inventors, filed Feb. 19, 2010, application Ser. No. 12/660,157, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. The United States Patent Office (USPTO) has published a notice to the effect that the USPTO's computer programs require that patent applicants reference both a serial number and indicate whether an application is a continuation or continuation-in-part. Stephen G. Kunin, Benefit of Prior-Filed Application, USPTO Official Gazette Mar. 18, 2003, available at http://www.uspto.gov/web/offices/com/sol/og/2003/week11/patbene.htm. The present Applicant Entity (hereinafter “Applicant”) has provided above a specific reference to the application(s) from which priority is being claimed as recited by statute. Applicant understands that the statute is unambiguous in its specific reference language and does not require either a serial number or any characterization, such as “continuation” or “continuation-in-part,” for claiming priority to U.S. patent applications. Notwithstanding the foregoing, Applicant understands that the USPTO's computer programs have certain data entry requirements, and hence Applicant is designating the present application as a continuation-in-part of its parent applications as set forth above, but expressly points out that such designations are not to be construed in any way as any type of commentary and/or admission as to whether or not the present application contains any new matter in addition to the matter of its parent application(s). The present disclosure generally relates to the thermal storage and subsequent utilization of nuclear reactor generated energy. In one aspect, a method includes but is not limited to diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, diverting at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir, and supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. In addition to the foregoing, other method aspects are described in the claims, drawings, and text forming a part of the present disclosure. In one or more various aspects, related systems include but are not limited to circuitry and/or programming for effecting the herein-referenced method aspects; the circuitry and/or programming can be virtually any combination of hardware, software, and/or firmware configured to effect the herein-referenced method aspects depending upon the design choices of the system designer. In one aspect, a system includes but is not limited to means for diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, means for diverting at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir, and means for supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. In addition to the foregoing, other system aspects are described in the claims, drawings, and text forming a part of the present disclosure. In one aspect, an apparatus includes but is not limited to a first energy transfer system configured to divert a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one additional energy transfer system configured to divert at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir, and a heat supply system configured to supply at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. In addition to the foregoing, other system aspects are described in the claims, drawings, and text forming a part of the present disclosure. In addition to the foregoing, various other method and/or system and/or program product aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is NOT intended to be in any way limiting. Other aspects, features, and advantages of the devices and/or processes and/or other subject matter described herein will become apparent in the teachings set forth herein. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. Referring now to FIG. 1, a system 100 for storing and subsequently utilizing energy generated by a plurality of nuclear reactor systems 102 is described in accordance with the present disclosure. A first energy transfer system 104 may divert energy (e.g., thermal energy or electrical energy) from a portion (e.g., first nuclear reactor 108 or first energy conversion system 110) of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to one or more heat storage materials 111 of one or more auxiliary thermal reservoirs 112, and a second energy transfer system 104 may divert energy from a portion (e.g., second nuclear reactor 108 or second energy conversion system 108) of a second nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more heat storage materials 111 of the one or more auxiliary thermal reservoirs 112. Further, an additional energy transfer system, up to and including an Nth energy transfer system 104, may divert energy from a portion (e.g., Nth nuclear reactor 108 or Nth energy conversion system 110) of an Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more heat storage materials 111 of the one or more auxiliary thermal reservoirs 112. Then, one or more heat supply systems 114 (e.g., first heat supply system 114, second heat supply system, or Nth heat supply system 114) may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the energy conversion system 110 may include, but is not limited to, a first energy conversion system 110 associated with the first nuclear reactor system 106, a second energy conversion system 110 associated with the second nuclear reactor system 106, or an Nth energy conversion system 110 associated with the Nth nuclear reactor system 106. It is further contemplated that the labeling of the various nuclear reactor systems 106 as the first nuclear reactor system 106, the second nuclear reactor system 106, the third nuclear reactor system 106, and the Nth nuclear reactor system 106 is for illustrative purposes only. As such, the first nuclear reactor system 106, the second nuclear reactor system 106, the third nuclear reactor system 106 and the Nth nuclear reactor system 106 are substantially interchangeable for the purposes described within the present disclosure. Similarly, it is contemplated that the labeling of the various energy conversion systems 110 as the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system 110 is for illustrative purposes only and, therefore, the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system 110 are substantially interchangeable for the purposes described in the present disclosure. Additionally, it is contemplated that the labeling of the various heat supply systems 114 as the first heat supply system 114, the second heat supply system 114, and the Nth heat supply system 114 is for illustrative purposes only and, therefore, the first heat supply system 114, the second heat supply system 114, and the Nth heat supply system 114 are substantially interchangeable for the purposes described in the present disclosure. It is further contemplated that the labeling of the various energy transfer systems 104 as the first energy transfer system 104, the second energy transfer system 104, and the Nth energy transfer system 104 is for illustrative purposes and therefore the first energy transfer system 104, the second energy transfer system 104, and the Nth energy transfer system 104 are substantially interchangeable for the purposes described in the present disclosure. Referring now to FIG. 2, one or more of the nuclear reactors 108 (i.e., the first nuclear reactor, the second nuclear reactor, or the Nth nuclear reactor) of one or more of the nuclear reactor systems 106 (i.e., first nuclear reactor system, second nuclear reactor system, or Nth nuclear reactor system) of the plurality of nuclear reactor systems 102 may include, but are not limited to, one or more thermal spectrum nuclear reactors 202, one or more fast spectrum nuclear reactors 204, one or more multi-spectrum nuclear reactors 206, one or more breeder nuclear reactors 208, or one or more traveling wave nuclear reactors 210. For example, the energy produced by a thermal spectrum nuclear reactor 202 of a nuclear reactor system 106 may be diverted from the thermal spectrum nuclear reactor 202 to one or more auxiliary thermal reservoirs 112 using an energy transfer system 104. Then, one or more heat supply systems 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 (e.g., the first energy conversion system, the second energy conversion system, or the Nth energy conversion system) of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of further example, the energy produced by a traveling wave nuclear reactor 210 of a nuclear reactor system 106 may be diverted from the traveling wave nuclear reactor 210 to one or more auxiliary thermal reservoirs 112 using an energy transfer system 104. Then, one or more heat supply systems 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the nuclear reactor systems 106. Further, it will be recognized by those skilled in the art that the first nuclear reactor 108, the second nuclear reactor 108, and the Nth nuclear reactor 108 need not consist of the same type of nuclear reactor. For instance, the first nuclear reactor 108 may include a traveling wave nuclear reactor 210, the second nuclear reactor 108 may include a breeder nuclear reactor 208, and the Nth nuclear reactor 108 may include a thermal spectrum nuclear reactor 202. In another aspect, one or more of the energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactors 102 may include, but are not limited to, one or more primary energy conversion systems 212, one or more auxiliary energy conversion systems 214, or one or more emergency energy conversion systems 216. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more primary energy conversion systems 212 of the one or more nuclear reactor systems 106 (e.g., the first nuclear reactor system, the second nuclear reactor system or the Nth nuclear reactor system) of the plurality of nuclear reactor systems 102 For instance, the primary energy conversion system 212 may include a turbine 218 coupled to an electric generator used to supply electrical power to the primary load 220 (e.g., electrical power grid) of one or more nuclear reactor systems 106. By way of an additional example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more auxiliary energy conversion systems 214 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the auxiliary energy conversion system 214 may include an energy conversion system that supplements or replaces the output of the primary energy conversion system 212. For example, the auxiliary energy conversion system 214 may include a turbine 218 coupled to an electric generator used to provide supplemental or backup electric power to the primary load 220 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of a further example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more emergency energy conversion systems 216 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the emergency energy conversion system may include a turbine 218 coupled to an electric generator used to supply electric power to an operation system 222 (e.g., monitoring system, safety system, control system, coolant system or security system) of one or more nuclear reactor systems 106 (e.g., first nuclear reactor, second nuclear reactor, or Nth nuclear reactor) of the plurality of nuclear reactor systems 102. It will be appreciated by those skilled in the art that the emergency energy conversion system 216 may be configured to operate at temperatures lower than the operational temperature of the primary energy conversion system 212, allowing the emergency energy conversion system 216 to supply electrical energy to various operation systems 222 of one or more nuclear reactors 106 of the plurality of nuclear reactors 102 during emergency situations when grid power is unavailable. Further, it will be recognized by those skilled in the art that the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system need not consist of the same type of energy conversion system. For instance, the first energy conversion system 110 may include a primary energy conversion system 212, the second energy conversion system 110 may include an auxiliary energy conversion system 214, and the Nth energy conversion system 110 may include an emergency energy conversion system 216. In another aspect, one or more of the energy conversion systems 110 may include, but are not limited to, one or more turbines 218 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more turbines 218 of one or more nuclear reactors 106 of the plurality of nuclear reactors 102. By way of further example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to a working fluid 224 of one or more turbines 218 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to a pressurized steam working fluid 224 of one or more turbines 218 of the one or more nuclear reactor systems 106. It will be appreciated by those skilled in the art that the thermal energy supplied from the auxiliary thermal reservoir 112, via the one or more heat supply systems 114, to the working fluid 224 of one or more turbines 218 of the one or more nuclear reactor systems 106 may be used to augment the thermal energy supplied to the working fluid 224 of the one or more turbines 218 from the one or more nuclear reactors 108 of the one or more nuclear reactor systems 106. In another aspect, one or more of the energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include, but are not limited to, one or more topping cycles 226, one or more bottoming cycles 228, or one or more low grade heat dumps 230. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more topping cycles 226 of one or more of the nuclear reactor systems 106. By way of another example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more bottoming cycles 228 of one or more of the nuclear reactor systems 106. By way of further example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more low grade heat dumps of one or more of the nuclear reactor systems 106. For instance, the low grade heat dump may include a portion of the surrounding environment (e.g., surrounding soil or atmosphere). It will be recognized by those skilled in the art that the low grade environmental heat dump serves as the ultimate heat sink, allowing for the effective removal of reactor core decay heat in the event the primary heat removal system(s) fail. In this context, the auxiliary thermal reservoir may serve as a thermal capacitor, residing upstream of the more thermally resistive low grade heat dump, such as the surrounding soil or surrounding atmosphere. As the reactor decay heat falls of exponentially, the auxiliary thermal reservoir, acting as a thermal capacitor, may act to absorb the high initial heat load, while the heat is dissipated at a lower rate to the low grade environmental heat dump. Further, it will be recognized by those skilled in the art that the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system 110 need not consist of the same type of energy conversion system. For instance, the first energy conversion system 110 may include a topping cycle 226 of the first nuclear reactor system 106, the second energy conversion system 110 may include a bottoming cycle 228 of the second nuclear reactor system 106, and the Nth energy conversion system 110 may include a low grade heat dump 230 of the Nth nuclear reactor system 106. In another aspect, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary reservoir 112 to one or more boiling loops 232 of the one or more nuclear reactor systems 106, wherein the one or more boiling loops 232 of the one or more nuclear reactor systems 106 are in thermal communication with one or more energy conversion systems 110 of the one or more nuclear reactor systems 106. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of an auxiliary reservoir 112 to a boiling loop 232 in thermal communication with a turbine 218 of one or more nuclear reactor systems 106. By way of further example, the boiling loop 232 may be in thermal communication with one or more topping cycles 226, one or more bottoming cycle 228 or one or more low grade heat dumps 230 of the one or more nuclear reactor systems 106. It will be appreciated by those skilled in the art that the thermal energy supplied to the boiling loop 232 of the one or more nuclear reactor systems 106 from the one or more auxiliary thermal reservoirs 112 may be used to augment the thermal energy supplied to the one or more boiling loops 232 from the one or more nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 3, one or more of the nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include a nuclear reactor having a liquid coolant 302. For example, the liquid coolant 302 of one or more of the nuclear reactors 108 may include, but is not limited to, a liquid metal salt coolant 304 (e.g., lithium fluoride, beryllium fluoride or other fluoride salts), a liquid metal coolant 306 (e.g., sodium, lead, or lead bismuth), a liquid organic coolant 308 (e.g., diphenyl with diphenyl oxide), or a liquid water coolant 310. For instance, an energy transfer system 104 may divert energy from a portion of a liquid sodium cooled nuclear reactor of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. In another instance, the energy transfer system 104 may divert energy from a portion of a liquid water cooled nuclear reactor 220 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. In an additional instance, the energy transfer system 104 may divert energy from a portion of a lithium fluoride cooled nuclear reactor of a nuclear reactor system 106 of the plurality of the nuclear reactor systems to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include one or more nuclear reactors having a pressurized gas coolant 312. For example, the pressurized gas coolant 222 may include, but is not limited to, pressurized helium gas or pressurized carbon dioxide gas. For instance, the energy transfer system 104 may divert energy from a portion of a pressurized helium cooled nuclear reactor 312 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include one or more nuclear reactors having a mixed phase coolant 314. For example, the mixed phase coolant 314 may include, but is not limited to, a gas-liquid mixed phase material (e.g., steam water-liquid water). For instance, the energy transfer system 104 may divert energy from a portion of a steam water-liquid water cooled nuclear reactor 314 of a nuclear reactor system 106 of the plurality of nuclear reactors 102 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 4A, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a liquid heat storage material 402 of one or more auxiliary thermal reservoirs 112. For example, the liquid heat storage material 402 may include, but is not limited to, an organic liquid 404 (e.g., diphenyl with diphenyl oxide), a liquid metal salt 406 (e.g., lithium fluoride, beryllium fluoride or other fluoride salts), a liquid metal 408 (e.g., sodium, lead, or lead bismuth), or liquid water 410. For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of liquid sodium of an auxiliary thermal reservoir 112. In another instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems to a mass of liquid water 410 of an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the liquid heat storage material 402 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a pressurized gas heat storage material 412 of one or more auxiliary thermal reservoirs 112. For example, the pressurized gas material 412 suitable for heat storage may include, but is not limited to, pressurized helium gas or pressurized carbon dioxide gas. For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of pressurized helium of an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the pressurized gas material 412 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a solid heat storage material 414 of one or more auxiliary thermal reservoirs 112. In one aspect, the solid heat storage material 414 may include a continuous solid material forming a solid object 416. For example, the solid object 416 suitable for heat storage may include, but is not limited to, a three dimensional monolithic object (e.g., a brick), a three dimensional porous object (e.g., a brick containing pores suitable for fluid flow), a three dimensional channeled object (e.g. a brick containing channels suitable for fluid flow), or a three dimensional engineered object (e.g., an object containing an engineered honeycomb pattern for increased heat transfer). For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to one or more solid monolithic objects, such as a brick, a rod, or a sheet of material. In another instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a solid engineered object, such as an object constructed of a high heat capacity honeycomb structured material. Further, the solid object 416 may include, but is not limited to, a ceramic solid object, such as a carbide ceramic (e.g., titanium carbide or silicon carbide) or a boride ceramic, a metal solid (e.g., iron or steel) object, or an environmentally present solid (e.g., rock or stone) object. For example, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a ceramic solid object. By way of further example, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to an environmentally preexisting rock or stone structure located in close proximity to one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the solid heat storage material 414 may include a particulate solid material 418. For example, the particulate solid material 418 may include, but is not limited to, a granular material (e.g. sand) or a powder material. For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of sand located in close proximity to one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of sand via heat pipes, wherein one portion of the heat pipes is in thermal communication with a portion of one or more nuclear reactors 108 of one or more nuclear reactor systems 106 and a second portion of the heat pipes is embedded in a volume of sand located in close proximity to one or more nuclear reactor systems 106. It will be recognized by those skilled in the art that the volume of the sand, and like solid materials, need not be constrained by the volume of a reservoir containment system 122, in that uncontained sand, stone, and like heat trapping materials surrounding one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may serve as a suitable heat storage material 111. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the solid heat storage material 414 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mixed phase heat storage material 420 of one or more auxiliary thermal reservoir 112. For example, the mixed phase material 420 suitable for heat storage may include, but is not limited to a gas-liquid mixed phase material (e.g., steam water-liquid water) or a liquid-solid mixed phase material (e.g. liquid sodium-solid sodium). For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of steam water-liquid water. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the mixed phase heat storage material 420 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of a heat storage material having a phase transition within the operating temperature 422 of the auxiliary thermal reservoir 112. For example, an auxiliary thermal reservoir 112 having a heat storage material 116 with a phase transition at approximately 100° C. may continuously operate at temperatures above and below the phase transition at 100° C. Those skilled in the art will recognize that this allows the heat supply system 114 to supply thermal energy from the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 at reservoir temperatures above, below and at the phase transition temperature of the heat storage material 111. For instance, given that sodium has an approximate melting temperature of 97.7° C., a sodium based auxiliary thermal reservoir 112 may operate in the liquid phase at temperatures above 97.7° C. and in the solid phase at temperatures below 97.7° C. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 having a phase transition within the operating temperature 422 of the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of one or more the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 4B, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of a heat storage material 111 contained in a reservoir containment system 424. For example, the reservoir containment system 424 may include, but is not limited to, an external vessel 426 or an external pool 432. By way of further example, the external vessel 426 may include, but is not limited to an external liquid vessel 428 or an external high pressure gas vessel 430. For instance, the one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of liquid metal 408 (e.g. liquid sodium) contained in an external liquid vessel 428. In another instance, the one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of pressurized gas 412 (e.g. pressurized helium) contained in an external high pressure vessel 430. By way of further example, the external pool 432 may include, but is not limited to, a liquid pool 434. For instance, the one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of liquid metal 408 (e.g. liquid sodium) contained in an external liquid pool 434. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 contained in the reservoir containment system 424 to one or more energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 4C, the one or more auxiliary thermal reservoirs 112 may store the energy diverted from the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 in the form of a temperature change 436 in the heat storage material 111 of the one or more auxiliary thermal reservoirs 112. For example, the energy diverted from the one or more nuclear reactor systems 106 to the heat storage material 111 of an auxiliary thermal reservoir 112 may cause the temperature of the heat storage material 111 to increase. For instance, the energy diverted from the one or more nuclear reactor systems 106 to the heat storage material 111 of an auxiliary thermal reservoir 112 may cause the temperature of the heat storage material 111, such as a liquid metal 408 (e.g., liquid sodium), to increase from an initial temperature of approximately 100° C. to a temperature of approximately 500° C. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 as a temperature increase 436 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the one or more auxiliary thermal reservoirs 112 may store the energy diverted from the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 in the form of a phase change 438 in the heat storage material 111 of the one or more auxiliary thermal reservoirs 112. For example, the phase change 438 in the heat storage material 111 may include a solid-liquid phase change 440 or a liquid-gas phase change 442. For instance, the energy diverted from the one or more nuclear reactor systems 106 to a solid heat storage material 414 of an auxiliary thermal reservoir 112 may be stored in the heat storage material 111 by melting the heat storage material 111. For example, the energy diverted from the one or more nuclear reactor systems 106 to a mass of solid sodium may liquefy the mass of sodium via a melting transition at approximately 97.7° C., thus storing a portion of the diverted energy in the liquid phase of the mass of sodium. It will be appreciated by those skilled in the art that the energy required to transform the heat storage material 111 from one phase (e.g. solid) to a new phase (e.g., liquid) is the heat of transformation (i.e., the “latent heat”). Then, a heat supply system 114 may supply a portion of the heat of transformation stored as thermal energy in the heat storage material 111 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactors 102. Referring now to FIG. 4D, the operational status of the auxiliary thermal reservoir 112 may be monitored using one or more reservoir monitoring systems 444. For example, the reservoir monitoring system 444 may include a temperature monitoring system 446, a pressure monitoring system 448, a system configured to determine the amount of energy stored in the thermal reservoir 450 or a system configured to determine the amount of available energy capacity of the thermal reservoir 452. For instance, a system configured to determine the amount of energy stored in the thermal reservoir 450 may include thermal and pressure monitoring devices configured to probe the temperature and pressure of the heat storage material 111 of the auxiliary thermal reservoir 112. Further, the thermal and pressure monitoring devices may be interfaced with a computer processing system configured to apply an established algorithm (e.g., established equation-of-state for the storage material in question) to the data outputs of the thermal and pressure monitoring devices, thus relating the temperature and pressure of the heat storage material 111 to the internal energy of the heat storage material 111 (e.g., liquid metal or pressurized gas). In another aspect, the temperature of the auxiliary thermal reservoir 112 may be controlled using a reservoir temperature control system 454. For example, the reservoir temperature control system 454 may be used to increase or decrease the temperature of the auxiliary thermal reservoir 112. For instance, in situations where the internal temperature of the auxiliary thermal reservoir reaches levels outside the predefined operational limits, the reservoir temperature control system 454 may signal the heat supply system 114 to transfer a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to the one or more energy conversion systems 110 of the nuclear reactor systems 106, such as a turbine 218 or a low grade heat dump 230. Referring now to FIG. 5A, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system configured to transfer thermal energy 502 from a portion of one or more nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112. For example, an energy transfer system configured to transfer thermal energy 502 from a portion (e.g., primary coolant system) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112 may divert thermal energy from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, one or more of the energy transfer systems configured to transfer thermal energy 502 from a portion of one or more nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112 may include, but are not limited to, one or more heat transfer systems 504. For example, a heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. For instance, the heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112 via thermal convection 506 (e.g., natural convection or forced convection via coolant pump(s)). In another instance, the heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112 via thermal conduction 508 (e.g., using a heat exchanger). Those having skill in the art will recognize that the one or more heat transfer systems 504 may be configured to transfer thermal energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112 using both thermal conduction 506 and thermal convection 508. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, the one or more heat transfer systems 504 may include, but are not limited to, one or more direct fluid exchange heat transfer systems 510. For example, a direct fluid exchange heat transfer system 510 may transfer thermal energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. The direct fluid exchange heat transfer system 510 may include a system configured to intermix the coolant of a nuclear reactor 108 of a nuclear reactor system 106 with the fluidic heat storage material 111 contained in the reservoir containment system 424. For instance, a fluid carrying loop may couple a primary coolant system of a nuclear reactor system 106 and the reservoir fluid containment system 424, allowing for the intermixing of the two fluids. The rate of reactor coolant-reservoir fluid intermixing may be controlled by the direct fluid exchange transfer system 510. For instance, a valve system and/or fluid pumps (e.g., mechanical pumps or magnetohydrodynamic pumps) may be employed to volumetrically limit the exchange of material between the reactor coolant system of a nuclear reactor system 106 and the reservoir fluid containment system 424. Moreover, the reservoir fluid and the reactor coolant may consist of identical or substantially similar materials. For example, both the reservoir fluid and the reactor coolant may consist of an identical liquid metal, such as liquid sodium. Additionally, the reservoir fluid and the reactor coolant may consist of different materials. For example, the reservoir fluid may consist of a liquid organic, such as diphenyl with diphenyl oxide, while the reactor coolant may consist of liquid sodium. Further, the one or more heat transfer systems 504 may include, but are not limited to, one or more reactor-reservoir heat exchangers 514. For example, a reactor-reservoir heat exchanger 514 may transfer thermal energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. For instance, the reactor-to-reservoir heat exchanger 514 may include a heat exchanger 515 having a first portion in thermal communication with the primary coolant system of the nuclear reactor system 106 and a second portion in thermal communication with the auxiliary thermal reservoir 112. Further, the heat transfer system 504 may include more than one reactor-reservoir heat exchanger 514. For example, a first portion of a first heat exchanger may be in thermal communication with the primary coolant system of the nuclear reactor system 106, while a second portion of the first heat exchanger may be in thermal communication with a heat exchange loop. Further, a first portion of a second heat exchanger may be in thermal communication with the auxiliary thermal reservoir 112, while a second portion of the second heat exchanger may be in thermal communication with the heat exchange loop. Collectively, the first heat exchanger-heat exchange loop-second heat exchanger system acts to transfer thermal energy from the primary coolant system of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system configured to transfer electrical energy 503 from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112. For example, an energy transfer system configured to transfer electrical energy 503 from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112 may transfer electrical energy from a portion (e.g., energy conversion system 110) of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion system 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, one or more of the energy transfer systems configured to transfer electrical energy 503 from a portion of one or more nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112 may include, but are not limited to, an electrical energy-to-thermal energy conversion system 516. For example, an electrical energy-to-thermal energy conversion system 516, such as a resistive heating device 517 (e.g., a heating coil 518), may convert a portion of the electrical energy produced by an energy conversion system 110 of a nuclear reactor system 106 to thermal energy. It will be recognized by those skilled in the art that the system for transferring electrical energy 503 from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112 may be utilized to convert excess electrical energy produced by an energy conversion system 110 of the nuclear reactor system 106 to thermal energy. Subsequently, a portion of that thermal energy may be transferred to and stored in the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 5B, one or more heat transfer systems 504 may transfer thermal energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112, wherein the portion of a nuclear reactor system 106 is in thermal communication with a heat source 522 of the nuclear reactor system 106. For example, a heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 of a nuclear reactor 108 of the nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, the portion of the nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 may include, but is not limited to, a portion of the primary coolant system 526 (e.g., portion of the primary coolant loop 528 or portion of the primary coolant pool 530). For example, a heat transfer system 504 may transfer thermal energy from a primary coolant system 526 of a nuclear reactor system 106 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one ore more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 5C, one or more heat transfer systems 504 may transfer thermal energy from a primary coolant system 526 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the primary coolant system 526 is in thermal communication (e.g., thermally communicating via a primary coolant system—secondary coolant system heat exchanger 536) with a secondary coolant system not in thermal communication 532 with the auxiliary thermal reservoir 112. For example, the auxiliary thermal reservoir 112 may be thermally coupled via a heat transfer system 504 to a primary coolant loop 528 of the primary coolant system 526. By way of further example, the auxiliary thermal reservoir 112 may be thermally coupled via a heat transfer system 504 to a primary coolant pool 530 of the primary coolant system 526. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems. Referring now to FIG. 5D, one or more heat transfer systems 504 may transfer thermal energy from a primary coolant system 526 of one or more nuclear reactor systems 106 to one or more auxiliary thermal reservoirs 112, wherein the primary coolant system 526 and a secondary coolant system 532 of the one or more nuclear reactor systems 106 are both in thermal communication with the auxiliary thermal reservoir 112. For example, the auxiliary thermal reservoir 112 may be thermally coupled to both a primary coolant loop 528 of the primary coolant system 526 of a nuclear reactor system 106 and a secondary coolant loop 534 of a secondary coolant system 532 of the nuclear reactor system 106, such that the thermal path coupling the primary coolant loop 526, the auxiliary thermal reservoir 112, and the secondary coolant loop 532 is parallel to the thermal path coupling the primary coolant loop 526, the primary-secondary coolant system heat exchanger 536, and the secondary coolant loop 532. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 6, the heat supply system 114 may include, but is not limited to, a heat exchange loop 602. For example, a first portion of a heat exchange loop 602 may be in thermal communication with a portion of the auxiliary thermal reservoir 112 and a second portion of the heat exchange loop 602 may be in thermal communication with an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, in response to a shutdown event of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102, the heat exchange loop 602 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the heat supply system 114 may include, but is not limited to, one or more heat pipes 604. For example, a first portion of a heat pipe 604 may be in thermal communication with a portion of the auxiliary thermal reservoir 112 and a second portion of the heat pipe 604 may be in thermal communication with an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, in response to a shutdown event of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102, the heat pipe 604 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the heat supply system 114 may include, but is not limited to, one or more heat exchangers 606. For example, a first portion of a first heat exchanger 608 may be in thermal communication with a portion of the auxiliary thermal reservoir 112 and a second portion of the first heat exchanger 606 may be in thermal communication with an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, the heat pipe 604 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that a combination of heat exchange loops 602, heat exchangers 606, and heat pipes 604 may be used in conjunction to supply heat from the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, a first heat exchanger 606, containing a number of heat pipes 604, may be used to thermally couple the auxiliary thermal reservoir 112 and a first portion of a heat exchange loop 602. Moreover, a second heat exchanger 606, also containing numerous heat pipes 604, may be used to thermally couple a portion of an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the heat exchange loop 602. Then, thermal energy may be supplied from the auxiliary thermal 112 reservoir to the energy conversion system 110 via the heat exchange loop-heat exchanger circuit. In another aspect, the heat supply system 114 may include, but is not limited to, one or more thermoelectric devices 608. For example, a first portion of a thermoelectric device 608 (e.g., p-type/n-type semiconductor thermoelectric junction) may be placed in thermal communication with the auxiliary thermal reservoir 112, while a second portion of the thermoelectric device 608 may be placed in thermal communication with a cold reservoir (e.g., an environmental reservoir or any portion of the nuclear reactor system at a temperature lower than the auxiliary thermal reservoir) of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, the electrical power produced by the thermoelectric conversion of the thermal energy stored in the auxiliary thermal reservoir 112 may be used to supplement or replace the electrical output of an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 7, an additional energy source 702 may supplement the auxiliary thermal reservoir 112 with an additional portion of energy. For example, excess energy from the load 220 (e.g., the external grid 703) of one or more of the nuclear reactor systems 106 may be used to provide supplemental energy to the auxiliary thermal reservoir 112. For instance, when grid requirements are such that an energy conversion system 110 is producing excess electrical power, the excess power may be converted to thermal energy via an electrical-to-thermal energy conversion process (e.g., heating coil) and transferred to the auxiliary thermal reservoir 112 using a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the energy transfer systems 104 during normal operation. By way of another example, the additional energy source 702 may include, but is not limited to, a non-nuclear reactor energy source 708, such as coal powered generator, a solar array, or wind powered turbine. For instance, electrical energy produced from a coal powered generator may be converted to thermal energy via an electrical-to-thermal energy conversion process and transferred to the auxiliary thermal reservoir 112 using a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the energy transfer systems 104 during normal operation. In another instance, excess electrical energy from a solar array or wind powered turbine may be converted to thermal energy via an electrical-to-thermal energy conversion process and transferred to the auxiliary thermal reservoir 112 using a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the energy transfer systems 104 during normal operation. In an additional instance, thermal energy produced by a coal generator may be transferred directly to the auxiliary thermal reservoir 112 via a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the primary energy transfer systems 104 during normal operation. It will be recognized by those skilled in the art that the supplemental energy supplied to the auxiliary thermal reservoir 112 by an additional energy source may be used to superheat the reservoir material of the auxiliary thermal reservoir to temperatures beyond normal operational capability. Referring now to FIG. 8A, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a condition 802. The conditions with which the energy transfer system is responsive may include, but are not limited to, nuclear reactor operational conditions (e.g., temperature, rate of change of temperature, pressure or rate of change of pressure, nuclear reactor capacity), power demand on the one or more nuclear reactor systems (e.g., electrical power requirements of the grid), nuclear reactor system operation system conditions (e.g., control system, monitoring system, or safety system (e.g., heat sink status or coolant pump status)). For example, in response to a coolant pump malfunction of one of the nuclear reactor systems 106, an energy transfer system 104 may divert energy from a portion of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. By way of further example, at or near a specified operating temperature of a portion of a nuclear reactor system 106 (e.g., nuclear reactor core or nuclear reactor coolant fluid), an energy transfer system 104 may initiate transfer of thermal energy from the nuclear reactor 108 of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a condition of a first nuclear reactor system 804. For example, in response to a coolant pump malfunction of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102, an energy transfer system configured to respond to a condition of the first nuclear reactor system 804 may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Additionally, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a condition of an additional nuclear reactor system 806 of the plurality of nuclear reactor systems 102. For example, in response to a drop in the energy output of a second nuclear reactor system 106 of the plurality of nuclear reactor systems, the energy transfer system configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. By way of further example, in response to a drop in the energy output of the second and third nuclear reactor systems 106 (e.g., a drop in both the individual outputs of the second nuclear reactor system and third nuclear reactor system or a drop in the collective output of the second and third nuclear reactors systems) of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. Further, in response to a drop in the energy output of the Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. More generally, in response to a condition of the Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102, the corresponding energy transfer systems configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106, the second nuclear system 106, or up the (N-1) nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to the determination of excess capacity 808 of one or more of the nuclear reactor systems of the plurality of the nuclear reactor systems 102. For example, in the event one or more of the nuclear reactor systems 106 is producing more energy than is required by the load (e.g., external electrical power grid) of the energy conversion system 110 of the nuclear reactor system 106, the energy transfer system may initiate transfer of thermal or electrical energy from a portion of one or more of the nuclear reactor systems 106 (e.g., a first nuclear reactor system 106, a second nuclear reactor system 106 or a Nth nuclear system 106) to the auxiliary thermal reservoir 112. For instance, in the event a first nuclear reactor system 106 is producing more energy than is required by the load (e.g., external electrical power grid) of the energy conversion system 110 of the first nuclear reactor system 106, the energy transfer system 104 may initiate transfer of thermal or electrical energy from a portion of the first nuclear reactor system 106, the second nuclear reactor system 106 or the Nth nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. In an additional aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to an operation system 810 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. For example, the energy transfer system responsive to an operation system 810 may include, but is not limited to, an energy transfer system responsive to a signal from an operation system 812. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from an operation system (e.g., shutdown system, warning system, or security system) of one or more of the nuclear reactor systems 106, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the energy transfer system responsive to an operation system 810 may include, but is not limited to, an energy transfer system responsive to a monitoring system 808 (e.g., temperature monitoring system or pressure monitoring system), an energy transfer system responsive to a control system 810, or an energy transfer system responsive to safety system 812. For instance, in response to a signal from a monitoring system 814 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102, one or more of the energy transfer systems 104 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. In another instance, in response to a signal from a control system 816 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102, one or more of the energy transfer systems 104 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. Further, in response to a signal from a safety system 818 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102, one or more of the energy transfer systems 104 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the one energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. Further, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a signal from an operation system of the first nuclear reactor system. For example, in response to a signal from an operation system of the first nuclear reactor system 106 of the plurality of nuclear reactor systems 102, an energy transfer system configured to respond to a signal of an operation system of the first nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. For instance, in response to a signal from the monitoring system of the first nuclear reactor system the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Moreover, an energy transfer system associated with a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a signal from an operation system of an additional nuclear reactor system. For example, in response to a signal from an operation system of an additional nuclear reactor system 106 (e.g., second nuclear reactor system 106, third nuclear reactor system 106, or Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a signal from an operation system of an additional nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. For instance, in response to a signal from a monitoring system of an additional nuclear reactor system, the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a signal from an operator 820 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. For example, in response to a signal from an operator (e.g., human user or human controlled system, such as a programmed computer system), one or more energy transfer systems responsive to a signal from an operator 820 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. For instance, an energy transfer system responsive to a signal from an operator 820, in response to a remote signal, such as a wireline or wireless signal from a computer terminal controlled by an operator, may initiate transfer of thermal energy from a nuclear reactor 108 of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. In an additional aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a pre-selected diversion start time 822. For example, the pre-selected diversion start time may include a time of elapse (e.g., time of elapse measured relative to a specific event, such as a shutdown event or satisfaction of grid demand requirements) or an absolute time. For instance, an energy transfer system responsive to a pre-selected diversion start time 822, at a pre-selected absolute time (e.g., 2:00 a.m. eastern standard time) may initiate transfer of energy from a nuclear reactor system 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. It will be recognized by those skilled in the art that historical grid power demand data may be utilized to determine the appropriate time in which to begin diversion of nuclear reactor generated energy to the auxiliary thermal reservoir 112. In another instance, the energy transfer system responsive to a pre-selected diversion start time 822, upon elapse of a pre-selected amount of time from a specific event, such as a nuclear reactor shutdown or achievement of power production in excess of external demand, may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a reservoir operation system 824 of one or more auxiliary thermal reservoirs 112. For example, an energy transfer system responsive to a reservoir operation system 824 may include, but is not limited to, an energy transfer system responsive to a signal from a reservoir operation system 826. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from a reservoir operation system of the auxiliary thermal reservoir 112, the energy transfer system responsive to a signal from a reservoir operation system 826 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the energy transfer system responsive to a reservoir operation system 824 may include, but is not limited to, an energy transfer system responsive to a reservoir monitoring system 828 (e.g., temperature monitoring system, pressure monitoring system, system for monitoring amount of stored energy, or system for monitoring the amount of available storage capacity), an energy transfer system responsive to a reservoir control system 830, or an energy transfer system responsive to a reservoir safety system 832. For instance, in response to a signal from a reservoir monitoring system, the energy transfer system responsive to a reservoir monitoring system 828 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. In another instance, in response to a signal from a reservoir control system, the energy transfer system responsive to a reservoir control system 830 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, in response to a signal from a reservoir safety system, the energy transfer system responsive to a reservoir safety system 8832 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 8B, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a shutdown event 834 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. For example, an energy transfer system responsive to a shutdown event 834 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 may include, but is not limited to, an energy transfer system responsive to a scheduled shutdown event 834 of one or more of the nuclear reactor systems 106 or an energy transfer system responsive to an emergency shutdown event 838 of one or more of the nuclear reactor systems 106. For instance, in response to a schedule shutdown event (e.g., routine maintenance), one or more of the energy transfer systems responsive to a scheduled shutdown event 836 of one or more of the nuclear reactors 106 may initiate transfer of energy from a portion of one or more of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. In another instance, in response to an emergency shutdown event (e.g., SCRAM), one or more of the energy transfer systems responsive to an emergency shutdown event 838 of one or more of the nuclear reactors 106 may initiate transfer of energy from a portion of one or more of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that, in response to a shutdown event of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102, energy may be diverted from a portion of the nuclear reactor system 106 to the auxiliary thermal reservoir 112 prior to, during, and following the shutdown of the nuclear reactor 108 of the nuclear reactor system 106, as part of the steps required to facilitate the nuclear reactor system 106 shutdown. Further, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a shutdown event of the first nuclear reactor system. For example, in response to a shutdown event of the first nuclear reactor system 106 of the plurality of nuclear reactor systems 102, an energy transfer system configured to respond to a shutdown event of the first nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. For instance, in response to an emergency shutdown event of the first nuclear reactor system the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Moreover, an energy transfer system associated with a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a shutdown event of an additional nuclear reactor system. For example, in response to a shutdown event of an additional nuclear reactor system 106 (e.g., second nuclear reactor system 106, third nuclear reactor system 106, or Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a shutdown event of an additional nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. For instance, in response to a scheduled shutdown event of an additional nuclear reactor system, the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system suitable for establishing thermal communication between a nuclear reactor system and the auxiliary thermal reservoir 840. For example, in response to a condition, the energy transfer system suitable for establishing thermal communication between the nuclear reactor system and the auxiliary thermal reservoir 840 may establish a thermal pathway between a portion of a nuclear reactor 108 (e.g., primary coolant system) of the nuclear reactor system 106 and the auxiliary thermal reservoir 112. For instance, in the case of a direct fluid exchange heat transfer system 510, a control valve may be used to initiate the intermixing of the reactor coolant and reservoir fluid. In another instance, in the case of a heat transfer system employing a reactor-reservoir heat exchanger 514, a control valve may be used to initiate reactor coolant flow through the heat exchanger. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 842. For example, in response to the determination of energy currently stored in the auxiliary thermal reservoir 112, the energy transfer system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 842 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the energy transfer system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 842 may include an energy transfer system responsive to the determination of the percentage of energy stored, relative to the overall storage capacity, in the auxiliary thermal reservoir 844. For example, in response to a determination of a set percentage level of stored energy (e.g., 25% of energy storage capacity is being utilized), the energy transfer system responsive to the determination of the percentage of stored energy 842 may initiate transfer of energy from a portion a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In an additional aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 846. For example, in response to the determination of available energy storage capacity, the energy transfer system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 846 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the energy transfer system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 846 may include an energy transfer system responsive to the determination of the percentage of available energy storage capacity in the auxiliary thermal reservoir 848. For example, in response to a determination of a set level of available energy storage (e.g., 75% storage capacity remains), the energy transfer system responsive to the determination of the percentage of available energy storage capacity 848 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 8C, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system suitable for diverting excess energy from a nuclear reactor system of the plurality of nuclear reactor systems to an auxiliary thermal reservoir 850. For example, an energy transfer system suitable for diverting excess energy from a nuclear reactor system to an auxiliary thermal reservoir 850 may transfer energy exceeding operational demand of an energy conversion system 852. For instance, in the event a turbine-generator system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 is producing electrical power in excess of grid demand, the energy transfer system 104 may transfer energy (e.g., thermal or electrical) from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, one or more of the energy transfer systems 104 may include an energy transfer system configured to divert a specified percentage of the energy output of a nuclear reactor system to an auxiliary thermal reservoir 854. For example, a control system or operator may choose to transfer a pre-selected percentage of a nuclear reactor system 106 output and transfer at least a portion of that energy to the auxiliary thermal reservoir 112. It will be recognized by those skilled in the art that the level of energy output pre-selected to be transferred to the auxiliary thermal reservoir may be a function of time and may be derived from historic external power demand curves. For example, in times of day or times of year historically displaying relatively low grid demand, the control system or operator may choose to divert a larger percentage of the output of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir than the percentage transferred during periods of higher demand. Referring now to FIG. 9A, one or more of the heat supply systems 114 may include, but are not limited to a heat supply system responsive to a condition 902. The conditions with which one or more of the heat supply systems are responsive may include, but are not limited to, nuclear reactor operational conditions (e.g., temperature, rate of change of temperature, pressure or rate of change of pressure, nuclear reactor capacity), power demand on the one or more nuclear reactor systems (e.g., electrical power requirements of the grid), nuclear reactor system operation system conditions (e.g., control system, monitoring system, or safety system (e.g., heat sink status or coolant pump status)), or reservoir operational conditions (e.g., temperature, rate of change of temperature, pressure or rate of change of pressure). For example, in response to a condition of one or more of the nuclear reactor systems 106, a heat supply system configured to respond to a condition of one or more of the nuclear reactor systems 904 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, in response to heightened power demand on a the nuclear reactor systems 106, a heat supply system responsive to heightened power demand on one or more of the nuclear reactor systems 906 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to a heat supply system responsive to a shutdown event 908. For example, in response to an emergency shutdown event (e.g., SCRAM), a heat supply system responsive to an emergency shutdown event 910 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of another example, in response to a scheduled shutdown event (e.g., routine maintenance), a heat supply system responsive to a schedule shutdown event 912 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that, in response to a shutdown event of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102, the thermal energy stored in the auxiliary thermal reservoir 112 may be transferred from the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more nuclear reactor systems 106 prior to, during, and following the shutdown of a nuclear reactor system 106 as part of the steps required to facilitate the nuclear reactor system 106 shutdown. In another aspect, one or more of the heat supply systems responsive to a shutdown event 908, may include, but are not limited to, a heat supply system responsive to a shutdown event established by an operation system 914. For example, in response to a shutdown event established by an operation system (e.g., shutdown system) of one or more of the nuclear reactor systems 106, a heat supply system responsive to a shutdown event established by an operation system 914 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of further example, a heat supply system responsive to a shutdown event established by a reactor control system 916 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems of the plurality of nuclear reactor systems. Further, the reactor control system may include a reactor control system responsive to a signal from one or more reactor safety systems 918. For example, a heat supply system responsive to a shutdown event established by a reactor control system responsive to a signal from a safety system 918 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. Even further, the safety system may include a safety system responsive to one or more sensed conditions of one or more of the nuclear reactor systems 106 (e.g., external conditions or internal conditions) 920. For instance, a safety system of one or more of the nuclear reactor systems 106, upon sensing a loss of heat sink, may send a signal to a reactor control system of one of the nuclear reactor systems 106. In turn, the reactor control system may establish a nuclear reactor system 106 shutdown and send a corresponding signal to a heat supply system responsive to a shutdown event established by a reactor control system. Then, the heat supply system responsive to a shutdown event established by a reactor control system may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 9B, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to an operation system 922 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from an operation system (e.g., control system, safety system, monitoring system, shutdown system, warning system, or security system) of one or more of the nuclear reactor systems, the heat supply system responsive to a signal from an operation system 924 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, upon receiving a signal from a monitoring system a nuclear reactor system 106 indicating the shutdown of the nuclear reactor system 106, a heat supply system responsive to a signal from an operation system 924 of one or more of the nuclear reactor systems 106 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to a reservoir operation system 926 of one or more of the auxiliary thermal reservoirs 112. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from a reservoir operation system (e.g., control system, safety system, monitoring system) of one or more of the auxiliary thermal reservoirs 112, a heat supply system responsive to a signal from a reservoir operation system 928 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, upon receiving a signal from a monitoring system of an auxiliary thermal reservoir 112 indicating the shutdown of a nuclear reactor system 106 (e.g., energy no longer being diverted to thermal reservoir), the heat supply system responsive to a signal from a reservoir operation system 928 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one of more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In an additional aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to an operator 930 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, in response to a signal from an operator (e.g., human user or human controlled system, such as a programmed computer system), a heat supply system responsive to a signal from an operator 932 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, a heat supply system responsive to a signal from an operator 932, in response to a remote signal, such as wireline or wireless signal from a computer terminal controlled by an operator, may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 9C, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to a pre-selected supply start time 934. For example, the pre-selected supply start time may include the amount of elapsed time relative to a specific event (e.g., shutdown event) or an absolute time. For instance, a heat supply system responsive to a pre-selected supply start time 934, at a pre-selected absolute time (e.g., 5:00 p.m. eastern standard time), may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that historical grid power demand data may be utilized to determine the appropriate time in which to begin transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another instance, a heat supply system responsive to a pre-selected supply start time 934, upon elapse of a pre-selected amount of time from a specific event, such as a nuclear reactor 108 shutdown, may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to the determination of the amount of energy stored in one or more of the auxiliary thermal reservoirs 936. For example, in response to the determination of energy currently stored in an auxiliary thermal reservoir 112, a heat supply system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 936 may initiate transfer of the thermal energy stored in the auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, the heat supply system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 936 may include a heat supply system responsive to the determination of the percentage of energy stored, relative to the overall storage capacity, in the auxiliary thermal reservoir 938. For example, in response to the determination of a set percentage level of stored energy (e.g., 80% of energy storage capacity is being utilized), a heat supply system responsive to the determination of the percentage of stored energy 938 may initiate transfer of the thermal energy stored in one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In an additional aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to the determination of the amount of available storage capacity in one or more of auxiliary thermal reservoirs 940. For example, in response to the determination of available energy storage capacity, a heat supply system responsive to the determination of the amount of available storage capacity in an auxiliary thermal reservoir 940 may initiate transfer of the thermal energy stored the auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, the heat supply system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 940 may include a heat supply system responsive to the determination of the percentage of available energy storage capacity in an auxiliary thermal reservoir 942. For example, in response to the determination of a set percentage level of available energy storage (e.g., 20% storage capacity remains), a heat supply system responsive to the determination of the percentage of available energy storage capacity 942 of an auxiliary thermal reservoir 112 may initiate transfer of the thermal energy stored in the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 9D, one or more of the heat supply systems 114 may include, but are not limited to, a heat supply system suitable for supplying a specified portion of the energy stored in one or more of auxiliary thermal reservoirs to an energy conversion system of one or more of the nuclear reactor systems of the plurality of nuclear reactor systems 944. For example, a heat supply system suitable for supplying a specified portion of the energy stored in an auxiliary thermal reservoir 944 may be utilized to transfer a specified amount of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the amount of energy transferred from an auxiliary thermal reservoir 112 to an energy conversion system 110 may be based on current load demand (e.g., grid demand), where a control system or operator may choose the amount of energy to be transferred to the energy conversion system based on the level of demand that the energy conversion system is currently undergoing. Further, the heat supply system suitable for supplying a specified portion of the energy stored in an auxiliary thermal reservoir to the energy conversion system 944 may include a heat supply system suitable for supplying a specified percentage of the energy stored in the auxiliary thermal reservoir to the energy conversion system 946. For example, a heat supply system suitable for supplying a specified percentage of the energy stored in the auxiliary thermal reservoir to the energy conversion system 946 may be utilized by a control system or operator to transfer a chosen percentage (e.g., 50% of the stored energy) of the energy stored in the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Following are a series of flowcharts depicting implementations. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an example implementation and thereafter the following flowcharts present alternate implementations and/or expansions of the initial flowchart(s) as either sub-component operations or additional component operations building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an example implementation and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular and/or object-oriented program design paradigms. FIG. 10 illustrates an operational flow 1000 representing example operations related to the storage and utilization of energy generated by a plurality of nuclear reactor systems In FIG. 10 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1 through 9D, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1 through 9D. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 1000 moves to a first diverting operation 1010. The first diverting operation 1010 depicts diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112. Then, the additional diverting operation 1020 depicts diverting at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a second energy transfer system 104 may transfer energy from a portion of a second nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more auxiliary thermal reservoirs 112. More generally, an Nth energy transfer system 104 may transfer energy from a portion of an Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more auxiliary thermal reservoirs 112. Then, the supplying operation 1030 depicts supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, one or more heat supply systems 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to one or more of energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems. FIG. 11 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 11 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1102, and/or an operation 1104. Operation 1102 illustrates diverting at least a first portion of excess energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer excess energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1104 illustrates diverting at least a first portion of energy exceeding operational demand of at least one energy conversion system from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer energy exceeding operational demand (e.g., energy in excess of grid requirements) of an energy conversion system associated with a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 from a portion of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. FIG. 12 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 12 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1202. Operation 1202 illustrates diverting a specified percentage of the energy output of a portion of a first nuclear reactor system of a plurality of nuclear reactor systems from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer a specified percentage of the energy output of a portion (e.g., nuclear reactor core or portion of nuclear reactor system in thermal communication with the nuclear reactor core, such as the primary coolant system) of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 from a portion of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. FIGS. 13A and 13B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 13 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1302, an operation 1304, an operation 1306, an operation 1308, an operation 1310, an operation 1312, and/or an operation 1314. Operation 1302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one energy transfer system. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1304 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one energy transfer system. For example, as shown in FIG. 5A, one or more of the energy transfer systems 104 may be suitable for transferring thermal energy 502. For instance, as shown in FIGS. through 9D, a first energy transfer system 104 may transfer thermal energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1306 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5A, one or more of energy transfer systems 104 may include a heat transfer system 504. For instance, as shown in FIG. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems to an auxiliary thermal reservoir 112. Further, the operation 1308 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system, the portion of the first nuclear reactor in thermal communication with at least one heat source of the first nuclear reactor system. For example, as shown in FIG. 5B, heat may be transferred from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the portion of the first nuclear reactor system 106 is in thermal communication with a heat source 522 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a first nuclear reactor system 106 (e.g., coolant system of the nuclear reactor system) in thermal communication with a heat source 522 of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, the operation 1310 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system, the portion of the first nuclear reactor system in thermal communication with at least one nuclear reactor core of the first nuclear reactor system. For example, as shown in FIG. 5B, the heat source 522 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 may include a nuclear reactor core 524. For instance, as shown in FIGS. 1 through 9D, a heat transfer system 504 of a first nuclear reactor system 106 may transfer thermal energy from a portion of the first nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, the operation 1312 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5B, the portion of the first nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 of the first nuclear reactor system 106 may include a portion of the primary coolant system 526 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant system 526 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1314 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant loop of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5B, the portion of the primary coolant system of the first nuclear reactor system 106 may include a portion of a primary coolant loop 528 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant loop 528 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIGS. 14A and 14B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 14 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1402. Operation 1402 illustrates diverting a first selected portion of thermal energy from at least one coolant pool of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5B, the portion of the primary coolant system of the first nuclear reactor system 106 may include a portion of a primary coolant pool 530, such as a liquid metal pool (e.g. liquid sodium) or a liquid metal salt pool (e.g., lithium fluoride pool), of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant pool 530 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIGS. 15A and 15B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 15 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1502. Operation 1502 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one primary coolant system of the first nuclear reactor system in thermal communication with the at least one auxiliary thermal reservoir and at least one secondary coolant system of the first nuclear reactor system, the at least one auxiliary thermal reservoir and the at least one secondary coolant system not in thermal communication. For example, as shown in FIG. 5C, the primary coolant system 526 of the first nuclear reactor system 106 may include a primary coolant system 526 in thermal communication with both an auxiliary thermal reservoir 112 and a secondary coolant system 532 of the first nuclear reactor system 106, wherein the auxiliary thermal reservoir 112 and the secondary coolant system 532 are not in thermal communication with each other. For instance, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant system 526 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the primary coolant system 526 is in thermal communication with both the auxiliary thermal reservoir 112 and a secondary coolant system 532 of the first nuclear reactor system 106, while the auxiliary thermal reservoir 112 and the at least one secondary coolant 532 system are not in thermal communication. FIGS. 16A and 16B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 16 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1602. Further, operation 1602 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir in thermal communication with the at least one primary coolant system of the first nuclear reactor system and at least one secondary coolant system of the first nuclear reactor system. For example, as shown in FIG. 5D, the primary coolant system 526 of the first nuclear reactor system 106 may include a primary coolant system in thermal communication with both an auxiliary thermal reservoir 112 and a secondary coolant system 532 of the first nuclear reactor system 106, wherein the auxiliary thermal reservoir 112 is in thermal communication with the primary coolant system 526 of the nuclear reactor system 106 and the secondary coolant system 532 of the nuclear reactor system 106. For instance, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant system 526 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the auxiliary thermal reservoir 112 is in thermal communication with both the primary coolant system 526 of the nuclear reactor system 106 and a secondary coolant system 532 of the nuclear reactor system 106. FIGS. 17A and 17B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 17 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1702, an operation 1704, and/or an operation 1706. Further, the operation 1702 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one direct fluid exchange heat transfer system. For example, as shown in FIG. 5A, a first energy transfer system 104 of a first nuclear reactor system 106 may include a direct fluid exchange heat transfer system 510. For instance, as shown in FIGS. 1 through 9D, a first direct fluid exchange system 510 may transfer thermal energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1704 illustrates intermixing at least one reservoir fluid of at least one auxiliary thermal reservoir with at least one coolant of a first nuclear reactor system of a plurality of nuclear reactor systems using at least one direct fluid exchange heat transfer system. For example, as shown in FIG. 5A, a first direct fluid exchange system 510 of a first nuclear reactor system 106 may include a system configured to intermix 511 the reservoir fluid of an auxiliary thermal reservoir 112 and the coolant of a nuclear reactor 108 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a system for intermixing 511 the reservoir fluid of an auxiliary thermal reservoir 112 and the reactor coolant of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 may transfer thermal energy from the first nuclear reactor system 106 to the auxiliary thermal reservoir 112 by directly mixing the two fluids. Further, the operation 1706 illustrates intermixing at least one reservoir fluid of at least one auxiliary thermal reservoir with at least one coolant of a first nuclear reactor system of a plurality of nuclear reactor systems using at least one direct fluid exchange heat transfer system, the at least one reservoir fluid substantially similar to the at least one coolant. For example, as shown in FIG. 5A, the auxiliary thermal reservoir fluid and the coolant of the first nuclear reactor system 106 may be substantially similar 512. For instance, the reservoir fluid and the nuclear reactor coolant of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 may both comprise the same liquid metal, such as liquid sodium, liquid lead, or liquid lead bismuth. In another instance, the reservoir fluid and the nuclear reactor coolant may both comprise the same liquid organic, such as diphenyl with diphenyl oxide. FIGS. 18A and 18B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 18 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1802. Further, the operation 1802 illustrates intermixing at least one reservoir fluid of at least one auxiliary thermal reservoir with at least one coolant of at a first nuclear reactor system of a plurality of nuclear reactor systems using at least one direct fluid exchange heat transfer system, the at least one reservoir fluid different from the at least one coolant. For example, as shown in FIG. 5A, the auxiliary thermal reservoir fluid and the coolant of the first nuclear reactor system 106 may be different 513. For instance, the reservoir fluid may comprise a liquid organic fluid (e.g., diphenyl with diphenyl oxide), while the nuclear reactor coolant of a first nuclear reactor system 106 of a plurality of nuclear reactor systems may comprise a liquid metal coolant (e.g., liquid sodium, lead, or lead bismuth). Similarly, the reservoir fluid may comprise a first liquid metal coolant, such as liquid sodium, while the nuclear reactor coolant may comprise a second liquid metal coolant, such as liquid lead. FIGS. 19A and 19B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 19 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1902, and/or an operation 1904. Operation 1902 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat exchanger. For example, as shown in FIG. 5A, a first energy transfer system configured to transfer thermal energy 502 may transfer thermal energy from a portion of the nuclear reactor system 101 to the auxiliary thermal reservoir 112 using one or more reactor-to-reservoir heat exchangers 514. Further, operation 1904 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat exchanger, a first portion of the at least one heat exchanger in thermal communication with a portion of at least one primary coolant system of the first nuclear reactor system and a second portion of the at least one heat exchanger in thermal communication with a portion of the at least one auxiliary thermal reservoir. For example, the reactor-to-reservoir heat exchanger 514 may include a heat exchanger 515 having a first portion in communication with a primary coolant system of the first nuclear reactor system 106 and a second portion in thermal communication with an auxiliary thermal reservoir 112. For instance, the energy transfer system configured to transfer thermal energy 502 may transfer energy from the first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112 using a heat exchanger 515 having a first portion in communication with the primary coolant system of the first nuclear reactor system 106 and a second portion in thermal communication with the auxiliary thermal reservoir 112. FIG. 20 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 20 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2002, an operation 2004, and/or an operation 2006. Operation 2002 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one energy transfer system. For example, as shown in FIG. 5A, a first energy transfer system 104 may include an energy transfer system configured to transfer electrical energy 503 from a portion of a first nuclear reactor system 106 (e.g., an energy conversion system 110 of the first nuclear reactor system 106) to an auxiliary thermal reservoir 112. For instance, an energy transfer system configured to transfer electrical energy 503 from a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112 may be used to transfer electrical energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Those skilled in the art will recognize that in the transfer process the electrical energy originating from a portion of the first nuclear reactor system 106 must be converted to thermal energy in order to be stored in the auxiliary thermal reservoir 112. Further, the operation 2004 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one electrical-to-thermal conversion system. For example, as shown in FIG. 5A, the energy transfer system suitable for transferring electrical energy 503 from a first nuclear reactor system 106 to an auxiliary thermal reservoir 112 may include an electrical energy-to-thermal energy conversion device 516. For instance, an electrical energy-to-thermal energy conversion device 516 may be used to convert electrical energy produced by a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102. The thermal energy may then be transferred to the auxiliary thermal reservoir 112. Further, the operation 2006 illustrates diverting a first selected portion of electrical energy from at least one energy conversion system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one electrical-to-thermal conversion system. For example, as shown in FIG. 5A, the energy transfer system suitable for transferring electrical energy 503 from a first nuclear reactor system 106 to an auxiliary thermal reservoir 112 may include an electrical energy-to-thermal energy conversion device configured to transfer electrical energy from an energy conversion device 110 of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. For instance, an electrical energy-to-thermal energy conversion device configured to transfer electrical energy from an energy conversion device 110 to the auxiliary thermal reservoir 112 may be used to convert electrical energy from the electrical output of an energy conversion device 110 (e.g., turbine-generator system) of the first nuclear reactor system 106 to thermal energy. The thermal energy may then be transferred to the auxiliary thermal reservoir 112. FIG. 21 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 21 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2102, and/or an operation 2104. Operation 2102 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one resistive heating device. For example, as shown in FIG. 5A, the electrical energy-to-thermal energy conversion device may include one or more than one resistive heating devices 517. For instance, a resistive heating device 517 may be utilized to convert electrical energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to thermal energy. The thermal energy may then be transferred to an auxiliary thermal reservoir 112. Further, the operation 2104 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heating coil. For example, as shown in FIG. 5A, the resistive heating device 517 may include one or more heating coils. For instance, a heating coil 518 may be used to convert electrical energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to thermal energy. The thermal energy may then be transferred to an auxiliary thermal reservoir 112. FIG. 22 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 22 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2202, and/or an operation 2204. Operation 2202 illustrates, responsive to at least one condition, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a condition (e.g., power demands on a nuclear reactor system, state of readiness of auxiliary thermal reservoir, thermal properties of nuclear reactor or thermal properties of reservoir), an energy transfer system responsive to a condition 802 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 2204 illustrates responsive to at least one condition of a first nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a condition of a first nuclear reactor system, an energy transfer system responsive to a condition of the first nuclear reactor system 804 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 23 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 23 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2302. Operation 2302 illustrates, responsive to at least one condition of at least one additional nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a condition of an additional nuclear reactor system, an energy transfer system responsive to a condition of an additional nuclear reactor system 806, such as a 2nd nuclear reactor system, a 3rd nuclear reactor system, or up to and including an Nth nuclear reactor system, may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 24 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 24 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2402. Operation 2402 illustrates, responsive to determination of excess capacity of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to the determination of excess capacity of at least one nuclear reactor system 106 of a plurality of nuclear reactor systems 102 (e.g., determination that current nuclear reactor power production exceeds current grid demand), an energy transfer system responsive to the determination of excess nuclear reactor capacity 808 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 25 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 25 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2502, and/or an operation 2504. Operation 2502 illustrates, responsive to at least one operation system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to an operation system (e.g., warning system, security system, or shutdown system) of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to an operation system 810 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 2504 illustrates, responsive to at least one monitoring system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting first a selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a monitoring system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a monitoring system 814 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 26 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 26 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2602. Operation 2602 illustrates, responsive to at least one control system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a control system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a control system 816 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 27 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 27 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2702. Operation 2702 illustrates, responsive to at least one safety system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a safety system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a safety system 818 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 28 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 28 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2802, and/or an operation 2804. Operation 2802 illustrates, responsive to at least one signal from at least one operation system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal (e.g., a digital wireline signal, an analog wireline signal, a digital wireless signal, or an analog wireless signal) from an operation system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 2804 illustrates, responsive to at least one signal from at least one operation system of a first nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal from an operation system of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 29 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 29 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2902. Operation 2902 illustrates, responsive to at least one signal from at least one operation system of at least one additional nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal from an operation system of an additional nuclear reactor system 106 (e.g., second nuclear reactor system 106, third nuclear reactor system, or up to an including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 30 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 30 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3002, and/or an operation 3004. Operation 3002 illustrates, responsive to at least one reservoir operation system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir operation system (e.g., monitoring system, warning system, or control system) of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir operation system 824 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 3004 illustrates, responsive to at least one signal from at least one reservoir operation system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal (e.g., a digital wireline signal, an analog wireline signal, a digital wireless signal, or an analog wireless signal) from a reservoir operation system of an auxiliary thermal reservoir, an energy transfer system responsive to a signal from a reservoir operation system 826 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 31 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 31 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3102. Operation 3102 illustrates, responsive to at least one reservoir monitoring system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir monitoring system (e.g., thermal monitoring system) of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir monitoring system 828 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 32 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 32 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3202. Operation 3202 illustrates, responsive to at least one signal from at least one operator of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to at least one signal from an operator of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal (e.g., wireless or wireline signal) from an operator 820 (e.g., human user or human controlled programmable computer system) may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 33 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 33 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3302. Further, the operation 3302 illustrates, upon a preselected diversion start time, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, upon a preselected diversion start time (e.g., absolute time or time of elapse relative to the occurrence of a predetermined event), an energy transfer system responsive to a preselected diversion start time 822 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 34 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 34 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3402, and/or an operation 3404. Operation 3402 illustrates, responsive to a shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a shutdown event 834 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 3404 illustrates, responsive to a scheduled shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a scheduled shutdown event (e.g., shutdown for routine maintenance) of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a scheduled shutdown event 836 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 35 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 35 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3502. Operation 3502 illustrates, responsive to an emergency shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to an emergency shutdown event (e.g., SCRAM) of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to an emergency shutdown event 838 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 36 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 36 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3602. Operation 3602 illustrates, responsive to a shutdown event of a first nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a shutdown event 834 of the first nuclear reactor system 106 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 37 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 37 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3702. Operation 3702 illustrates, responsive to a shutdown event of at least one additional nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of an additional nuclear reactor system 106 (e.g., the second nuclear reactor system, the third nuclear reactor system 106, or up to an including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a shutdown event 834 of the additional nuclear reactor system 106 may initiate transfer of energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 38 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 38 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3802. Further, the operation 3802 illustrates, responsive to a shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, establishing thermal communication between a portion of a first nuclear reactor system of the plurality of nuclear reactor systems and at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system configured to establish thermal communication between a first nuclear reactor system and an auxiliary thermal reservoir 840 may establish thermal communication between a portion of the first nuclear reactor system (e.g., primary coolant system) of the plurality of nuclear reactor systems and the auxiliary thermal reservoir 112. FIG. 39 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 39 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3902. Operation 3902 illustrates, preceding shutdown of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, prior to shutdown of a nuclear reactor 108 of a nuclear reactor system 106, an energy transfer system responsive to a shutdown event 834 of the nuclear reactor system 106 may initiate the transfer of energy from a portion of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIG. 40 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 40 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4002, and/or an operation 4004. Operation 4002 illustrates, during shutdown of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, during shutdown of a nuclear reactor 108 of a nuclear reactor system 106, an energy transfer system responsive to a shutdown event 834 of the nuclear reactor system 106 may initiate the transfer of energy from a portion of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the operation 4004 illustrates, following shutdown of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, following shutdown of a nuclear reactor 108 of a nuclear reactor system 106, an energy transfer system responsive to a shutdown event 834 of the nuclear reactor system 106 may initiate the transfer of energy from a portion of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIG. 41 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 41 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4102, and/or an operation 4104. Operation 4102 illustrates, responsive to determination of the amount of energy stored in at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the amount of energy stored in an auxiliary thermal reservoir 842 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 4104 illustrates, responsive to determination of the percentage of energy stored in at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the percentage of energy stored in an auxiliary thermal reservoir 844 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 42 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 42 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4202, and/or an operation 4204. Operation 4202 illustrates, responsive to determination of the amount of available energy storage capacity of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the amount of available energy storage capacity of at least one auxiliary thermal reservoir 846 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 4204 illustrates, responsive to determination of the percentage of available energy storage capacity of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the percentage of available energy storage capacity of at least one auxiliary thermal reservoir 848 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 43 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 43 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4302, and/or an operation 4304. The operation 4302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112. Further, the operation 4304 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one solid heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of a solid heat storage material 414, such a solid object (e.g., solid ceramic object, solid metal object, or solid stone object) or a particulate solid (e.g., sand), of an auxiliary thermal reservoir 112. FIG. 44 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 44 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4402, and/or an operation 4404. Operation 4402 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one liquid heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid material 402 (e.g., liquid metal, liquid metal salt, liquid organic, or liquid water) of an auxiliary thermal reservoir 112. Further, the operation 4404 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one organic liquid heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid organic material 404 (e.g., diphenyl with diphenyl oxide) of an auxiliary thermal reservoir 112. FIG. 45 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 45 illustrates example embodiments where the diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4502. Further, the operation 4502 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one liquid metal salt heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid metal salt 406 (e.g., lithium fluoride) of an auxiliary thermal reservoir 112. FIG. 46 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 46 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4602. Operation 4602 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one liquid metal heat storage material of at least one auxiliary thermal reservoir.]. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid metal 408 (e.g., sodium) of the auxiliary thermal reservoir 112. FIG. 47 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 47 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4702. Operation 4702 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of liquid water of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid water 410 of an auxiliary thermal reservoir 112. FIG. 48 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 48 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4802. Further, the operation 4802 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one pressurized gaseous mass of material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of pressurized gaseous material 412 (e.g., pressurized helium or pressurized carbon dioxide) of the auxiliary thermal reservoir 112. FIG. 49 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 49 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4902. Operation 4902 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one mixed phase material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of a mixed phase material 420 (e.g., steam water-liquid water) of the auxiliary thermal reservoir 112. FIG. 50 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 50 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5002. Further, the operation 5002 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one material of at least one auxiliary thermal reservoir, the mass of at least one material having a phase transition within the operating temperature of the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of a material having a phase transition within the operating temperature 422 of the auxiliary thermal reservoir 112. FIG. 51 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 51 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5102, an operation 5104, and/or an operation 5106. Operation 5102 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in a reservoir containment system. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112 contained in a reservoir containment system 424 (e.g., vessel). Further, the operation 5104 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external vessel. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112 contained in an external vessel 426. Further, the operation 5106 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external high pressure gas vessel. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112 contained in a high pressure gas vessel 430. For instance, the energy transfer system 104 may transfer a selected portion of energy from a portion a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of high pressurized gaseous helium contained in an external high pressure helium vessel. FIG. 52 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 52 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5202. Operation 5202 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external liquid vessel. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of the auxiliary thermal reservoir 112 contained in an external liquid vessel 428. For instance, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid water contained in an external water vessel. FIG. 53 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 53 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5302. Further, the operation 5302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external liquid pool. For example, as shown in FIGS. 1 through 9D, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid heat storage material 402 of the auxiliary thermal reservoir 112 contained in an external liquid pool 434. For instance, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid sodium contained in an external liquid sodium pool. FIG. 54 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 54 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5402. Operation 5402 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a temperature change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir stores the energy in the form of an increase in temperature of the heat storage material 436. For instance, the energy transferred to the auxiliary thermal reservoir 112 may cause a liquid heat storage material 402 to increase in temperature from 100° C. to 200° C. FIG. 55 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 55 illustrates example embodiments where the operation 1010 may include at least one additional operation. Additional operations may include an operation 5502, and/or an operation 5504. Operation 5502 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a phase change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir stores the energy in the form of a phase change in the heat storage material 438. For instance, the energy transferred to the auxiliary thermal reservoir 112 may cause a solid reservoir material to undergo a phase change into a liquid reservoir material, where the energy is stored in the reservoir material as a latent heat of transformation. Further, the operation 5504 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a solid-liquid phase change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir 112 stores the energy in the form of a solid-liquid phase change 440 (e.g., solid sodium-liquid sodium phase change). FIG. 56 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 56 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5602. Operation 5602 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a liquid-gas phase change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir 112 stores the energy in the form of a liquid-gas phase change 442 (e.g., liquid water-steam water phase change). FIG. 57 illustrates an operational flow 5700 representing example operations related to the storage and utilization of energy generated by a plurality of nuclear reactor systems. FIG. 57 illustrates an example embodiment where the example operational flow 1000 of FIG. 10 may include at least one additional operation. Additional operations may include an operation 5710, and/or an operation 5712. After a start operation, a first diverting operation 1010, an additional diverting operation 1020, and a supplying operation 1030, the operational flow 5700 moves to a temperature maintaining operation 5710. Operation 5710 illustrates maintaining the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above a selected temperature. For example, as shown in FIG. 4D, the temperature of a heat storage material 111 of an auxiliary thermal reservoir 112 may be maintained with a reservoir temperature control system 454 (e.g., thermostat). The operation 5712 illustrates maintaining the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above the melting temperature of the at least one heat storage material. For example, as shown in FIG. 4D, the temperature of a heat storage material 111 of an auxiliary thermal reservoir 112 may be maintained with a reservoir temperature control system 454 above a specified temperature, such as the melting temperature of the heat storage material 111. FIG. 58 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 58 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5802, and/or an operation 5804. The operation 5802 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid cooled 302 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 5804 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid metal salt coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid metal salt (e.g., lithium fluoride or other fluoride salts) cooled 304 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 59 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 59 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5902. Operation 5902 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid metal coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid metal (e.g., liquid sodium or liquid lead) cooled 306 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 60 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 60 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6002. Operation 6002 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid organic coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid organic (e.g., diphenyl with diphenyl oxide) cooled 308 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 61 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 61 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6102. Operation 6102 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid water coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid water cooled 310 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 62 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 62 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6202, and/or an operation 6204. Operation 6202 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one pressurized gas coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first pressurized gas (e.g., pressurized helium or carbon dioxide) cooled 312 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. The operation 6204 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one mixed phase coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first mixed phase (e.g., liquid water-steam water) cooled 314 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 63 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 63 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6302, and/or an operation 6304. Operation 6302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a thermal spectrum nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a thermal spectrum nuclear reactor 202. The operation 6304 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a fast spectrum nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a fast spectrum nuclear reactor 204. FIG. 64 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 64 illustrates example embodiments where the diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6402, and/or an operation 6404. Operation 6402 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a multi-spectrum nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a multi-spectrum nuclear reactor 206. The operation 6404 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a breeder nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a breeder nuclear reactor 208. FIG. 65 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 65 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6502. Operation 6502 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a traveling wave nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a traveling wave nuclear reactor 210. FIG. 66 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 66 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 6602, and/or an operation 6604. Operation 6602 illustrates supplying at least a portion of thermal energy from a first auxiliary thermal reservoir and a portion of thermal energy from at least a second thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, thermal energy stored in a first auxiliary reservoir 112 and thermal energy stored in an additional thermal reservoir (e.g., second thermal reservoir, third thermal reservoir, or up to and including an Nth thermal reservoir) may be supplied to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. For instance, a first heat supply system 114 may supply thermal energy stored in the first auxiliary thermal reservoir 112 to an energy conversion system 110 and a second heat supply system 114 may supply thermal energy stored in the second auxiliary thermal reservoir 112 to the energy conversion system 110. Further, the operation 6604 illustrates supplying at least a portion of thermal energy from a first auxiliary thermal reservoir and a portion of thermal energy from at least a second thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems, the first auxiliary thermal reservoir and the at least a second thermal reservoir not in thermal communication. For example, as shown in FIGS. 1 through 9D, thermal energy stored in a first auxiliary reservoir 112 and thermal energy stored in an additional thermal reservoir (e.g., second thermal reservoir, third thermal reservoir, or up to and including an Nth thermal reservoir) may be supplied to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102, wherein the first thermal reservoir 112 and the second thermal reservoir 112 are not in thermal communication. FIG. 67 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 67 illustrates example embodiments where the supply operation 1030 may include at least one additional operation. Additional operations may include an operation 6702. Operation 6702 illustrates supplying at least a portion of thermal energy from a first auxiliary thermal reservoir and a portion of thermal energy from at least a second thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems, the first auxiliary thermal reservoir and the at least a second thermal reservoir in thermal communication. For example, as shown in FIGS. 1 through 9D, thermal energy stored in a first auxiliary reservoir 112 and thermal energy stored in an additional thermal reservoir (e.g., second thermal reservoir, third thermal reservoir, or up to and including an Nth thermal reservoir) may be supplied to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102, wherein the first thermal reservoir 112 and the second thermal reservoir 112 are in thermal communication. It will be recognized by those skilled in the art that even though the first thermal reservoir 110 and the second thermal reservoir 110 are thermally coupled the two reservoirs can for practical purposes be treated as two distinct thermal reservoirs under non-steady state conditions. FIG. 68 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 68 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 6802, and/or an operation 6804. The operation 6802 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat supply system. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 (e.g., topping cycle 226 or turbine 218) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 6804 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat exchange loop. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more heat exchange loops 602. FIG. 69 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 69 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 6902. Operation 6902 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat exchange pipe. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more heat pipes 604. FIG. 70 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 70 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7002. Operation 7002 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat exchanger. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more heat exchangers 606. For instance, a first portion of a heat exchanger 606 may be in thermal communication with an auxiliary thermal reservoir 112, while the second portion of the heat exchanger 606 may be in thermal communication with an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 71 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 71 illustrates example embodiments where the supply operation 1030 may include at least one additional operation. Additional operations may include an operation 7102. Operation 7102 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one thermoelectric device. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more thermoelectric devices 608. For instance, a first portion of a thermoelectric device 608 may be in thermal communication with an auxiliary thermal reservoir 112 and a second portion of the thermoelectric device 608 may be in thermal communication with a heat sink (e.g., environmental heat sink) of a nuclear reactor system 106 of the plurality of the nuclear reactor systems 102. FIG. 72 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 72 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7202, and/or an operation 7204. Operation 7202 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one primary energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a primary energy conversion system 212 (e.g., energy conversion system coupled to the primary boiling loop) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7204 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one auxiliary energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an auxiliary energy conversion system 214 (e.g., energy conversion system coupled to a non-primary boiling) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 73 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 73 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7302, and/or an operation 7304. Operation 7302 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one emergency energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an emergency energy conversion system 216 (e.g., energy conversion system supplying electric power to various operation systems of the nuclear reactor system) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7304 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one boiling loop of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a boiling loop 232 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 74 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 74 illustrates example embodiments where the supply operation 1030 may include at least one additional operation. Additional operations may include an operation 7402, and/or an operation 7404. The operation 7402 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one turbine of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a turbine 218 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7404 illustrates [supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one working fluid of at least one turbine of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to the working fluid of a turbine 224 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 75 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 75 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7502, and/or an operation 7504. The operation 7502 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one low grade heat dump. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a low grade heat dump 230 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7504 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one topping cycle. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a topping cycle 226 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 76 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 76 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7602, and/or an operation 7604. The operation 7602 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one bottoming cycle. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a bottoming cycle 228 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7604 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of the first nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 77 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 77 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7702. The operation 7702 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of the at least one additional nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of an additional nuclear reactor system 106 (e.g., a second nuclear reactor system 106, a third nuclear reactor system 106 or up to and including an Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102. FIG. 78 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 78 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7802, an operation 7804, and/or an operation 7806. The operation 7802 illustrates, responsive to at least one condition, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a condition 902 (e.g., grid demand, thermal properties of one or more of the auxiliary thermal reservoirs) may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7804 illustrates, responsive to at least one condition of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a condition of one or more of the nuclear reactor systems 904 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7806 illustrates, responsive to heightened power demand on at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to heightened power demand of one or more of the nuclear reactor systems 906 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 79 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 79 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7902, and/or an operation 7904. Operation 7902 illustrates, responsive to at least one operation system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to an operation system (e.g., monitoring system, control system, safety system, or security system) of a nuclear reactor system 922 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7904 illustrates, responsive to at least one signal from at least one operation system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a signal (e.g., wireless or wireline) from an operation system of a nuclear reactor system 924 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 80 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 80 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8002, and/or an operation 8004. Operation 8002 illustrates, responsive to at least one reservoir operation system of the at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a reservoir operation system 926 (e.g., reservoir monitoring system, reservoir control system, or reservoir safety system) may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8004 illustrates responsive to at least one signal from at least one reservoir operation system of at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a signal (e.g., wireless or wireline) from a reservoir operation system 928 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 81 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 81 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8102, and/or an operation 8104. Operation 8102 illustrates, responsive to at least one operator of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to an operator (e.g., human or human programmed computer control system) of a nuclear reactor system of the plurality of nuclear reactor systems 930 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8104 illustrates responsive to at least one signal from at least one operator of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a signal from an operator of a nuclear reactor system of the plurality of nuclear reactor systems 932 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 82 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 82 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8202, and/or an operation 8204. Operation 8202 illustrates, responsive to a shutdown event of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8204 illustrates, responsive to a scheduled shutdown event of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a scheduled shutdown event (e.g., shutdown for routine maintenance) of a nuclear reactor system of the plurality of nuclear reactor systems 912 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 83 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 83 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8302. Operation 8302 illustrates, responsive to an emergency shutdown event of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to an emergency shutdown event (e.g., SCRAM) of a nuclear reactor system of the plurality of nuclear reactor systems 910 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 84 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 84 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8402. Operation 8402 illustrates, preceding shutdown of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, preceding the shutdown of a nuclear reactor system 106, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 85 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 85 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8502. Operation 8502 illustrates, following shutdown of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, following the shutdown of a nuclear reactor system 106, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIGS. 86A and 86B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 86 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8602, an operation 8604, an operation 8606, and/or an operation 8608. Operation 8602 illustrates, responsive to a shutdown event established by at least one operation system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems established by an operation system of a nuclear reactor system 914 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8604 illustrates, responsive to a shutdown event established by at least one reactor control system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems established by a reactor control system of a nuclear reactor system 916 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8606 illustrates, responsive to a shutdown event established by at least one reactor control system responsive to at least one signal from at least one safety system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems established by a reactor control system that is responsive to a safety system of a nuclear reactor system 918 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8608 illustrates, responsive to a shutdown event established by at least one reactor control system responsive to at least one signal from at least one safety system of at least one nuclear reactor system of the plurality of nuclear reactor systems, the safety system responsive to at least one sensed condition, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 920 responsive to a shutdown event of a nuclear reactor system established by a reactor control system that is responsive to a safety system, where the safety system is responsive to a sensed condition (e.g., external condition or internal condition) of a nuclear reactor system, may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 87 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 87 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8702. Operation 8702 illustrates, upon a pre-selected supply start time, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to the elapse of a preselected supply start time 934 (e.g., time of elapse measured relative to the initiation of a nuclear reactor system or system shutdown event or absolute time) may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 88 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 88 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8802, and/or an operation 8804. Further, the operation 8802 illustrates, responsive to determination of the amount of energy stored in at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the amount of energy stored in an auxiliary thermal reservoir 936 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8804 illustrates, responsive to determination of the percentage of energy stored in at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the percentage of energy stored in an auxiliary thermal reservoir 938 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 89 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 89 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8902, and/or an operation 8904. Operation 8902 illustrates, responsive to determination of the amount of available energy storage capacity of at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the amount of available energy storage capacity of an auxiliary thermal reservoir 940 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8904 illustrates, responsive to determination of the percentage of available energy storage capacity of at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the percentage of available energy storage capacity of an auxiliary thermal reservoir 942 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 90 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 90 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 9002, and/or an operation 9004. The operation 9002 illustrates supplying a specified portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system suitable for supplying a specified portion of the energy stored in an auxiliary thermal reservoir to an energy conversion system 944 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may initiate the transfer of a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 9004 illustrates supplying a specified percentage of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system suitable for supplying a specified percentage of the energy stored in an auxiliary thermal reservoir to an energy conversion system 946 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may initiate the transfer of a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 91 illustrates an operational flow 9100 representing example operations related to storage and utilization of energy generated by a plurality of nuclear reactor systems. FIG. 91 illustrates an example embodiment where the example operational flow 1000 of FIG. 10 may include at least one additional operation. Additional operations may include an operation 9110, and/or an operation 9112. After a start operation, a first diverting operation 1010, an additional diverting operation 1020, and a supplying operation 1030, the operational flow 9100 moves to a supplementing operation 9110. Operation 9110 illustrates supplementing the at least one auxiliary thermal reservoir with an additional portion of thermal energy from at least one additional energy source. For example, as shown in FIG. 7, the thermal energy stored in an auxiliary thermal reservoir 112 may be supplemented with an additional portion of energy transferred from an additional energy source 702, such as a non-nuclear energy source (e.g., coal powered generator, diesel powered generator, or solar cell array) via a supplementary energy transfer system 706. The operation 9112 illustrates supplementing the at least one auxiliary thermal reservoir with an additional portion of energy from at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIG. 7, the thermal energy stored in an auxiliary thermal reservoir 112 may be supplemented with an additional portion of energy transferred from an energy conversion device 110 of a nuclear reactor system of the plurality of nuclear reactor systems 102 the via a supplementary energy transfer system 706. FIG. 92 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 92 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 9202. Operation 9202 illustrates, responsive to at least one reservoir control system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir control system (e.g., thermal control system) of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir control system 830 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 93 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 93 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 9302. Operation 9302 illustrates, responsive to at least one reservoir safety system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir safety system of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir safety system 832 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 94 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 94 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 9402. Further, the operation 9402 illustrates, during shutdown of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, during the shutdown of a nuclear reactor system 106, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 95 illustrates an operational flow 9500 representing example operations related to storage and utilization of energy generated by a plurality of nuclear reactor systems. FIG. 95 illustrates an example embodiment where the example operational flow 1000 of FIG. 10 may include at least one additional operation. Additional operations may include an operation 9510, an operation 9512, and/or an operation 9514. After a start operation, a first diverting operation 1010, an additional diverting operation 1020, and a supplying operation 1030, the operational flow 9500 moves to a monitoring operation 9510. Operation 9510 illustrates monitoring at least one condition of the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a condition, such as the operational status (e.g., state of readiness, temperature pressure, or storage capacity), of an auxiliary thermal reservoir 112 may be monitored. Further, the operation 9512 illustrates monitoring at least one condition of the at least one auxiliary thermal reservoir using at least one reservoir monitoring system. For example, as shown in FIG. 4D, a reservoir monitoring system 444 maybe used to monitor a condition of an auxiliary thermal reservoir 112. Further, the operation 9514 illustrates monitoring the temperature of the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a reservoir temperature monitoring system 446 maybe used to monitor the temperature of an auxiliary thermal reservoir 112. FIG. 96 illustrates alternative embodiments of the example operational flow 9500 of FIG. 95. FIG. 96 illustrates example embodiments where the monitoring operation 9510 may include at least one additional operation. Additional operations may include an operation 9602, and/or an operation 9604. The operation 9602 illustrates monitoring the pressure of the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a reservoir pressure monitoring system 448 maybe used to monitor the pressure of an auxiliary thermal reservoir 112. The operation 9604 illustrates determining the amount of energy stored in the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a system configured to determine the amount of stored energy 450 in an auxiliary thermal reservoir 112 may be utilized to monitor the energy storage level in the auxiliary thermal reservoir 112. FIG. 97 illustrates alternative embodiments of the example operational flow 9500 of FIG. 95. FIG. 97 illustrates example embodiments where the monitoring operation 9510 may include at least one additional operation. Additional operations may include an operation 9702. The operation 9702 illustrates determining the amount of available energy storage capacity in the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a system configured to determine the amount of available energy storage capacity 452 in an auxiliary thermal reservoir 112 may be utilized to monitor the available energy storage capacity of the auxiliary thermal reservoir 112. Those having skill in the art will recognize that the state of the art has progressed to the point where there is little distinction left between hardware, software, and/or firmware implementations of aspects of systems; the use of hardware, software, and/or firmware is generally (but not always, in that in certain contexts the choice between hardware and software can become significant) a design choice representing cost vs. efficiency tradeoffs. Those having skill in the art will appreciate that there are various vehicles by which processes and/or systems and/or other technologies described herein can be effected (e.g., hardware, software, and/or firmware), and that the preferred vehicle will vary with the context in which the processes and/or systems and/or other technologies are deployed. For example, if an implementer determines that speed and accuracy are paramount, the implementer may opt for a mainly hardware and/or firmware vehicle; alternatively, if flexibility is paramount, the implementer may opt for a mainly software implementation; or, yet again alternatively, the implementer may opt for some combination of hardware, software, and/or firmware. Hence, there are several possible vehicles by which the processes and/or devices and/or other technologies described herein may be effected, none of which is inherently superior to the other in that any vehicle to be utilized is a choice dependent upon the context in which the vehicle will be deployed and the specific concerns (e.g., speed, flexibility, or predictability) of the implementer, any of which may vary. Those skilled in the art will recognize that optical aspects of implementations will typically employ optically-oriented hardware, software, and or firmware. In some implementations described herein, logic and similar implementations may include software or other control structures. Electronic circuitry, for example, may have one or more paths of electrical current constructed and arranged to implement various functions as described herein. In some implementations, one or more media may be configured to bear a device-detectable implementation when such media hold or transmit device-detectable instructions operable to perform as described herein. In some variants, for example, implementations may include an update or modification of existing software or firmware, or of gate arrays or programmable hardware, such as by performing a reception of or a transmission of one or more instructions in relation to one or more operations described herein. Alternatively or additionally, in some variants, an implementation may include special-purpose hardware, software, firmware components, and/or general-purpose components executing or otherwise invoking special-purpose components. Specifications or other implementations may be transmitted by one or more instances of tangible transmission media as described herein, optionally by packet transmission or otherwise by passing through distributed media at various times. Alternatively or additionally, implementations may include executing a special-purpose instruction sequence or invoking circuitry for enabling, triggering, coordinating, requesting, or otherwise causing one or more occurrences of virtually any functional operations described herein. In some variants, operational or other logical descriptions herein may be expressed as source code and compiled or otherwise invoked as an executable instruction sequence. In some contexts, for example, implementations may be provided, in whole or in part, by source code, such as C++, or other code sequences. In other implementations, source or other code implementation, using commercially available and/or techniques in the art, may be compiled/implemented/translated/converted into a high-level descriptor language (e.g., initially implementing described technologies in C or C++ programming language and thereafter converting the programming language implementation into a logic-synthesizable language implementation, a hardware description language implementation, a hardware design simulation implementation, and/or other such similar mode(s) of expression). For example, some or all of a logical expression (e.g., computer programming language implementation) may be manifested as a Verilog-type hardware description (e.g., via Hardware Description Language (HDL) and/or Very High Speed Integrated Circuit Hardware Descriptor Language (VHDL)) or other circuitry model which may then be used to create a physical implementation having hardware (e.g., an Application Specific Integrated Circuit). Those skilled in the art will recognize how to obtain, configure, and optimize suitable transmission or computational elements, material supplies, actuators, or other structures in light of these teachings. The foregoing detailed description has set forth various embodiments of the devices and/or processes via the use of block diagrams, flowcharts, and/or examples. Insofar as such block diagrams, flowcharts, and/or examples contain one or more functions and/or operations, it will be understood by those within the art that each function and/or operation within such block diagrams, flowcharts, or examples can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or virtually any combination thereof. In one embodiment, several portions of the subject matter described herein may be implemented via Application Specific Integrated Circuits (ASICs), Field Programmable Gate Arrays (FPGAs), digital signal processors (DSPs), or other integrated formats. However, those skilled in the art will recognize that some aspects of the embodiments disclosed herein, in whole or in part, can be equivalently implemented in integrated circuits, as one or more computer programs running on one or more computers (e.g., as one or more programs running on one or more computer systems), as one or more programs running on one or more processors (e.g., as one or more programs running on one or more microprocessors), as firmware, or as virtually any combination thereof, and that designing the circuitry and/or writing the code for the software and or firmware would be well within the skill of one of skill in the art in light of this disclosure. In addition, those skilled in the art will appreciate that the mechanisms of the subject matter described herein are capable of being distributed as a program product in a variety of forms, and that an illustrative embodiment of the subject matter described herein applies regardless of the particular type of signal bearing medium used to actually carry out the distribution. Examples of a signal bearing medium include, but are not limited to, the following: a recordable type medium such as a floppy disk, a hard disk drive, a Compact Disc (CD), a Digital Video Disk (DVD), a digital tape, a computer memory, etc.; and a transmission type medium such as a digital and/or an analog communication medium (e.g., a fiber optic cable, a waveguide, a wired communications link, a wireless communication link (e.g., transmitter, receiver, transmission logic, reception logic, etc.), etc.). In a general sense, those skilled in the art will recognize that the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, and/or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, electro-magnetically actuated devices, and/or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, a Micro Electro Mechanical System (MEMS), etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.), and/or any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, medical devices, as well as other systems such as motorized transport systems, factory automation systems, security systems, and/or communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, and/or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Those skilled in the art will recognize that at least a portion of the devices and/or processes described herein can be integrated into a data processing system. Those having skill in the art will recognize that a data processing system generally includes one or more of a system unit housing, a video display device, memory such as volatile or non-volatile memory, processors such as microprocessors or digital signal processors, computational entities such as operating systems, drivers, graphical user interfaces, and applications programs, one or more interaction devices (e.g., a touch pad, a touch screen, an antenna, etc.), and/or control systems including feedback loops and control motors (e.g., feedback for sensing position and/or velocity; control motors for moving and/or adjusting components and/or quantities). A data processing system may be implemented utilizing suitable commercially available components, such as those typically found in data computing/communication and/or network computing/communication systems. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken limiting. Although a user is shown/described herein as a single illustrated figure, those skilled in the art will appreciate that the user may be representative of a human user, a robotic user (e.g., computational entity), and/or substantially any combination thereof (e.g., a user may be assisted by one or more robotic agents) unless context dictates otherwise. Those skilled in the art will appreciate that, in general, the same may be said of “sender” and/or other entity-oriented terms as such terms are used herein unless context dictates otherwise. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that such terms (e.g., “configured to”) can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B. With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. |
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description | FIG. 1 is an illustration of several related art nuclear fuel bundles 10 and core components commonly encountered in existing nuclear power technology. As shown in FIG. 1, one or more fuel bundles 10 containing several individual fuel rods may be placed within a reactor core in conventional fuel placement strategies. A channel 20 may surround the fuel rods in each bundle 10, providing directed coolant and/or moderator flow within bundles 10 and/or facilitating manipulation of bundles 10 as a single rigid body. Control rods or cruciform control blades 60 may be extended from set core locations between bundles to absorb neutrons and control reactivity and ultimately control reactivity by a degree of insertion or withdrawal from between the fuel bundles 10. Fuel support 70 may support and align bundles 10 at constant positions within the core. FIG. 2 is a quadrant map of a related art Boiling Water Reactor (BWR) core, illustrating fuel bundle locations within one quarter of the core. Reactor cores typically are conveniently symmetrical about at least two perpendicular axes, such that a quadrant map of FIG. 2 conveys the makeup of the entire core. As shown in FIG. 2, individual bundle locations are occupied by fresh (shown with diagonal or cross-hatched fill) or burnt (shown with no fill) fuel bundles at the start of a fuel cycle, before commencement of power operations in the core. Fresh bundles are bundles that have not previously been exposed to neutron flux during power operations, i.e., never been burnt, whereas burnt fuel bundles have received such exposure, typically over one or more fuel cycles lasting 1-2 years. As such, burnt fuel bundles typically have exposure, or burnup, of several GWd/ST. Fresh fuel bundles may have different starting enrichments of fissile material content. For example, in some BWR designs, outer-enrichment bundles (shown in cross-hatched fill) may include approximately 4.3% Uranium-235 fuel, and inner-enrichment bundles (shown in diagonal fill) may include approximately 4.2% Uranium-235 fuel. Varying enrichments, such as the one shown in FIG. 2, may permit a flatter radial power profile in the core and/or achieve other operational effects. Further, in some BWR designs, bundles may also possess varying distributions and concentrations of burnable poisons/neutron absorbers to suppress reactivity and optimize operational characteristics. As shown in FIG. 2, at startup related art nuclear fuel cores include an outer peripheral ring of stale fuel bundles surrounding an inner peripheral ring of fresh, high-enrichment fuel bundles. A central region may include 50% or more fresh bundles in order to maximize fresh fuel content over an even distribution, permitting longer operating cycles with lower downtime. In related art BWRs, cruciform control blades 60 extend centrally between groupings of four fuel bundles in order to absorb neutrons and control the nuclear chain reaction in the core. As shown in FIG. 2, the groupings of four fuel bundles, between which control blades extend, are identified in bolded outline as controlled bundles, or control cells. Bundles within the controlled bundle groups conventionally have one face closest to a control blade used during the fuel cycle; such bundles are referred to as controlled bundles and their positions as controlled positions in control cells of four bundles. Different control blades in different control cells, usually four or five per quadrant, are conventionally alternately inserted and withdrawn in different and complex control blade sequences in order to manage reactivity and power distribution and spread control blade usage across several different blades and fresh fuel bundles within the core. As shown in FIG. 2, in order to maximize the number of fresh fuel bundles used in a longer cycle over an even core distribution, several fresh fuel bundles may be placed in controlled positions adjacent to operated control blades within the inner portion of the core. Due to conventional operation of control blades, all fresh fuel bundles in the central core portion may be controlled—having direct exposure to control blades actively moved to finely control reactivity—throughout an entire fuel cycle. Use of fresh fuel bundles in controlled locations causes several problems, including corrosion and channel bowing that worsens in later cycles, and a need to perform complex and/or lower-power control blade sequence exchanges due to this positioning that worsen plant economics. Some related art fuel cores have avoided this problem by using a Control Cell Core loading strategy, where only burnt fuel bundles are placed closest to operated control blades, resulting in fewer fresh bundles used in the central portion of the core and shorter operating cycles. Example embodiments include nuclear cores with at least two control cell types that differ in total reactivity. The different control cell types may be placed in numbers and/or positions the enhance fuel and core performance. Example cores may include an outermost region with lower reactivity fuel bundles, an inner peripheral region lining the outer peripheral region and having higher reactivity fuel bundles and at least portions of the outermost control cells, and an inner core lining the inner peripheral region and having inner control cells with only fuel bundles of lower reactivity. The lower reactivity bundles may be burnt, and the higher reactivity bundles may be fresh, for example, the outer control cells can include two fresh fuel bundles and the inner control cells can include only burnt fuel bundles. However, reactivity differences may also be achieved through fuel enrichment variation, burnable poison presence, etc. In an example with a conventional BWR, the inner peripheral region may be three bundles thick, most of which can be higher reactivity fuel bundles, and the outer peripheral region may be three bundles thick. In this instance, there may be thirteen inner control cells. Example embodiments are not limited to BWRs or specific placements, but are compatible with any type of core control cell setup, including control cells formed with control rods or cruciform control blades having four fuel bundles positioned in each corner the blades. Different core geometries are easily outfitted with example embodiments; for example, in an ESBWR, the inner core region may have twenty-five inner control cells. Example methods include creating and/or operating nuclear cores with multiple types of control cells. For example, a core may be loaded to form an example embodiment core. In example methods, control elements in only the inner control cells may be moved to control core reactivity, except at sequence exchanges after several weeks or months of operation, such as after 3 GWd/ST. At such a sequence exchange, a single coarse movement of control elements in the outer control cells may be made in order to resume controlling day-to-day reactivity with the inner control cells. Near the end of a cycle, when reactivity is lowest, reactivity in the core may be controlled only with inner control cells, and control elements in the outer control cells can be fully withdrawn. Example embodiments and methods can provide high (approximately 50%) fresh fuel volumes for each cycle, enabling longer cycles and better plant economics. Example methods and embodiments further provide high power density and low leakage through segregating fuel types by reactivity in the periphery and inner portions of the core. Example methods and embodiments further may enable simplified and non-interrupting movement of control elements in the inner core to fully control reactivity without causing negative control element and fuel interactions. This is a patent document, and general broad rules of construction should be applied when reading and understanding it. Everything described and shown in this document is an example of subject matter falling within the scope of the appended claims. Any specific structural and functional details disclosed herein are merely for purposes of describing how to make and use example embodiments or methods. Several different embodiments not specifically disclosed herein fall within the claim scope; as such, the claims may be embodied in many alternate forms and should not be construed as limited to only example embodiments set forth herein. It will be understood that, although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of example embodiments. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that when an element is referred to as being “connected,” “coupled,” “mated,” “attached,” or “fixed” to another element, it can be directly connected or coupled to the other element or intervening elements may be present. In contrast, when an element is referred to as being “directly connected” or “directly coupled” to another element, there are no intervening elements present. Other words used to describe the relationship between elements should be interpreted in a like fashion (e.g., “between” versus “directly between”, “adjacent” versus “directly adjacent”, etc.). Similarly, a term such as “communicatively connected” includes all variations of information exchange routes between two devices, including intermediary devices, networks, etc., connected wirelessly or not. As used herein, the singular forms “a”, “an” and “the” are intended to include both the singular and plural forms, unless the language explicitly indicates otherwise with words like “only,” “single,” and/or “one.” It will be further understood that the terms “comprises”, “comprising,”, “includes” and/or “including”, when used herein, specify the presence of stated features, steps, operations, elements, ideas, and/or components, but do not themselves preclude the presence or addition of one or more other features, steps, operations, elements, components, ideas, and/or groups thereof. It should also be noted that the structures and operations discussed below may occur out of the order described and/or noted in the figures. For example, two operations and/or figures shown in succession may in fact be executed concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. Similarly, individual operations within example methods described below may be executed repetitively, individually or sequentially, so as to provide looping or other series of operations aside from the single operations described below. It should be presumed that any embodiment having features and functionality described below, in any workable combination, falls within the scope of example embodiments. Applicants have recognized problems existing in several diverse types of nuclear fuel cores with control element placement near certain fuel bundles. Particularly, Applicants have identified that while a maximization of fresh fuel within a nuclear fuel core at any beginning of cycle will permit longer cycle operating times and reduce outage intervals, such maximization can also force fresh fuel bundles to be placed directly next to control elements, which may cause several problems over the life of the fuel, including corrosion, channel-blade interference, and pellet-cladding interactions. Applicants have further recognized that Control Cell Core management techniques, where fresh fuel bundles are not placed directly adjacent to control elements, restricts the amount of fresh fuel that can be placed within a core, as well as limiting placement of fresh fuel in optimal positions for power management, resulting in worsened burnup/efficiency and shorter operating cycles. Example embodiments and methods below uniquely address these and other problems identified by Applicants in related nuclear fuel management technologies for a diverse array of nuclear plants. Example embodiments of the invention include nuclear fuel cores having higher reactivity fuel in lower proportions adjacent to control elements. Lower reactivity fuel is placed in greater proportions adjacent to control elements, while permitting overall fuel content and operating lifespan of the core to be substantially maintained. Example embodiments form two or more different types of positions subject to direct control element exposure—a larger number of controlled positions of a first type having a higher population of burnt and/or lower-enrichment fuel; and a smaller number of controlled positions of a second type having a higher population of fresh and/or higher-enrichment fuel. Specific example embodiments describing how this configuration may be achieved across several different core designs are discussed below, with the understanding that specific placements of the differing types of controlled positions within various regions in example embodiments can be varied based on core design and reactivity needs. It is further understood that any specific plant type, fuel type, enrichment level, exposure level, and/or control element configuration discussed in these example embodiments are not limiting but merely examples of the breadth of nuclear reactor technologies across which example embodiments may be implemented. Example methods of forming and using example embodiments are described thereafter. FIG. 3 is a quadrant map of an example embodiment fuel core 100; for example, FIG. 3 may be an initial loading map for a particular cycle. Core 100 may be useable in existing boiling water reactors; for example, core 100 may be useable in similar plants as the related fuel core loading strategy of FIG. 2. As shown in FIG. 3, core 100 may include a typical BWR fuel core geometry, such as a 17-bundle radius quadrant. Example embodiment core 100 can be visualized in three regions: an outer periphery 120; an inner periphery 130; and an inner core 140. Outer periphery 120 may be up to three fuel bundles thick from an edge of the core in a reactor and include mostly burnt fuel bundles 111 (no fill). Burnt fuel bundles 111 are bundles that have experienced burnup in previous operating cycles or otherwise have been exposed to neutron flux or have significantly lower reactivity than fresh fuel bundles. Inner periphery 130 may be up to three fuel bundles thick and include a larger proportion of higher enrichment fresh fuel bundles 110 (cross-hatched fill). Inner core 140 includes the remainder of the core within inner periphery 130 and includes a mix of lower enrichment fresh fuel bundles 112 (diagonal fill) and burnt fuel bundles 111. Fresh fuel bundles 110 and 112 may have little or no previous neutron flux exposure compared to burnt fuel bundles 111. For example, fresh fuel bundles 110 and 112 may be newly-manufactured bundles previously unused in core operations. Higher enrichment fresh fuel bundles 110 and lower enrichment fresh fuel bundles 112 may differ in fissile material enrichment by any degree required for core 100 operations and optimization. For example, higher enrichment fresh fuel bundles 110 may contain 4.3% Uranium-235 fuel, and lower enrichment fresh fuel bundles 112 may include approximately 4.2% Uranium-235 fuel. Fuel bundles 110 and 112 may each have distinct distributions and concentrations of burnable absorber as well. In other example embodiments, burnt fuel bundles 111, higher enrichment fresh fuel bundles 110, and lower enrichment fresh fuel bundles 112 may be replaced with fuel bundles having a same age but varying initial enrichment and burnable absorber concentration in order to achieve the same reactivity differences as between bundles 110, 111, and 112 in example embodiment core 100. Similarly, reactivity differences may be achieved by using bundles of a same initial enrichment but having three different operating exposure levels, such as fresh, burnt 1-cycle, or burnt 2-cycles in place of higher enrichment fresh fuel bundles 110, lower enrichment fresh fuel bundles 112, and burnt fuel bundles 111. Yet further, reactivity and enrichment differences between all fresh fuel bundles 110 and 112 may be non-existent or minimal, such as where a single fuel type and enrichment is used throughout an entire example core having only differently-aged fuel bundles. Comparing FIGS. 2 and 3, it can be seen that example embodiment core 100 includes more fresh fuel bundles in inner periphery region 130 and does not adhere to a strict checkerboard pattern for fresh and stale fuel bundles in inner core 140. In this way, example embodiment core 100 may include substantially the same amount of fresh fuel bundles 110 and 112 and/or fissile mass and reactivity as related art cores loaded for maximum operation cycle length. Instead of a strict checkerboard alteration between burnt fuel bundles 111 and lower enrichment fresh bundles 112 in inner core 140, example embodiment core 100 includes some groupings of fuel bundles that include more burnt bundles 111. As seen in FIG. 3, four burnt fuel bundles 111 may be grouped about a control blade so as to form a priority control cell 142 that includes less fresh fuel bundles than non-priority control cells 141. For example, as shown in FIG. 3, priority control cells 142, shown by a solid black line surrounding bundle locations so controlled, may include only burnt fuel bundles 111. Priority control cells 142 may be inner-most control cells within inner region 140. Non-priority control cells 141, shown by a broken black line surrounding bundle locations so controlled, may include a mix of burnt fuel bundles 111 and fresh bundles 112 similar to related art core of FIG. 1 and may be positioned closer to or in inner periphery 130, outside of priority control cells 142. FIG. 4 is a quadrant map of an example embodiment fuel core 200; for example, FIG. 4 may be an initial loading map for a particular cycle. Core 200 in the example of FIG. 4 may be useable in an Economic Simplified Boiling Water Reactor (ESBWR). As shown in FIG. 4, core 200 may include a typical ESBWR fuel core geometry, such as a 19-bundle radius quadrant. Example embodiment core 200 can be visualized in three regions: an outer periphery 220; an inner periphery 230; and an inner core 240. Outer periphery 220 may be up to three fuel bundles thick from an edge of the core in a reactor and include mostly once-burnt fuel bundles 213 (dashed fill) and twice-burnt fuel bundles 211 (no fill). Burnt fuel bundles 211 and 213 are bundles that have experienced burnup in previous operating cycles or otherwise have been exposed to neutron flux or have significantly lower reactivity than fresh fuel bundles. For example, once-burnt fuel bundles 213 have approximately 15-23 GWd/ST exposure from a single two-year operating cycle in known ESBWR cores, and twice-burnt fuel bundles 211 may have more burnup, such as 35-40 GWd/ST exposure. Inner periphery 230 may be one to three fuel bundles thick and include a larger proportion of higher enrichment fresh fuel bundles 210 (cross-hatched fill). Inner core 240 includes the remainder of the core within inner periphery 230 and includes a mix of mostly lower enrichment fresh fuel bundles 212 (diagonal fill) and once-burnt fuel bundles 213. Fresh fuel bundles 210 and 212 may have little or no previous neutron flux exposure compared to burnt fuel bundles 211 and 213. For example, fresh fuel bundles 210 and 212 may be newly-manufactured bundles previously unused in core operations. Higher enrichment fresh fuel bundles 210 and lower enrichment fresh fuel bundles 212 may differ in fissile material enrichment by any degree required for core 200 operations and optimization. For example, higher enrichment fresh fuel bundles 210 may contain 4.3% Uranium-235 fuel, and lower enrichment fresh fuel bundles 212 may include approximately 4.2% Uranium-235 fuel. Fuel bundles 210 and 212 may each have distinct distributions and concentrations of burnable absorber as well. In other example embodiments, twice-burnt fuel bundles 211, once-burnt fuel bundles 213, higher enrichment fresh fuel bundles 210, and lower enrichment fresh fuel bundles 212 may be replaced with fuel bundles having a same age but varying initial enrichment and burnable absorber concentration in order to achieve the same reactivity differences as between bundles 210, 211, 212, and 213 in example embodiment core 200. Similarly, reactivity differences may be achieved by using bundles of a same initial enrichment but having three different operating exposure levels, such as fresh, burnt 1-cycle, or burnt 2-cycles in place of higher enrichment fresh fuel bundles 210, lower enrichment fresh fuel bundles 212, and burnt fuel bundles 211 and 213. Yet further, reactivity and enrichment differences between fresh fuel bundles 210 and 212 may be non-existent or minimal, such as where a single fuel type and enrichment is used throughout an entire example core having only differently-aged fuel bundles. Example embodiment core 200 may include substantially the same amount of fresh fuel bundles 210 and 212 and/or fissile mass as related art ESBWR cores loaded for maximum operation cycle length. Example embodiment core 200 includes some groupings of fuel bundles that include more burnt bundles 211 and/or 213. As seen in FIG. 4, four once-burnt fuel bundles 213 may be grouped about a control blade so as to form a priority control cell 242 that includes less fresh fuel bundles and less reactivity than non-priority control cells 241. For example, as shown in FIG. 4, priority control cells 242, shown by a solid black line surrounding bundle locations so controlled, may include only once-burnt fuel bundles 213. Priority control cells 242 may be inner-most control cells within inner region 240. Non-priority control cells 241, shown by a broken black line surrounding bundle locations so controlled, may include a mix of burnt fuel bundles 211 and 213 and fresh bundles 112 and may be positioned closer to or in inner periphery 230. Example embodiment cores are useable with fuel assemblies described in co-owned application Ser. No. 12/843,037 filed Jul. 25, 2010 titled “OPTIMIZED FUEL ASSEMBLY CHANNELS AND METHODS OF CREATING THE SAME,” which is incorporated herein by reference in its entirety. For example, fuel bundles that are to be placed in controlled positions in example embodiment cores may use channels with Zircaloy-4 to additionally guard against shadow corrosion. Other example embodiment cores may be useable in Advanced Boiling Water Reactors, other Light and Heavy Water Reactors, or any nuclear reactor having nuclear chain reaction control structures extending into the core that are useable to control reactivity, with modifications of size and initial enrichments made for the appropriate type of core and control element placement. Example methods include loading and/or operating nuclear cores. Example methods may take particular advantage of nuclear cores loaded as described above in example embodiments, but it is understood that example methods and embodiments may be used separately. During an operating outage or other time when a core is available for loading, an operator or other party may load a core so as to achieve loading patterns consistent with those described in the above example embodiments. For example, existing fuel bundles may be shuffled into stale fuel positions based on their age, enrichment, and/or reactivity. Such shuffling may open a number of positions about an inner periphery and non-primary controlled locations within the inner core. A desired number of oldest or least functional fuel bundles may be removed from the core. Fresh fuel bundles may be procured and installed in locations vacated by the fuel shuffle, based on enrichment or other parameters. Such shuffling may create a fuel core resembling example embodiments described above or related embodiments. During operation of a core, control elements may be used to control the nuclear chain reaction. For example, in related art BWRs, a cruciform control blade may be extended between four adjacent bundles in a control cell to control reactivity. Example methods include using only control elements directly adjacent to fuel bundles having relatively lower reactivity and/or being previously burnt and not fresh for fine, day-to-day reactivity control within a core. In example methods, control elements directly adjacent to fresh or higher reactivity fuel bundles are relatively stationary and used for only coarse reactivity adjustments at a few set points during the fuel cycle; these control elements may be entirely removed from the core—i.e., not used at all for reactivity control—during the later portions of the cycle. As a specific example method in connection with the example embodiment of FIG. 4, an operator or other party may load an ESBWR core 200 so as to create priority control cells 242 including only burnt fuel associated with control blades in a central area of inner core 240 of core 200. Non-priority control cells 241 including some fresh and/or high-reactivity fuel bundles are created at control blade positions nearer or in the inner periphery region 230, outside of the priority control cells 242. During sequence exchange intervals occurring at approximately every 3 GWd/ST of operation, for example, control blades in non-priority cells 241 may be moved to a desired coarse reactivity control position. Otherwise, control blades in non-priority cells 241 are not required to be moved for reactivity control, and control blades only in priority control cells 242 may be moved for fine reactivity control throughout the sequence. During the final quarter of operation, at approximately 15 GWd/ST cycle average exposure, for example, control blades in non-priority cells 241 may be fully withdrawn and not necessary to control reactivity. At all points during the cycle, control blades associated with priority cells 242 may be freely moved to make fine adjustments to core reactivity. During the final quarter of the cycle, control blades in priority cells 242 alone may be used to control core reactivity; that is, control blades in priority cells 242 may be the only blades within core 200 after approximately 15 GWd/ST. Example embodiments and/or methods may provide fuel cores in existing and future-designed reactors with large enough fresh fuel reload batch sizes to accommodate longer operating cycles with higher power densities, while reducing or eliminating concerns associated with placing fresh or higher reactivity fuel directly adjacent to control elements. Placement of fresh fuel in greater numbers about an inner periphery of the core and in limited number of controlled positions may provide a low-leakage core having several inner controlled positions not including fresh or high reactivity fuel. In this way, shadow corrosion, pellet-cladding interaction, and resulting channel distortions and negative control element-channel interaction may be reduced by avoiding placement of the newest and/or highest reactivity fuel bundles closest to active control elements. In addition to longer operating cycle compatibility, high power density, lower leakage, and reduced channel distortion, example embodiments and/or methods may permit nuclear fuel cores to be operated with simplified control element maneuvers; particularly, example embodiments and methods may permit only a subset of control elements to be used for immediate, fine reactivity control and reduce a number of total control element sequences and exchanges throughout an entire operating cycle and/or reduce any need to lower power during such complicated exchanges. These and other advantages and solutions to newly-identified core operating problems are addressed by the various example embodiments and methods described above. Example embodiments and methods thus being described, it will be appreciated by one skilled in the art that example embodiments may be varied and substituted through routine experimentation while still falling within the scope of the following claims. For example, a variety of different nuclear fuel types and core designs are compatible with example embodiments and methods simply through loading and operational strategy—and without any core geometry or structural changes—and fall within the scope of the claims. Such variations are not to be regarded as departure from the scope of these claims. |
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RE0361623 | summary | FIELD OF THE INVENTION The present invention relates generally to radiographic instruments and more particularly to an apparatus for assembling broad area images from narrow beam radiographic scans. BACKGROUND OF THE INVENTION Scanning radiographic equipment differs from conventional radiography in that it employs a narrowly collimated beam of radiation, typically x-rays formed into, for example, a fan beam, rather than a broad area cone beam. The small beam size used in scanning radiographic equipment allows replacement of an image forming sheet of radiographic film, used with conventional radiographic equipment, with a small area array of detector elements. The detector elements receiving the transmitted radiation produce electrical signals which may be converted to digital values by an analog to digital converter for the later development of an image or for other processing by computer equipment. The ability to quantify the measurement of the transmitted radiation, implicit in the digitization by the analog to digital converter, allows not only the formation of a radiographic "attenuation" image but also the mathematical analysis of the composition of the attenuating material by dual energy techniques. See generally, "Generalized Image Combinations in Dual KVP Digital Radiography", by Lehmann et al. Med. Phys. 8(5) September/October 1981. Such dual energy techniques quantitatively compare the attenuation of radiation at two energies to distinguish, for example, between bone and soft tissue. Dual energy techniques allow the measurement of bone mass, such measurement being important in the treatment of osteoporosis and other bone diseases. The limited area of the beam of radiation used in scanning radiographic systems allows the use of limited area detectors permitting high resolution with relatively lower cost. The limited area of the detectors, however requires that the beam be scanned along several adjacent path if large area images are to be constructed. Typically, a fan beam will be scanned in a raster pattern over the area to be measured, each line of the scan separated by somewhat less than the width of fan beam to ensure complete illumination of the entire volume of the imaged object, with the directions of scanning being generally perpendicular to the direction of the radiation and the plane of the fan beam. Images formed by a scanning radiographic system are potentially more accurate than those produced by a typical broad beam radiograph system. This accuracy arises from the limited divergence, in the scanning direction, of the rays of the fan beam, as compared to a broad area cone beam. This narrow collimation of the fan beam reduces "parallax" in the projected image, particularly of anatomical planar surfaces that are nearly parallel with the plane of the tan beam--such as the superior and interior borders of the vertebrae in the spine when the scanning ..directions.!. .Iadd.direction .Iaddend.is along the superior-inferior axis of the body. Morphological measurements of the vertebrae, and other structures, which benefit from reduced parallax are used to evaluate various dimensions of a vertebra to detect crushing or other deformation that are one element of certain bone diseases such as osteoporosis. See e.g. Minne et al., "A Newly Developed Spine Deformity Index (SDI) to Quantitate Vertebral Crush Factors in Patients with Osteoporosis," Bone and Mineral, 3:335-349 (1988); J. C. Gallagher et al, "Vertebral Morphometry: Normative Data," Bone and Mineral, 4:189-196 (1988); Hedlund et al, "Vertebral Morphometry in Diagnosis of Spinal Fractures," Bone and Mineral, 5:59-67 (1988); and Hedlund et al, "Change in Vertebral Shape in Spinal Osteoporosis," Calcified Tissue International, 44:168-172 (1989). Automatic techniques for morphological measurements of bone are described in U.S. patent application Ser. No. 07/944,626 filed Sep. 14, 1992 and entitled: "Method for Analyzing Vertebral Morphology Using Digital Radiography" assigned to the same assignee as the present application and hereby incorporated by reference. Nevertheless images developed with scanning fan beam equipment can include certain distortions or artifacts. In particular, it has been noted that objects at the interface between two adjacent scan paths contain a blurring or distortion in a direction perpendicular to the scan path. SUMMARY OF THE INVENTION The present invention provides a method and apparatus for constructing broad area images from a sequence of narrow fan beam scans. The invention recognizes that a source of image artifacts in combining narrow, fan beam scans is the varying amount of overlap between the fan beams when the axes of the fan beams are held parallel. This overlap causes some volume elements of the patient to be measured with rays at two different angles. The amount of overlap depends on the height of the structure being imaged, as measured along the path of the fan beams, and thus cannot, in general, be determined or corrected in a two dimensional image. The present invention vanes the angle of the axis of each fan beam so as to create a larger, effective fan beam of arbitrary width and to eliminate any height dependent overlap. The elimination of height dependent overlap ensures that each volume element of the patient is measured by rays at only one angle. Specifically, the invention employs an imaging system having a radiation source directing a fan beam of radiation toward the patient, where the fan beam diverges about a radiation axis, substantially within a beam plane, from a focal spot. A radiation detector opposing the radiation source along the radiation axis receives the diverging beam of radiation after passage through the patient to produce a projection signal indicating the attenuation of the beam of radiation for multiple rays within the beam. The radiation axis may be moved along a first and second path across the patient, the first and second paths being spaced apart and substantially perpendicular to the beam plane. In moving between the first and second paths of the scan, the radiation axis is rotated about the focal spot by a displacement angle, within the beam plane. The signals obtained along the first and second path are then combined to produce a two dimensional projection image. It is thus one object of the invention to reduce image artifacts, caused by combining image data obtained from multiple scannings of a narrow fan beam. Creating a larger, effective fan beam eliminates areas of overlap or produces areas of overlap that, with appropriate projections, are constant regardless of the height of the imaged structures and which therefore can be eliminated by a constant weighting factor applied to the data of the overlapping area. The radiation detector may be a linear array of detector elements, each subtending a first width of the fan beam along the linear array, where the ..projections.!. .Iadd.projection .Iaddend.signals include a plurality of elements signals from each detector element. A projector may be employed to map the element signals to pixels of a non-planar image surface generally normal to the radiation axis, each pixel subtending second widths of the fan beam varying from the first widths. The non-planar image surface may be positioned midway along the height of the patient as measured along the radiation access. It is thus another object of the invention to reduce the distortion caused by the divergence of rays in both the marrow measuring fan beams and the larger, effective fan beam by mapping the element signals to pixels of a non-planar surface so that each such pixel represents rays of the fan beam passing through equal areas of the patient. This reduces variations, for example, in bone mineral density measurements, which are sensitive to distortion in the measured area. It is another object of the invention to reduce the magnitude of ..magnifications.!. .Iadd.magnification .Iaddend.induced errors on the projected image. By positioning the non planar image surface to approximately bisect the body, distance between the imaging plane and any particular structure in the body, such as affects magnifications, is reduced to a minimum. The foregoing and other objects and advantages of the invention will appear from the following description. In the description, reference is made to the accompanying drawings which form a pan hereof and in which there is shown by way of illustration, a preferred embodiment of the invention. Such embodiment does not necessarily represent the full scope of the invention, however, and reference must be made therefore to the claims herein for interpreting the scope of the invention. |
claims | 1. A method for mitigating a stress corrosion cracking of structural material in a nuclear power plant, comprising the steps of:injecting hydrogen into a reactor water of a boiling water nuclear power plant,wherein in said step of injecting said hydrogen into the reactor water, said hydrogen is injected until a main steam line dose rate comes to a prescribed value, andwherein in said step of injecting said hydrogen into the reactor water, the amount of injected hydrogen is decreased when the main steam line dose rate exceeds the prescribed dose rate;injecting a reductive nitrogen compound containing nitrogen having a negative oxidation number into said reactor water of a boiling water nuclear power plant after injection of said hydrogen,wherein in said step of injecting said reductive nitrogen compound into the reactor water, said reductive nitrogen compound is injected after injection of said hydrogen; andcontrolling amount of injection of said reductive nitrogen compound by using an oxygen concentration and an ammonia concentration in said reactor water as indication,wherein the injection of said reductive nitrogen compound is controlled to cause said oxygen concentration to be higher than a lower limit of a target value, and to cause said ammonia concentration to be lower than an upper limit of a target value,wherein the lower limit of the target value of the oxygen concentration is 10 ppb, andwherein the upper limit of the target value of the ammonia concentration is 4.2×10−6 mol/l. |
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claims | 1. An apparatus for driving a plate-like driven mass, comprising:a drive unit;a wobble bearing to convert a rotary motion of the drive unit into a linear motion of the driven mass; anda plate-like compensatory mass disposed parallel to the plate-like driven mass, the compensatory mass mechanically coupled with the wobble bearing so as to move linearly contrary to a linear motion of the plate-like driven mass during operation of the drive unit. 2. The apparatus of claim 1, wherein the plate-like driven mass is a scattered radiation grid of a diagnostic X-ray system. 3. The apparatus of claim 1, wherein the compensatory mass is connected to the wobble bearing via a thrust rod and at least one decoupling element which converts the tumbling motion exerted by the wobble bearing into the contrary linear motion. 4. The apparatus of claim 3, wherein the thrust rod is connected by a second end to a guide carriage secured to the compensatory mass. 5. The apparatus of claim 4, further comprising a bearing element, supported on one side in the guide carriage and connected on the other side to the thrust rod for permitting a linear compensatory motion. 6. The apparatus of claim 5, wherein the linear compensatory motion is perpendicular to the linear motion of the compensatory mass. 7. The apparatus of claim 5, wherein the thrust rod is connected by the second end to the guide carriage via a second joint head. 8. The apparatus of claim 7, wherein the trust rod is supported in the bearing element. 9. The apparatus of claim 4, further comprising a bearing housing in which the thrust rod is supported in sliding fashion and which is secured to a frame component or housing component. 10. The apparatus of claim 3, further comprising a bearing housing in which the thrust rod is supported in sliding fashion and which is secured to a frame component or housing component. 11. The apparatus of claim 3 further comprising a retaining element secured to the wobble bearing, and having a connection part to receive a first end of the thrust rod. 12. The apparatus of claim 3, wherein the wobble bearing is connected to the driven mass via a coupling element, and the thrust rod extends along a side of the coupling element. 13. The apparatus of claim 12, wherein the coupling element is a peg. 14. The apparatus of claim 12, wherein a first end of the thrust rod is connected to the wobble bearing in a diametrically opposite configuration to the coupling element. 15. The apparatus of claim 14, further comprising a retaining element secured to the wobble bearing, and having a connection part to receive a first end of the thrust rod. 16. The apparatus of claim 15, wherein the retaining element is clamped to the wobble bearing. 17. The apparatus of claim 15, wherein the thrust rod is supported in the connection part via a first joint head. 18. The apparatus of claim 17, wherein the first joint head is one decoupling element of a plurality of decoupling elements configured for converting a tumbling motion into the linear motion. 19. A diagnostic radiology system having an apparatus for driving a scattered radiation grid, the apparatus comprising:a drive unit;a wobble bearing to convert and transmit a rotary motion of the drive unit into a linear motion of the scattered radiation grid, anda compensatory mass disposed generally in parallel to the scattered radiation grid, the compensatory mass being mechanically connected with the wobble bearing so as to execute during operation of the drive unit a linear motion that is contrary to a linear motion of the scattered radiation grid. 20. The diagnostic radiology system of claim 19, wherein the compensatory mass is connected to the wobble bearing via a thrust rod and at least one decoupling element which converts the tumbling motion exerted by the wobble bearing into the contrary linear motion. |
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description | Priority is claimed as a continuation application to U.S. patent application Ser. No. 12/260,914, filed Oct. 29, 2008 and now issued as U.S. Pat. No. 8,576,976, which claims priority to U.S. Provisional Patent Application Ser. No. 60/983,566, filed Oct. 29, 2007 and to U.S. Provisional Patent Application Ser. No. 61/038,525, filed Mar. 21, 2008. The aforementioned priority applications are incorporated herein by reference in their entireties. The present invention relates generally to apparatus and methods for supporting high level radioactive waste, and specifically to apparatus and methods for supporting radioactive fuel assemblies. The subject inventions can be implemented into a wide variety of structures for transferring, supporting and/or storing spent nuclear fuel assemblies, including without limitation underwater fuel racks and fuel baskets incorporated into either canisters or casks. In the nuclear power industry, the nuclear energy source is in the form of hollow zircaloy tubes filled with enriched uranium, known as fuel assemblies. Upon being deleted to a certain level, spent fuel assemblies are removed from a reactor. At this time, the fuel assemblies not only emit extremely dangerous levels of neutrons and gamma photons (i.e., neutron and gamma radiation) but also produce considerable amounts of heat that must be dissipated. It is necessary that the neutron and gamma radiation emitted from the spent fuel assemblies be adequately contained at all times upon being removed from the reactor. It is also necessary that the spent fuel assemblies be cooled. Because water is an excellent radiation absorber, spent fuel assemblies are typically submerged under water in a pool promptly after being removed from the reactor. The pool water also serves to cool the spent fuel assemblies by drawing the heat load away from the fuel assemblies. The water may also contain a dissolved neutron shielding substance. The submerged fuel assemblies are typically supported in the fuel pools in a generally upright orientation in rack structures, commonly referred to as fuel racks. It is well known that neutronic interaction between fuel assemblies increases when the distance between the fuel assemblies is reduced. Thus, in order to avoid criticality (or the danger thereof) that can result from the mutual inter-reaction of adjacent fuel assemblies in the racks, it is necessary that the fuel racks support the fuel assemblies in a spaced manner that allows sufficient neutron absorbing material to exist between adjacent fuel assemblies. The neutron absorbing material can be the pool water, a structure containing a neutron absorbing material, or combinations thereof. Fuel racks for high density storage of fuel assemblies are commonly of cellular construction with neutron absorbing plate structures (i.e., shields) placed between the cells in the form of solid sheets. The cells are usually long vertical square tubes which are open at the top through which the fuel elements are inserted. The cells are sometimes with double walls that encapsulate the neutron shield sheets to protect the neutron shield from corrosion or other deterioration resulting from contact with water. Each fuel assembly is placed in a separate cell so that the fuel assemblies are shielded from one another. An example of a typical existing fuel rack, is described in U.S. Pat. No. 4,382,060, to Maurice Holtz et al., issued May 3, 1983, the entirety of which is hereby incorporated by reference. The Holtz rack is comprised of structural elements including elements which are hollow and cruciform in section. Each leg of the cruciform structural element includes a neutron shield therein. The free end of the legs of the cruciform structural element converge so as to have an included angle of approximately 90 degrees. The rack is comprised of such cruciform elements as well as cooperating elements which are generally T and L shaped in section. In certain regions of the world, the fuel assemblies used in the nuclear reactors do not have a rectangular horizontal cross-section. Instead, the fuel assemblies have a horizontal cross-section that is generally hexagonal. In such instances, existing racks having cells with rectangular horizontal cross-sections are less than optimal. Even after removal from the pool, the fuel assemblies still emit extremely dangerous neutrons (i.e., neutron radiation) and gamma photons (i.e., gamma radiation) and it is thus still imperative that these neutrons and gamma photons be contained at all times during transfer and storage. It also imperative that the residual heat emanating from the fuel assemblies be lead away and escape from the fuel assemblies. Thus, containers used to transfer and/or store fuel assemblies must not only safely enclose and absorb the radioactivity of the fuel assemblies, they must also allow for adequate cooling. In the art, there are two type of container systems used to transport and/or store fuel assemblies, canister-based systems and cask-based systems. Generally speaking, there are two types of casks used for the transportation and/or storage of SNF, ventilated vertical overpacks (“VVOs”) and thermally conductive casks. VVOs typically utilized in conjunction with a sealable canister that is loaded with the fuel assemblies and positioned within a cavity of the VVO. Such canisters, which are often multi-purpose canisters, often contain a fuel basket for receiving the fuel assemblies. An example of a canister and basket assembly designed for use with a VVO is disclosed in U.S. Pat. No. 5,898,747 (Singh), issued Apr. 27, 1999, the entirety of which is hereby incorporated by reference. The second type of casks are thermally conductive casks. In a typical thermally conductive cask, the fuel assemblies are loaded directly into a cavity formed by the cask body. A basket assembly is typically provided within the cavity itself to provide support for the fuel assemblies. The fuel basket generally acts in conjunction with the cask to support the fuel in a particular pattern, minimize load transfer to the fuel, transfer heat to the cask and control criticality. It is an object of the present invention to provide a fuel rack that can safely accommodate fuel assemblies. Yet another object of the present invention is to provide a fuel rack having a horizontal cross-section that is non-rectangular in shape, such as hexagonal. Still another object of the present invention is to provide a fuel rack that is cost effective to manufacture. A further object of the present invention is to provide a fuel rack that can withstand high inertia loads acting in concert with hydraulic loads from moving water. A yet further object of the present invention is to provide a fuel rack that allows natural thermosiphon flow of the pool water through the cells. A still further object of the present invention is to provide a fuel rack that eliminates the need for neutron absorber plates. It is a further object of the present invention to provide a fuel rack constructed of slotted plates. Another object of the present invention is to provide a fuel rack that is compact and maximizes the storage space of a fuel pool. A yet further object of the present invention is to provide a fuel rack that resists water corrosion. Still another object of the present invention is to provide a fuel rack that maintains structural stability under radiation exposure. Another object of the present invention is to provide a fuel rack that provides flux traps. It is an object of the present invention to provide a fuel basket that provides higher structural integrity. It is a further object of the present invention to provide a fuel basket that has fuel cells which correspond to the shape of the fuel assembly to be stored therein. A yet further object of the present invention is to provide a fuel basket that maximizes the packing density of the spent nuclear fuel while maintaining a reactivity of 0.95 or less. A yet further object of the present invention is to provide a fuel basket that is easy to manufacture and lightweight. These and other objects are met by the present invention which, in one aspect can be a fuel rack having an array of cells for holding fuel assemblies comprising: a base plate having a top surface; a plurality of tubes, each tube having an inner surface that forms one of the cells; and the tubes connected to the top surface of the base plate in a substantially vertical orientation and in a pattern so that one or more of the cells are formed by outside surfaces of the adjacent tubes. In another aspect, the invention can be a fuel rack for supporting fuel assemblies comprising: a plurality of hexagonal tubes having an internal cavity; a base plate having a top surface; the hexagonal tubes connected to the top surface of the base plate in a substantially vertical orientation and spaced from one another so that a flux trap space exists between all adjacent hexagonal tubes; and a plurality of spacers positioned in the flux trap spaces for maintaining the existence of the flux trap spaces, the spacers connected to the hexagonal tubes. In yet another aspect, the invention can be a fuel rack having an array of cells for holding fuel assemblies comprising a plurality of slotted plates that are slidably interlocked with one another to form the array of cells. In another aspect, the invention can be a fuel basket having a honeycomb-like grid that forms a plurality of substantially vertically oriented elongated cells. Most preferably, the basket assembly comprises one or more flux traps and is positioned within the cavity. The basket assembly can be constructed of a metal matrix composite material. In one embodiment, the basket assembly may utilize variable flux traps to maximize packing density. In such an embodiment, as the periphery of the basket assembly is approached, the width of the flux traps may decrease. In another embodiment, the basket assembly may utilize tubular elements of varying heights in a vertically staggered formation so that no two adjacent cells have interfaces which are vertically aligned. In a further aspect, the invention can be an apparatus for supporting radioactive fuel assemblies comprising: a grid of cells for housing radioactive fuel assemblies, the grid formed by a plurality of hexagonal tubes having an outer surface and an inner surface that forms one of the cells, the plurality of hexagonal tubes arranged in an adjacent manner and in a pattern so that one or more of the cells is a resultant cell formed by the outside surfaces of surrounding hexagonal tubes. In a yet further aspect, the invention can be an apparatus for supporting radioactive fuel assemblies comprising: a grid of cells for housing radioactive fuel assemblies, the grid formed by a plurality of tubes having inner surfaces that form the cells, the tubes arranged in an axially aligned and adjacent manner; each of the tubes formed by a plurality of tubular segments stacked in axial alignment, an interface formed between the adjacent tubular segments of each tube; and wherein the lengths of the tubular segments and the pattern in which the tubes are arranged to form the grid is such that none of the interfaces of adjacent tubes are aligned with one another. In a still further aspect, the invention can be an apparatus for supporting radioactive fuel assemblies comprising: a bottom section comprising a plurality of bottom tubular segments of varying length, the bottom tubular segments arranged in an axially adjacent manner and in a pattern so that no two adjacent bottom tubular segments are the same length, the bottom edges of the bottom tubular segments being aligned; at least one middle section comprising a plurality of middle tubular segments of equal length, the middle section stacked atop the bottom section so that the middle tubular segments are axially aligned with the bottom tubular segments and the bottom edges of the middle tubular segments abut the top edges of the bottom tubular segments; and a top section comprising a plurality of top tubular segments of varying length, the top section stacked atop the middle section so that the top tubular segments are axially aligned with the middle tubular segments, the bottom edges of the top tubular segments abut the top edges of the bottom tubular segments, and the top edges of the top tubular segments are aligned. In even another aspect, the invention can be a fuel rack having a grid of cells for holding fuel assemblies comprising: a base plate having a top surface; a plurality of hexagonal tubes, each hexagonal tube having inner surfaces that forms one of the cells; and the hexagonal tubes connected to the top surface of the base plate in a substantially vertical orientation and in a pattern so that one or more of the cells are formed by outside surfaces of the adjacent hexagonal tubes. In yet another aspect, the invention can be a fuel rack having a grid of cells for holding fuel assemblies comprising: a base plate having a top surface; a plurality of tubes, each tube having an inner surface that forms one of the cells; and the tubes connected to the top surface of the base plate in a substantially vertical orientation and in a pattern so that one or more of the cells are formed by outside surfaces of the adjacent tubes. In still another aspect, the invention can be a fuel rack for supporting fuel assemblies comprising: a plurality of hexagonal tubes having an internal cavity; a base plate having a top surface; the hexagonal tubes connected to the top surface of the base plate in a substantially vertical orientation and spaced from one another so that a flux trap space exists between all adjacent hexagonal tubes; and a plurality of spacers positioned in the flux trap spaces for maintaining the existence of the flux trap spaces, the spacers connected to the hexagonal tubes. In another aspect, the invention can be a fuel rack for supporting fuel assemblies comprising: a plurality of tubes having an internal cavity; a base plate having a top surface; the tubes connected to the top surface of the base plate in a substantially vertical orientation and spaced from one another so that a flux trap space exists between all adjacent tubes; and a plurality of spacers positioned in the flux trap spaces for maintaining the existence of the flux trap spaces, the spacers connected to the tubes. In a still further aspect, the invention can be a fuel rack having perimeter cells and non-perimeter cells for supporting fuel assemblies comprising: a base plate having a top surface; a plurality of hexagonal tubes, each hexagonal tube having inner surfaces that form one of the perimeter cells or the non-perimeter cells; and the hexagonal tubes connected to the top surface of the base plate in a substantially vertical orientation and in a pattern so that every third non-perimeter cell is formed by outside surfaces of six surrounding hexagonal tubes. In yet another aspect, the invention can be a fuel basket for supporting radioactive fuel assemblies comprising: a plurality of tubes having an internal cavity for receiving a radioactive fuel assembly; the tubes arranged in a substantially vertical orientation and spaced from one another so that a flux trap space exists between all adjacent tubes, the tubes forming a storage grid having a central axis and a perimeter; a plurality of spacers positioned in the flux trap spaces for maintaining the existence of the flux trap spaces; and wherein the width of the flux trap space between adjacent tubes decreases with distance from the central axis of the storage grid. The present invention will now be described in relation to exemplary embodiments. It is to be understood that while certain details and structural arrangements are explained in detail with respect to a certain embodiment, the details and structural arrangements can be implemented into any of the embodiments. I. Flux Trap Fuel Rack Embodiment Referring to FIG. 1, a perspective view of a fuel rack 100 according to one embodiment of the present invention is disclosed. The fuel rack 100 is a cellular, upright, prismatic module. The illustrated embodiment of the fuel rack 100 is specifically designed to accommodate hexagonal fuel assemblies, such as VVER 1000 fuel assemblies. To this extent, each cell 101 of the fuel rack 100 is also generally hexagonal in shape (i.e., have a hexagonal horizontal cross-section) so as to geometrically accommodate a single hexagonal fuel assembly. However, it is to be understood that the concepts of the present invention can be modified to accommodate any shaped fuel assembly, including rectangular, octagonal, round, etc. In describing the fuel rack 100 and its component parts below, relative terms such as top, bottom, above, below, horizontal, vertical upper and lower will be used in relation to the fuel rack 100 being in the illustrated substantially vertical orientation of FIG. 1. Additionally, in order to avoid clutter in the drawings, only a few of each component are numbered with the understanding that the reader will be able to identify duplicate elements. The fuel rack 100 generally comprises a base plate 110, a plurality of hexagonal tubes 120, and a plurality of spacing rods 130 (best visible in FIG. 2). The hexagonal tubes 120 are connected to the top surface 111 of the base plate 110 in a substantially vertical orientation. In this embodiment, the axis of each hexagonal tube 120 is not only substantially vertical but also substantially perpendicular to the top surface 111 of the base plate 110. The connection between the hexagonal tubes 120 and the base plate 110 is achieved by welding the bottom edge of the hexagonal tubes 120 to the top surface of the base plate 110. Of course, other connection techniques can be utilized with minor modification, including mechanical connections such as bolting, clamping, threading, etc. The top ends of the hexagonal tubes 120 remain open so that a fuel assembly can be slid into the internal cavity 101 (also referred to as a cell) formed by the inner surfaces of the hexagonal tubes 120. Each hexagonal tube 120 can be a single-part tube that extends the entire desired height H1 or can be constructed of multiple partial height tubes that together add up to the desired height H1. It is preferred that the height H1 be sufficient so that the entire height of the fuel assembly is within the hexagonal tube 120 The hexagonal tubes 120 are connected to the rectangular base plate 110 in an adjacent and spaced pattern to form a honeycomb-like grid of the cells 101. The cells 101 are substantially vertical elongated cavities for receiving the radioactive fuel assemblies via their open top ends. While a generally rectangular gridwork of cells 101 is illustrated, the fuel rack 100 can be designed to take on any desired shape. The geometric arrangement of the hexagonal tubes 120 will be discussed in greater detail below with respect to FIGS. 2-3. The hexagonal tubes 120 preferably constructed of a metal-matrix composite material, and more preferably a discontinuously reinforced aluminum/boron carbide metal matrix composite material, and most preferably a boron impregnated aluminum. One such suitable material is sold under the tradename Metamic™. The hexagonal tubes 120 perform the dual function of reactivity control as well as structural support. The base plate 110 is preferably constructed of a metal that is metallurgically compatible with the material of which the hexagonal tubes 120 are constructed for welding. Referring now to FIGS. 2-3 concurrently, each hexagonal tube 120 is arranged so as to be spaced from all adjacent hexagonal tubes 120 so that a gap 140 exists between each hexagonal tube 120 and its immediately adjacent hexagonal tubes 120. The gap 140 acts a neutron flux trap that decreases and/or eliminates the danger of criticality. The flux trap space 140 can be designed to be any desired width and the exact width will depend on the radiation levels of the fuel assemblies to be stored, the material of construction of the tubes 120, and properties of the pool water in which the fuel rack 100 will be submerged. In one embodiment, the flux trap spaces 140 will have a width between 30 and 50 millimeters and more preferably between 25 to 35 millimeters and most preferably about 38 millimeters. Spacers, which are in the form of spacing rods 130 in the illustrated embodiment, are inserted into the flux trap spaces 140 so as to maintain the existence of the flux trap spaces 140 at the desired width and to provide added structural stability. While the spacers are illustrated as elongated rods 130 that extend the entire height H1 of the hexagonal tubes 120, the spacers are not so limited and can take on a wide variety of shapes and sizes. For example, the spacers could be merely blocks or pins if desired in some embodiments. A spacing rod 130 is positioned at the juncture between the edges of three adjacent hexagonal tubes 120. Thus, each spacing rod 130 (with the exception of those spacing rods 130 along the perimeter) contacts three hexagonal tubes 120. For added integrity and ease of construction, the spacing rods 130 have three axial grooves along their length that act as nesting volumes for receiving the edge of the hexagonal tubes 120. In the illustrated embodiment, the spacing rods 130 have a horizontal cross-section that is generally in the shape of a truncated triangle, wherein a nesting groove is formed into each truncated apex. Of course, the spacing rods 130 can take on other shapes with or without the grooves. The spacing rods 130 are preferably made of aluminum or a metal matrix material, such as boron impregnated aluminum. The spacing rods 130 are plug welded to the hexagonal tubes 120 in which they are in contact with via elongated holes 121 located at the edges/corners or the hexagonal tubes 120. The shape, location and number of plug weld holes 120 will vary depending on design considerations and is in no way limiting of the present invention. The plug holes 121 are uniformly on each corner the hexagonal tubes 120 to facilitate uniform manufacture but this is not necessary. The plug holes 121 can be formed by punching, cutting, or during a molding process. Referring now to FIGS. 3-4 concurrently, the base plate 110 also comprises a plurality of flow holes 115 extending through the base plate 110 from its bottom surface 112 to its top surface 111. Similarly, the base plate 110 also comprises four oblong holes 116 (second row in from the corners) for lifting and installing the fuel rack 100 within the fuel pool. A special lifting beam with four long reach rods is used to interact with the oblong holes 116 to grapple the fuel rack 100 and place it in the pool. The flow holes 115 (and oblong holes 116) create passageways from below the base plate 100 into the cells 101 formed by the hexagonal tubes 120. Preferably, a single flow hole 115 is provided for each cell 101. The flow holes 115 are provided as inlets to facilitate natural thermosiphon flow of pool water through the cells 101 when fuel assemblies having a heat load are positioned therein. More specifically, when heated fuel assemblies are positioned in the cells 101 in a submerged environment, the water within the cells 101 surrounding the fuel assemblies becomes heated, thereby rising due to increased buoyancy. As this heated water rises and exist the cells 101 via their open top ends, cool water is drawn into the bottom of the cells 101 via the flow holes 115. This heat induced water flow along the fuel assemblies then continues naturally. As can best be seen in FIG. 5, a plurality of auxiliary cutouts/holes 121 are provided in the hexagonal tubes 120 at or near their bottom edge. The auxiliary holes 121 act as additional inlet openings for incoming pool water to facilitate the thermosiphon flow during the cooling process. Moreover, as will be described below, the flow holes 115 of certain cells 101 are blocked by the attachment of adjustable height pedestals 150. The auxiliary holes 121 of the hexagonal tubes 120 that form these cells 101 are thus the sole source of incoming cool water for fuel assemblies stored therein. While an auxiliary hole 121 is provide din each face of each and every hexagonal tube 120 in the fuel rack 100, it is to be understood that this may not be necessary in all instances. As a side note, the flow holes 115 (and holes 116) perform an additional function of providing an access-way into to the cells 101 for a “goose-neck welder” for welding the hexagonal tubes 120 to the top surface of the base plate 110. Referring back to FIGS. 3-4, the base plate 110 also comprises a plurality of adjustable height pedestals 150 connected to the bottom surface 112 of the base plate 110. The adjustable height pedestals 150 ensure that a space exists between the floor of the fuel pool and the bottom surface 112 of the base plate 110, thereby creating an inlet plenum for water to flow through the flow holes 115. The adjustable height pedestals 150 are spaced to provide uniform support of the base plate 110 and thus the fuel rack 100. Each pedestal 150 is individually adjustable to level and support the rack on a non-uniform spent fuel pool floor. The pedestals 150 are bolted to the base plate 110. Of course, if desired, the pedestals 150 can be attached top the base plate 110 by other means, including welding or threading. In the event of welded pedestal 15, an explosion-bonded stainless-Aluminum plate may be used to make the transition. For a welded pedestal, the bolts and bolt holes are eliminated. Referring now to FIGS. 6A-6B, the construction details of the adjustable height pedestals 150 will be described. Each of the adjustable height pedestals 150 comprises a block 151 and a cylindrical peg 152 that acts as the foot. The block 152 is connected to the base plate 110 via bolts 155. The block 151 comprises a central hole 153 which has a threaded inner surface (not visible). Similarly, the outer surface of a portion of the peg 151 is also threaded with corresponding threads. The peg 152 is inserted into the hole 153 and threadably engaged therein to the block 151. The peg 152 also comprises a rectangular depression 154 in its top surface for receiving a tool for turning the peg 152. Of course, the depression can be any shape that will facilitate rotational engagement with a tool. Moreover, other means for engaging and turning the peg 152 can be sued including a tab, a screw head, a bolt head, etc. Because of the threaded connection between the peg 152 and the block 151, turning the peg 152 via the depression 154 results in increasing or decreasing the height the peg 152 protrudes from the bottom surface of the block 151. Adjustment of the peg 152 is facilitated by a long-handled tool that is inserted into the cell 101. The depression 154 of the peg 152 is accessible through the flow hole 115 in that cell 101 (see FIG. 3). The bottom portion of the peg 152 has a rounded edge to prevent catching and tearing of the liner in a seismic-induced slide of the fuel rack 100. A break in the liner means problems for the site because of leakage. If desired, the bottom surface of the peg 152 can be formed or covered with a low friction sliding material. II. Non-Flux Trap Fuel Rack Embodiment Incorporating Resultant Cells Referring now to FIGS. 7-10 concurrently, a second embodiment of a fuel rack 200 is illustrated. Similar to the rack above, the fuel rack 200 is a cellular, upright, prismatic module. The illustrated embodiment of the fuel rack 200 is specifically designed to accommodate hexagonal fuel assemblies, such as VVER 1000 fuel assemblies. Each cell 201 of the fuel rack 200 is also generally hexagonal in shape (i.e., have a hexagonal horizontal cross-section) so as to geometrically accommodate a single hexagonal fuel assembly. However, it is to be understood that the concepts of the present invention can be modified to accommodate any shaped fuel assembly, including rectangular, octagonal, round, etc. In describing the fuel rack 200 and its component parts below, relative terms such as top, bottom, above, below, horizontal, vertical upper and lower will be used in relation to the fuel rack 200 being in the illustrated substantially vertical orientation of FIG. 7. Additionally, in order to avoid clutter in the drawings, only a few of each component are numbered with the understanding that the reader will be able to identify duplicate elements. The driving factor that leads to the structural differences between fuel rack 100 (discussed above) and the fuel rack 200 is that the fuel rack 200 is designed to be used with fuel assemblies that do not require the presence of a neutron flux trap between adjacent cells 201. Thus, the inclusion of neutron flux traps in fuel racks when not needed is undesirable because valuable pool floor area is unnecessarily wasted. Of course, both fuel rack types 100, 200 may be stored side by side in the same pool. Because many of the structural and functional features of the fuel rack 200 are identical to the fuel rack 100, only those aspect of the fuel rack 200 that are different will be discussed below with the understanding that the other concepts discussed above with respect to fuel rack 100 are applicable. The fuel rack 200 comprises a plurality of hexagonal tubes 220. The hexagonal tubes 220 are connected to the top surface 211 of the base plate 210 in a substantially vertical orientation. In this embodiment, the axis of each hexagonal tube 220 is not only substantially vertical but also substantially perpendicular to the top surface 211 of the base plate 210. The connection between the hexagonal tubes 220 and the base plate 210 is achieved by welding the bottom edge of the hexagonal tubes 220 to the top surface of the base plate 110. Of course, other connection techniques can be utilized with minor modification, including mechanical connections such as bolting, clamping, threading, etc. The top ends of the hexagonal tubes 220 remain open so that a fuel assembly can be slid into the internal cavity 201A (also referred to as a cell) formed by the inner surfaces each hexagonal tube 220. The hexagonal tubes 220 are connected atop the rectangular base plate 110 in a special geometric arrangement so that certain non-perimeter cells 201D are formed by the outside surfaces of the surrounding hexagonal tubes 220. Additionally, certain perimeter cells 201B-C are formed by the outside surfaces of the surrounding hexagonal tubes 220 and an added plate, which depending on the location is either a two-panel plate 225 or a single panel plate 226. In other words, the cells 201B-201D are not the internal cavities of any tubular structures but are resultant cavities formed by either (1) the outer surfaces of the surrounding hexagonal tubes 220; or (2) the outer surfaces of the surrounding hexagonal tubes 220 and an additional plate structure 225, 226. As used herein, all three cell types 201B-201D will be referred to as “resultant cells” or “developed cells.” Despite their different methods of formation, all of the cells 201A-201D have a horizontal cross-section that is generally hexagonal. Of course, the result cell concept can be applied to a host of other geometries if desired. The special geometric arrangement of the hexagonal tubes 220 and an additional plate structure 225-226 atop the base plate 210 will now be discussed. As can best be seen in FIG. 9, the hexagonal tubes 220 are geometrically arranged atop the base plate 210 in rows 1-11 (indicated numerically in a circle). Of course, any number of rows or columns may be created for the fuel rack 200. The details of the formation of three different kinds of resultant cells will be described in reference to rows 1 through 3 with the understanding that the certain patterns repeat and thus the entire fuel rack 200 can be constructed. A. Formation of Perimeter Resultant Cells X The formation of perimeter resultant cells formed by the combination of outside surfaces of the hexagonal tubes and a two-panel plate structure (referred to above as type 201C cells) will now be described. For ease of reference and to avoid clutter, all perimeter resultant cells formed by the combination of outside surfaces of the hexagonal tubes and a two-panel plate structure are marked with an X. In row 1, the hexagonal tubes 220(1), 220(2) are arranged atop the base plate 210 in the top left corner in an adjacent and abutting manner so that the opposing outside surfaces of the hexagonal tubes 220(1), 220(2) are in surface contact. The internal cavities of the hexagonal tubes 220(1), 220(2) act as the first two cells 201A (not marked). To reduce further clutter, all cells 201A that are formed by the inner surfaces of a single hexagonal tube 220 are left blank in FIG. 9. A second pair hexagonal tubes 220(3), 220(4) are arranged atop the base plate 210 within row 1 and spaced from the first pair of hexagonal tubes 220(1), 220(2). Similar to the first pair of hexagonal tubes 220(1), 220(2), the second pair of hexagonal tubes 220(3), 220(4) are in arranged in an adjacent and abutting manner so that the opposing outside surfaces of the hexagonal tubes 220(3), 220(4) are in surface contact. While the second pair of hexagonal tubes 220(3), 220(4) are aligned with the first pair of hexagonal tubes 220(1), 220(2) in row 1, they are also spaced so as to leave room for a first perimeter resultant cell X. A two-panel plate structure 225(1) is connected to the left sides of the hexagonal tubes 220(3), 220(4) to enclose the open lateral side of the perimeter resultant cell X. Specifically, the two panel plate structure 225(1) is connected to the hexagonal tube 220(2) and the hexagonal tube 220(3) by welding or another technique. The perimeter resultant cell X is completed by the outside surfaces of the hexagonal tubes 220(8), 220(9) located in row 2. Thus, a complete hexagonal resulting cell X is formed by the cooperation of the outside surfaces of the hexagonal tubes 220(2), 220(3), 220(8), 220(9) and the plate structure 225(1). The resultant cell X has a horizontal cross-sectional shape that corresponds to the shape of all other cells in the fuel rack 200. B. Formation of Resultant Perimeter Cells # The formation of perimeter resultant cells formed by the combination of outside surfaces of the hexagonal tubes and a single-panel plate structure (referred to above as type 201B cells) will now be described. For ease of reference and to avoid clutter, all perimeter resultant cells formed by the combination of outside surfaces of the hexagonal tubes and a single-panel plate structure are marked with an #. Turning now to row 2, the pair of hexagonal tubes 220(8), 220(9) are arranged atop the base plate 210 in an adjacent and abutting manner with each other and the hexagonal tubes 220(2), 220(3) from row 1. When so arranged, the opposing outside surfaces of the hexagonal tubes 220(8), 220(9) are in surface contact with one another. The opposing outside surfaces of the hexagonal tubes 220(8), 220(2) are also in surface contact with one another. And, the opposing outside surfaces of the hexagonal tubes 220(9), 220(3) are in surface contact with one another. The hexagonal tubes 220(14), 220(15) are arranged in row 3 atop the base plate 210 in an adjacent and abutting manner with each other and so that the outside surface of the hexagonal tube 220(15) is in surface contact with the outside surface of the hexagonal tube 220(8) from row 2. A single-panel plate structure 226(1) is connected to the hexagonal tubes 220(1), 220(14) to enclose the open lateral side of the perimeter resultant cell #. Specifically, the single-panel plate structure 226(1) is connected to the hexagonal tube 220(1) and the hexagonal tube 220(14) by welding or another technique. Thus, a complete hexagonal resulting cell # is formed by the cooperation of the outside surfaces of the hexagonal tubes 220(1), 220(2), 220(8), 220(14), 220(15) and the single-panel plate structure 226(1). The resultant cell # has a horizontal cross-sectional shape that corresponds to the shape of all other cells in the fuel rack 200. C. Formation of Resultant Perimeter Cells * The formation of non-perimeter resultant cells formed completely by the cooperation of outside surfaces of the surrounding hexagonal tubes (referred to above as type 201D cells) will now be described. For ease of reference and to avoid clutter, all non-perimeter resultant cells formed by the cooperation of the outside surfaces of the hexagonal tubes are marked with an *. In order to avoid redundancy, the arrangement and interaction of the hexagonal tubes atop the base plate will be omitted with the understanding that the discussion above is applicable. In row 2, a complete hexagonal resulting cell * is formed by the cooperation of the outside surfaces of the hexagonal tubes 220(3), 220(4), 220(9), 220(10), 220(16), 220(17). The resultant cell * has a horizontal cross-sectional shape that corresponds to the shape of all other cells in the fuel rack 200. Turning back to the general manufacture and formation of the fuel rack 200, all connections between the hexagonal tubes 220 and the base plate 210 are accomplished as described above with respect to the fuel rack 100. Additionally, connections between adjacent the hexagonal tubes 220 can be accomplished via the plug holes described above. Furthermore, in order to ensure that the resultant cells 201B-D are properly sized, the fuel rack may be formed in the following manner. First, an array of hexagonal tubes 220 are arranged in the desired geometric configuration so that all six outside surfaces of all non-perimeter hexagonal tubes 220 are in contact with the outside surface of the adjacent hexagonal tubes 220. In other words, at this stage, the fuel rack 200 only comprises cells of the type 201A that formed by the internal cavities of the hexagonal tubes 220 themselves. However, due to pre-planning, the location of those spots which are to be resultant cells 201B-D are noted. The array of hexagonal tubes 220 are positioned atop the base plate 210 and the necessary welding is performed. However, any hexagonal tubes 220 that are located in the spots where a resultant cell is desired are not welded to either the base plate 210 or to the adjacent hexagonal tubes 220. These hexagonal tubes 220 are then slidably removed from the array, thereby leaving the resultant cell 201B-D. As necessary, the single-panel plates 226 and the two-panel plates 225 are then connected to enclose the perimeter resultant cells 220B-C. Furthermore, if desired, neutron absorbing panels can be added to the array as necessary. The fuel rack 200 does not contain any flux traps. III. Slotted-Plate Fuel Rack Embodiment Referring now to FIG. 10, a fuel rack 300 that is formed from a plurality of slotted-plates arranged in a self-interlocking arrangement is illustrated. The fuel rack 300 is designed so as to have flux traps 340 and rectangular cells 301. However, it is to be understood that the slotted-plate concept described below can be utilized to form non-flux trap fuel racks and can be utilized to create fuel racks having any shaped cells, including without limitation the fuel racks discussed above. In describing the fuel rack 300 and its component parts below, relative terms such as top, bottom, above, below, horizontal, vertical upper and lower will be used in relation to the fuel rack 300 being in the illustrated substantially vertical orientation of FIG. 10. Additionally, in order to avoid clutter in the drawings, only a few of each component are numbered with the understanding that the reader will be able to identify duplicate elements. Because many of the structural and functional features of the fuel rack 300 are identical to the fuel racks 100, 200 above, only those aspect of the fuel rack 300 that are different will be discussed below with the understanding that the other concepts and structures discussed above with respect to the fuel racks 100, 200 are applicable. The fuel rack 300 generally comprises an array of cells 301 that are formed by a gridwork of slotted plates 370-372 that are slidably assembled in an interlocking rectilinear arrangement. The gridwork of slotted plates 370-372 are positioned atop and connected to a base plate 310. The entire fuel rack body is formed out of three types of slotted plates, a middle plate 370, a top plate 371 and a bottom plate 372. The bottom plate comprises the auxiliary holes 321 as discussed above for facilitating thermosiphon flow into the cells 301. Referring now to FIGS. 11A-11C, one of the middle plates 370, top plates 371 and bottom plates 372 are illustrated individually. As can be seen, the bottom plate 372 is merely a top half of the middle plate 370 with the auxiliary holes 321 cutout at its bottom edge. Similarly, the top plate 371 is merely a bottom half of the middle plate 370. The bottom and top plates 372, 371 are only used at the bottom and top of the fuel rack body to cap the middle body segments 380 (FIG. 12) formed from the middle plates 370 so that the fuel rack body has a level top and bottom edge. Each of the plates 370-372 comprise a plurality of slots 374 and end tabs 375 strategically arranged to facilitate sliding assembly to create the fuel rack body. The slots 374 are provided in both the top and bottom edges of the plates 370-372. The slots 374 on the top edge of each plate 370-372 are aligned with the slots 374 on the bottom edge of that same plate 370-372. The slots 374 extend through the plates 370-372 for one-fourth of the height of the plates 370-372. The end tabs 375 extend from lateral edges of the plates 370-372 and are preferably about one-half of the height of the plates 370-372. The end tabs 375 slidably mate with the indentations 376 in the lateral edges of adjacent plates 370-372 that naturally result from the existence of the tabs 375. The plates 370-372 are preferably constructed of a metal-matrix composite material, and more preferably a discontinuously reinforced aluminum/boron carbide metal matrix composite material, and most preferably a boron impregnated aluminum. One such suitable material is sold under the tradename Metamic™. Referring now to FIG. 12, a single middle segment 380 of the basket is illustrated. Each middle segment 380 of the fuel rack 300 comprises a gridwork of middle plates 370 arranged in a rectilinear configuration so as to form a vertical portion of the cells 301 and the flux traps 340. In creating a middle segment 380, a first middle plates 370 is arranged vertically. A second middle plate 370 is then arranged above and at a generally 90 degree angle to the first middle plate 370 so that its corresponding slots 374 are aligned. The second middle plate 370 is then lowered onto the first middle plate 370, thereby causing the slots 374 to interlock as illustrated. This is repeated with all middle plates 370 until the desired rectilinear configuration is created, thereby creating the segment 380. In creating the fuel rack body, the slots 374 and end tabs 375 of the segments 380 interlock the adjacent segments 380 together so as to prohibit relative horizontal and rotational movement between the segments 380. The segments 380 intersect and interlock with one another to form a stacked assembly that is the fuel rack body. The fuel rack 300 preferably comprises at least four of the segments 380, and more preferably at least ten segments 380. All of the segments 380 have substantially the same height and configuration. Therefore, the entire fuel rack 300 is formed of slotted plates 370-372 having what is essentially a single configuration which is the middle plate 370, with the exception that the top and bottom plates 371, 372 have to be formed by cutting the middle plate 370 and adding the cutouts 321. Furthermore, as a result of the interlocking nature of the slotted plates 370-372, spacers are not needed to maintain the flux traps 340. Thus, in some embodiments, the fuel rack 300 will be free of spacers in the flux traps 340. IV. Non-Flux Trap Fuel Basket Embodiment Referring to FIGS. 13-17, a fuel basket 1000 according an embodiment of the present invention is illustrated. The complete and assembled fuel basket 1000 is shown in FIGS. 15 and 17. While the fuel basket 1000 (and its components) are described throughout this specification in conjunction with storing and/or transporting spent nuclear fuel assemblies having a hexagonally shaped horizontal cross-sectional profile, the invention is in no way limited by the type of high level radioactive waste it is use din conjunction with. The fuel basket assembly 1000 (and its components) can be used to transport and/or store any shape of fuel assemblies. Referring now to FIG. 13, a perspective view of the storage grid portion 1001 of the fuel basket 1000 is illustrated. The storage grid 1001 is a cellular structure comprising a plurality of tubes 10 forming cells 20 for receiving and holding fuel assemblies. The tubes 10 form a honeycomb-like grid of cells 20 arranged in a polar configuration. For ease of representation (and in order to void clutter), only a few of the tubes 10 and the cells 20 are numerically identified in FIG. 13. The tubes 10 have a horizontal cross-sectional profile that is hexagonal in shape. The invention is not so limited however, and the tubes 10 will have a horizontal cross-sectional profile that corresponds with the shape of the fuel assembly to be stored within the cavities 20. For example, other polygonal-shaped SNF assemblies may be stored in the fuel basket 1000, in which case the tubes 10 will be of the appropriate horizontal cross-sectional shape. The cells 20 are substantially vertically oriented elongate spaces/cavities having a generally hexagonal horizontal cross-sectional configuration. The horizontal cross-sectional profile of the cells 20 is also not limited to hexagonal, and could be any shape including other polygons. Each cell 20 is designed to accommodate a single fuel assembly. The storage grid 1001 (and thus the cells 20) has a height that is equal to or slightly greater than the height of the fuel assembly for which the basket 1000 is designed to accommodate. The fuel basket 1000 preferably comprises 85 cells 20 and has a weight of approximately 4800 lbs. Each storage tube 10 comprises five plates 11 having an inner surface 12 and an outer surface 13. The tubes 10 could have less or more plates 11 depending upon the desired horizontal cross-sectional profile. The inner surface 12 of the tubes 10 form the cells 20. Preferably, the tubes 10 are bundled together in an axially adjacent arrangement to form a honeycomb storage grid of cells 20. The tubes 10 are formed by staking tubular segments 10A-10C so as to create cells 20 having a height equal to, or greater than the height of the fuel assembly to be stored therein. An interface/junction 21 is formed between the contacting edges of the tubular segment 10A-10C in each vertical stack that forms the tube 10. For ease of representation (and in order to void clutter), only a few of the contact interfaces are numerically identified in FIG. 13. As will be discussed in further detail below, the tubular segments 10A-10C are of varying height so that the interfaces 21 of adjacent tubes 10/cells 20 are not aligned. By ensuring that the interfaces 21 are not aligned for adjacent tubes 10/cells 20, the structural integrity of the fuel basket 1000 is enhanced. Preferably, the tubes 10 (and tubular segments 10A-10C) are made by extruding or forming plate stock followed by welding each of the plates 11 at their lateral edges. The tubes 10 are made of a material containing a neutron absorber isotope embedded in the microstructure, such as elemental boron or boron carbide. Metamic, produced by Metamic, LLC, made of an aluminum alloy matrix with embedded boron carbide is an example of an acceptable material. In some embodiments, however, the fuel basket 1000 and its components can be constructed of alternate materials, such as steel or borated stainless steel. A plurality of cutouts 23 are provided in the plates 11 at the bottom of the tubes 10. For ease of representation (and in order to void clutter), only a few of the cut-outs 23 are numerically identified in FIG. 13. The cutouts 23 form passageways through the plates 11 so that all of the cells 20 are in spatial communication. As a result, the cutouts 23 at or near the bottom of the storage grid 1001 act as a bottom plenum that helps circulate fluids (air or water) within the fuel basket 1000 (and the cells 20) to effectuate convective cooling of the stored fuel assemblies during storage and/or transportation. This natural circulation of air or water can be further facilitated by leaving one or more of the cells 20 along the periphery of the basket 100 empty so that they can act as downcomers (the support tubes 30A, 30B can also act as downcomers if the cutouts are added). The cutouts 23 are rectangular in shape in the illustrated embodiment but can take on a wide variety of shapes. Referring to FIGS. 13 and 15 concurrently, the storage grid 1001 is formed by a plurality of sections 150A-D of the tubular segments 10A-10C that are arranged in a stacked assembly. The sections 150A-D and the tubular segments 10A-C are joined with one another to form the stacked assembly that is the storage grid 1001. Each section 150A-D of the storage grid 1001 is a vertical portion of the storage grid 1001 that itself comprises the honeycomb-like grid of tubular segments 10A-C arranged in the polar configuration. The tubular segments 10A-C are of three different heights, each different height delineated by the letter A-C. A single bottom section 150A of the storage grid 1001 is illustrated in FIG. 15, the bottom section 150A having tubular segments 10A-C arranged in a polar configuration. The bottom edges of the tubular segments 10A-C are aligned at the same elevation. The two middle sections 150B-C comprise tubular members 10C, all having the same height. The top section 150D comprises tubular members 10A-C arranged so that the top surfaces of the tubular members 10A-C are aligned at the same elevation. The tubular segments 10A have a height that is preferably equal to one foot. The tubular segments 10B have a height that is preferably equal to two feet. The tubular segments 10C have a height that is preferably equal to three feet. The invention is not so limited however, and the tubular segments 10A-C may be of any height so long as the cells formed are at least equal to the height of the fuel assembly. In order to ensure that the interfaces 21 of adjacent tubes 10 are not aligned, it is preferable that no adjacent tubular members 10A-C of the bottom section 150A be of the same height. Thus when assembling the bottom section 150A, the one foot tubular segment 10A is surrounded by alternating tubular members 10B and 10C. Referring now to FIGS. 14 and 16, the fuel basket 1000 further comprises a plurality of resultant cells 20A that are formed by the outer surface 13 of the walls 11 of six of the tubular members 10. A single resultant cell 20A is illustrated in FIG. 14. Where six tubular members are joined in a polar configuration, a resultant cell 20A having a hexagonal horizontal cross-sectional profile is formed therebetween. This arrangement allows for less plates 11 (i.e., tubes 10) to be used to create the same number of cells 20, thereby creating a lighter fuel basket 1000. In the illustrated embodiment there are a total of twenty-nine resultant cells 20A. The invention is not so limited however, and the number of resultant cells may vary. The resultant cells 20A have about a ½ inch larger opening and thus can accommodate fuel assemblies that are damaged or dimensionally-deviant. Additionally, the resultant cells 20A allow for the fuel basket 1000 to be fabricated using fillet welds because the resultant cells provide easier access to the tubes 10. As best seen in FIG. 16, some resultant cells 20A that are located along the periphery of the fuel basket 1000 require the use of a closure plate 15A-B to complete the cell. The closure plates 15A-B are par-hexagonal plates. The fuel basket assembly 1000 comprises two types of closure plates 15A-B depending on the configuration of the resultant cell 20A. Some resultant cells 20A require only a single panel to enclose the cell 20A, while other resultant cells require two panels to enclose the resultant cell 20A. Where two panels are required, a closure plate 15A is formed by bending a plate into two panels of equal length and attaching the lateral edges of the closure plate 15A to the lateral edges of the tubes 10 that are forming the applicable resultant cell 20A. The basket assembly 1000 further comprises basket support tubes 30A, B placed adjacent the storage grid 1001 along its periphery. The basket support tubes 30A,B are used to provide conformal contact with the container/vessel in which the basket assembly 100 is to be used. The basket support tubes 30A, B comprise a par-hexagonal plate 31A, B connected to the concave side of a curved plate 32A, B. The curved plates 32A, 32B form a substantially circular outer perimeter for the fuel basket 1000. Because the basket support tubes 30A,B are not located at every position on the periphery of the storage grid, the substantially circular outer perimeter is circumferentially segmented. For resultant cells 20A that require only a single panel to form a six-walled cell (i.e., one side is open), the par-hexagonal plate 31B of the support tubes 30B can also function as a closure plate. The basket support tubes 30A can be used where there is no resultant cell 20A, or where the resultant cell 20A is enclosed by the closure plate 15A. Referring now to FIG. 17, the basket support tubes 30A, B are shown removed from the fuel basket assembly 1000. The basket support tubes 30A,B are provided so that the basket assembly is centered within the container (which can be a canister or a cask) in which it is to be used. Referring now to FIG. 18, the basket assembly 100 with the basket supports 30A, B attached is illustrated. The basket supports 30A, B provide for a circular outer perimeter of the fuel basket 1000. The gaps 31 in the basket supports 30A,B do not have fuel assemblies stored therein, thus they may improve the cooling of the fuel assemblies stored in the cells 20. V. Flux Trap Fuel Basket Embodiment Referring now to FIG. 19, a fuel basket 2000 according to another embodiment of the present invention is illustrated. The design aspects of the fuel assembly 2000 are substantially similar to those discussed above with respect to the fuel basket 1000. To avoid redundancy, only those design aspects of the fuel basket 2000 that substantially differ from the fuel basket 1000 will be discussed. The basket assembly 2000 comprises a plurality of flux trap spaces 50 that regulate the production of neutron radiation and prevent reactivity in a flooded condition. The flux traps 50 are small spaces that extend the height of the basket 2000. The flux traps 50 are formed between two of the tubular members 210 that are close to one another and substantially parallel. As will be discussed below, the flux traps 50 are designed to be of variable width to maximize the number of fuel assemblies that can be stored in the fuel basket 2000 while maintaining a reactivity equal to or less than 0.95. The flux traps 50 are formed between the outer surface of the storage tubes 2210. The tubes 2210 have a plurality of spacers 60 that maintain the spacing between tubular members 10 that forms the flux trap 50. Referring now to FIG. 20, a single storage tube 2210 is shown removed from the fuel basket 2000 so that it's design aspects can be more clearly visible. The storage tube 2210 has a hexagonal horizontal cross-sectional profile but can be any configuration. Preferably, the cross-sectional profile will be determined by the type of fuel assembly to be loaded and stored in the fuel basket assembly 2000. The tube 2210 comprises a plurality of notches 2215. The notches 2215 provide spaces through which the spacer 2260 (shown in FIG. 24) can be attached. Preferably, the spacers 60 are initially tack welded to the tubes 2210 so that the fuel basket 2000 can be assembled. After the fuel basket 2000 is assembled it is placed on its side and the spacers 2260 that are on the bottom (closer to the ground surface on which the basket assembly 2000 is resting) are plug welded to the storage tubes 2210. This allows for gravity to aid in the welding procedure. The fuel basket 2000 is rotated so that the next set of spacers 2260 is now closer to the ground surface for welding to the tubes 2210. The rotating and welding procedure is repeated for all of the spacers 2260 and tubes 2210. Referring to FIGS. 21 and 22 concurrently, the fuel basket 2000 is formed by a plurality of sections 2250A-F of the tubular segments 2210A-C that are arranged in a stacked assembly. The sections 2250A-F and the tubular segments 2210A-C are joined with one another to form the stacked assembly that is the fuel basket 2000. Each section 2250A-F of the fuel basket 2000 comprises a honeycomb-like grid of tubular segments 2210A-C arranged in the polar configuration. The tubular members 2210A-C of the basket assembly 2000 are of three varying heights, each height delineated by the letter A-C. A single bottom section 2250A of the basket 2000 is illustrated in FIG. 21, the bottom section 2250A has tubular segments 2210A-C arranged in a polar configuration. The middle sections 2250B-E comprise tubular segments 2210C, all having the same height. The top section 2250F comprises tubular segments 210A arranged so that the top surfaces of the tubular segments 2210A-C are aligned at the same elevation. The storage tubes 2210A have a height that is preferably equal to one foot. The tubular segments 2210B have a height that is preferably equal to two feet. The tubular segments 2210C have a height that is preferably equal to three feet. The invention is not so limited however, and the tubular segments 2210A-C may be of any height so long as the cells formed are at least equal to the height of the fuel assembly. In order that the interfaces 2221 are not aligned, it is preferable that no adjacent tubular segments 2210A-C of the bottom section 2250A be of the same height. Thus when assembling the bottom section 2250A, the one foot tubular segments 2210A are surrounded by alternating tubular segments 2210B and 2210C. The bottom edges of the tubular segments 2210A-C of the bottom section 2250A are aligned. Referring now to FIG. 23, a top perspective view of the fuel basket assembly 2000 is illustrated. The basket assembly 2000 comprises three types of spacers 2260A-C that form flux traps 50 between the tubes 2210. The first type of spacer 2260A is a unilateral triangle that maintains the largest gap between the tubes 2210 located near the central axis of the basket 2000. In other words, the spacer 2260A is used with the tubes 2210 that are closest to the center of the basket 2000 because the neutrons emitted by the fuel assemblies in the center of the basket 2000 cannot easily escape to the perimeter of the basket assembly 2000. Thus, for maintaining a reactivity of less than 0.95, a larger flux trap 50 is required. A second shape of spacer 2260B is used between the storage tubes 2210 that are nearer the outer perimeter of the fuel basket 2000. A rectangular spacer 2260C is used for the storage tubes 2210 that are nearest to the outer perimeter formed by basket supports 2230A,B. The fuel baskets 1000, 2000 of the present invention are not limited to use with any particular type of surrounding vessels. For example, in one embodiment, the basket assemblies 1000, 2000 can be incorporated into a hermetically sealable multi-purpose canister for use in conjunction with VVO style containment systems. In such an embodiment, the basket assemblies 1000, 2000 will be provided in a cavity formed by a cylindrical metal shell. The metal shell will encircle the basket assembly 1000, 2000 and a metal base plate may be welded to the bottom of the metal shell. A metal closure plate can be fitted on top of the cylinder formed by the metal shell, thereby forming a canister. Thermally conductive casks can also be used to house the fuel baskets 1000, 2000 directly. While the invention has been described and illustrated in sufficient detail that those skilled in this art can readily make and use it, various alternatives, modifications, and improvements should become readily apparent without departing from the spirit and scope of the invention. |
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abstract | This disclosure concerns a new method for preparing radioisotopes, such as molybdenum-99, by alpha particle irradiation, such as by alpha particle irradiation of zirconium-96. Molybdenum-99 is a precursor to the medically-significant radioisotope technetium-99m. Also disclosed are novel compositions containing one or more of technetium-99m, molybdenum-99 and zirconium species. Systems for producing molybdenum-99 and technetium-99m, including alpha particle generators and irradiation targets, also are described. |
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description | This application is a divisional application of U.S. patent application Ser. No. 12/460,160 filed Jul. 13, 2009, which is a continuation application of U.S. patent application Ser. No. 12/386,495 filed Apr. 16, 2009, issued as U.S. Pat. No. 8,320,513, the disclosures of which are incorporated herein for all purposes. This application generally relates to processes involving induced nuclear reactions and structures which implement such processes including orifices or fluid control means at inlet, outlet or coolant channels and more particularly relates to a nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. It is known that, in an operating nuclear fission reactor, neutrons of a known energy are absorbed by nuclides having a high atomic mass. The resulting compound nucleus separates into fission products that include two lower atomic mass fission fragments and also decay products. Nuclides known to undergo such fission by neutrons of all energies include uranium-233, uranium-235 and plutonium-239, which are fissile nuclides. For example, thermal neutrons having a kinetic energy of 0.0253 eV (electron volts) can be used to fission U-235 nuclei. Fission of thorium-232 and uranium-238, which are fertile nuclides, will not undergo induced fission, except with fast neutrons that have a kinetic energy of at least 1 MeV (million electron volts). The total kinetic energy released from each fission event is about 200 MeV. This kinetic energy is eventually transformed into heat. In nuclear reactors, the afore-mentioned fissile and/or fertile material is typically housed in a plurality of closely packed together fuel assemblies, which define a nuclear reactor core. It has been observed that heat build-up may cause such closely packed together fuel assemblies and other reactor components to undergo differential thermal expansion leading to misalignment of the reactor core components. Heat build-up may also contribute to fuel rod creep that can increase risk of fuel rod swelling and fuel rod cladding rupture during reactor operation. This may increase the risk that fuel pellets might crack and/or fuel rods might bow. Fuel pellet cracking may precede pellet-cladding failure mechanisms, such as pellet-clad mechanical interaction, and lead to fission gas release. Fission gas release can produce higher than normal radiation levels in the reactor core. Fuel rod bow may lead to obstruction of coolant flow channels. Attempts have been made to provide adequate coolant flow to nuclear reactor fuel assemblies. U.S. Pat. No. 4,505,877, issued Mar. 19, 1985 in the name of Jacky Rion and titled “Device for Regulating the Flow of a Fluid”, discloses a device comprising a series of gratings perpendicular to the fluid flow and that change direction of the fluid flow. According to the Rion patent, this device is intended for use in the regulation of the direction of a cooling fluid circulating in the base of a liquid metal-cooled nuclear reactor assembly. The device is directed toward bringing about a given pressure drop for a given nominal flow rate and a given down-stream pressure, without producing cavitation. Another attempt to provide adequate coolant flow to nuclear reactor fuel assemblies is disclosed in U.S. Pat. No. 5,066,453, issued Nov. 19, 1991 in the names of Neil G. Heppenstall et al. and titled “Nuclear Fuel Assembly Coolant Control.” This patent discloses an apparatus for controlling the flow of coolant through a nuclear fuel assembly, the apparatus comprising a variable flow restrictor locatable in the fuel assembly, means responsive to neutron radiation at a location in the fuel assembly in a manner to cause neutron induced growth of the responsive means, and a connecting means for connecting the neutron radiation responsive means to the variable flow restrictor for controlling the flow of coolant through the fuel assembly. The variable flow restrictor comprises a plurality of longitudinally aligned ducts, and a plugging means having an array of plugging members locatable in some of the ducts, the plugging members being of different lengths so that longitudinal displacement of the plugging means by the connecting means progressively opens or closes some of the ducts. Yet another attempt to provide adequate coolant flow to nuclear reactor fuel assemblies is disclosed in U.S. Pat. No. 5,198,185 issued Mar. 30, 1993 in the name of John P. Church and titled “Nuclear Reactor How Control Method and Apparatus.” This patent appears to disclose a coolant flow distribution that results in improved flow during accident conditions without degrading flow during nominal conditions. According to this patent, a universal sleeve housing surrounds a fuel element. The universal sleeve housing has a plurality of holes to allow passage of coolant. A variation is imposed in the number and size of holes in the sleeve housings from one sleeve to another to increase amount of coolant flowing to the fuel in the center of the core and decrease, relatively, flow to the peripheral fuel. Also, according to this patent, varying the number of holes and size of holes can meet a particular power shape across the core. According to an aspect of this disclosure, there is provided a nuclear fission reactor, comprising a nuclear fission module configured to have at least a portion of a traveling burn wave at a location relative to the nuclear fission module; and a flow control assembly configured to be coupled to the nuclear fission module and configured to modulate flow of a fluid in response to the traveling burn wave at the location relative to the nuclear fission module. According to an another aspect of the disclosure there is provided a nuclear fission reactor, comprising a heat-generating nuclear fission fuel assembly configured to have at least a portion of a traveling burn wave at a location relative to the nuclear fission fuel assembly; and a flow control assembly configured to be coupled to the nuclear fission fuel assembly and capable of modulating flow of a fluid stream in response to the traveling burn wave at the location relative to the nuclear fission fuel assembly. According to yet another aspect of the disclosure there is provided, for use in a traveling wave nuclear fission reactor, a flow control assembly, comprising a flow regulator subassembly. According to another aspect of the disclosure there is provided, for use in a nuclear fission reactor, a flow control assembly, comprising a flow regulator subassembly, the flow regulator subassembly including a first sleeve having a first hole; a second sleeve configured to be inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole, the first sleeve being configured to rotate for bringing the first hole into alignment with the second hole; and a carriage subassembly configured to be coupled to the flow regulator subassembly. According to still another aspect of the disclosure there is provided, for use in a traveling wave nuclear fission reactor, a flow control assembly configured to be connected to a fuel assembly, comprising an adjustable flow regulator subassembly configured to be disposed in a fluid stream. According to a further aspect of the disclosure there is provided, for use in a nuclear fission reactor, a flow control assembly configured to be connected to a fuel assembly, comprising an adjustable flow regulator subassembly configured to be disposed in a fluid stream, the adjustable flow regulator subassembly including a first sleeve having a first hole; and a second sleeve configured to be inserted into the first sleeve, the second sleeve having a second hole, the first hole being progressively alignable with the second hole, whereby a variable amount of the fluid stream flows through the first hole and the second hole as the first hole progressively aligns with the second hole, the first sleeve being configured to axially translate relative to the second sleeve for aligning the second hole with the first hole. According to an additional aspect of the disclosure there is provided, for use in a nuclear fission reactor, a flow control assembly configured to be connected to a fuel assembly, comprising an adjustable flow regulator subassembly; and a carriage subassembly coupled to the adjustable flow regulator subassembly for adjusting the adjustable flow regulator subassembly. According to another aspect of the disclosure there is provided, for use in a nuclear fission reactor, a flow control assembly couplable to a selected one of a plurality of nuclear fission fuel assemblies arranged for disposal in the nuclear fission reactor, comprising an adjustable flow regulator subassembly for modifying flow of a fluid stream flowing through the selected one of the plurality of nuclear fission fuel assemblies, the adjustable flow regulator subassembly including an outer sleeve having a plurality of first holes; an inner sleeve inserted into the outer sleeve, the inner sleeve having a plurality of second holes, the first holes being progressively alignable with the second holes for defining a variable flow area, whereby a variable amount of the fluid stream flows through the first holes and the second holes as the first holes and the second holes progressively align to define the variable flow area; and a carriage subassembly coupled to the adjustable flow regulator subassembly for adjusting the adjustable flow regulator subassembly. According to a further aspect of the disclosure there is provided a method of operating a nuclear fission reactor, comprising producing at least a portion of a traveling burn wave at a location relative to a nuclear fission module; and operating a flow control assembly coupled to the nuclear fission module to modulate flow of a fluid in response to the location relative to the nuclear fission module. According to another aspect of the disclosure there is provided a method of assembling a flow control assembly for use in a traveling wave nuclear fission reactor, comprising receiving a flow regulator subassembly. According to another aspect of the disclosure there is provided a method of assembling a flow control assembly for use in a traveling wave nuclear fission reactor, comprising receiving a carriage subassembly. According to another aspect of the disclosure there is provided a method of assembling a flow control assembly for use in a nuclear fission reactor, comprising receiving a first sleeve having a first hole; inserting a second sleeve into the first sleeve, the second sleeve having a second hole alignable with the first hole, the first sleeve being configured to rotate for axially translating the first hole into alignment with the second hole; and coupling a carriage assembly to the flow regulator subassembly. According to an additional aspect of the disclosure there is provided, for use in a traveling wave nuclear fission reactor, a flow control assembly system, comprising a flow regulator subassembly. According to another aspect of the disclosure there is provided, for use in a nuclear fission reactor, a flow control assembly system, comprising a flow regulator subassembly, the flow regulator subassembly including a first sleeve having a first hole; a second sleeve configured to be inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole, the first sleeve being configured to rotate for axially translating the first hole into alignment with the second hole; and a carriage subassembly configured to be coupled to the flow regulator subassembly. According to yet another aspect of the disclosure there is provided, for use in a nuclear fission reactor, a flow control assembly system configured to be connected to a nuclear fission fuel assembly, comprising an adjustable flow regulator subassembly configured to be disposed in a fluid stream. According to another aspect of the disclosure there is provided, for use in a nuclear fission reactor, a flow control assembly system couplable to a selected one of a plurality of nuclear fission fuel assemblies disposed in the nuclear fission reactor, comprising an adjustable flow regulator subassembly for controlling flow of a fluid stream flowing through the selected one of the plurality of nuclear fission fuel assemblies, the adjustable flow regulator subassembly including an outer sleeve having a plurality of first holes; an inner sleeve inserted into the outer sleeve, the inner sleeve having a plurality of second holes, the first holes being progressively alignable with the second holes for defining a variable flow area, whereby a variable amount of the fluid stream flows through the first holes and the second holes as the first holes and the second holes progressively align to define the variable flow area; and a carriage subassembly coupled to the adjustable flow regulator subassembly for adjusting the adjustable flow regulator subassembly. A feature of the present disclosure is the provision of a flow control assembly capable of controlling flow of a fluid in response to location of a burn wave. Another feature of the present disclosure is the provision of a flow control assembly comprising a flow regulator subassembly including an outer sleeve and an inner sleeve, the outer sleeve having a first hole and the inner sleeve having a second hole alignable with the first hole, whereby an amount of a fluid stream flows through the first hole and the second hole as the second hole aligns with the first hole. An additional feature of the present disclosure is the provision of a carriage subassembly configured to be coupled to the flow regulator subassembly for carrying and configuring the flow regulator subassembly. In addition to the foregoing, various other method and/or device aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, references is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented herein. In addition, the present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Moreover, the herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. With respect to the present disclosure and as previously mentioned, in many cases, for every neutron that is absorbed in a fissile nuclide, more than one neutron is liberated until the fissile nuclei are depleted. This phenomenon is used in a commercial nuclear reactor to produce continuous heat that, in turn, is used to generate electricity. However, heat damage to reactor structural materials may occur due to “peak” temperature (i.e., hot channel peaking factor) which occurs due to uneven neutron flux distribution in the reactor core. As well known in the art, neutron flux is defined as the number of neutrons passing through a unit area per unit time. This peak temperature is, in turn, due to heterogeneous control rod/fuel rod distribution. The heat damage may occur if the peak temperature exceeds material limits. In addition, reactors operating in the fast neutron spectrum may be designed to have a fertile fuel “breeding blanket” material present at the core periphery. Such reactors will tend to breed fuel into the breeding blanket material through neutron absorption. This results in an increasing power output in the reactor periphery as the reactor approaches the end of a fuel cycle. Flow of coolant through the peripheral assemblies at the beginning of a reactor fuel cycle can maintain a safe operating temperature and account for the increase in power which will occur as burn-up increases during the fuel cycle. A “reactivity” (i.e., change in reactor power) is produced because of fuel “burnup”. Burn-up is typically defined as the amount of energy generated per unit mass of fuel and is usually expressed in units of megawatt-days per metric tonne of heavy metal (MWd/MTHM) or gigawatt-days per metric tonne of heavy metal (GWd/MTHM). More specifically, reactivity change is related to the relative ability of the reactor to produce more or less neutrons than the exact amount to sustain a critical chain reaction. Responsiveness of a reactor is typically characterized as the time derivative of a reactivity change causing the reactor to increase or decrease in power exponentially. In this regard, control rods made of neutron absorbing material are typically used to adjust and control the changing reactivity. Such control rods are reciprocated in and out of the reactor core to variably control neutron absorption and thus the neutron flux level and reactivity in the reactor core. The neutron flux level is depressed in the vicinity of the control rod and potentially higher in areas remote from the control rod. Thus, the neutron flux is not uniform across the reactor core. This results in higher fuel burnup in those areas of higher neutron flux. Also, it may be appreciated by a person of ordinary skill in the art of nuclear power production, that neutron flux and power density variations are due to many factors. Proximity to a control rod may or may not be the primary factor. For example, the neutron flux typically drops significantly at core boundaries with no nearby control rod. These effects, in turn, may cause overheating or peak temperatures in those areas of higher neutron flux. Such peak temperatures may undesirably reduce the operational life of structures subjected to such peak temperatures by altering the mechanical properties of the structures. Also, reactor power density, which is proportional to the product of the neutron flux and the fissile fuel concentration, is limited by the ability of core structural materials to withstand such peak temperatures without damage. Therefore, referring to FIG. 1, by way of example only and not by way of limitation, there is shown a nuclear fission reactor, generally referred to as 10, that addresses the concerns recited hereinabove. As described more fully hereinbelow, reactor 10 may be a traveling wave nuclear fission reactor. Nuclear fission reactor 10 generates electricity that is transmitted over a plurality of transmission lines (not shown) to users of the electricity. Reactor 10 alternatively may be used to conduct tests, such as tests to determine effects of temperature on reactor materials. Referring to FIGS. 1, 1A, 1B and 2, a reactor 10 comprises a nuclear fission reactor core, generally referred to as 20, that includes a plurality of nuclear fission fuel assemblies or, as also referred to herein, nuclear fission modules 30. Nuclear fission reactor core 20 is sealingly housed within a reactor core enclosure 35. By way of example only and not by way of limitation, each nuclear fission module 30 may form a hexagonally-shaped structure in transverse cross-section, as shown, so that more nuclear fission modules 30 may be closely packed together within reactor core 20, as compared to most other shapes for nuclear fission module 30, such as cylindrical or spherical shapes. Each nuclear fission module 30 comprises a plurality of fuel rods 40 for generating heat due to the aforementioned nuclear fission chain reaction process. Fuel rods 40 may be surrounded by a fuel rod canister 43, if desired, for adding structural rigidity to nuclear fission modules 30 and for segregating nuclear fission modules 30 one from another. Segregating nuclear fission modules 30 one from another avoids transverse coolant cross flow between adjacent nuclear fission modules 30. Avoiding transverse coolant cross flow prevents transverse vibration of nuclear fission modules 30. Such transverse vibration might otherwise increase risk of damage to fuel rods 40. In addition, segregating nuclear fission modules 30 one from another allows control of coolant flow on an individual module-by-module basis, as described more fully hereinbelow. Controlling coolant flow to individual, preselected nuclear fission modules 30 efficiently manages coolant flow within reactor core 20, such as directing coolant flow substantially according to the nonuniform temperature distribution in reactor core 20. Canister 43 may include an annular shoulder portion 46 (see FIG. 7) for resting bundled together fuel rods 40 thereon. The coolant may have an average nominal volumetric flow rate of approximately 5.5 m.sup.3/sec (i.e., approximately 194 cubic ft.sup.3/sec) and an average nominal velocity of approximately 2.3 msec (i.e., approximately 7.55 ft/sec) in the case of an exemplary sodium cooled reactor during normal operation. Fuel rods 40 are adjacent one to another and define a coolant flow channel 47 (see FIG. 7) therebetween for allowing flow of coolant along the exterior of fuel rods 40. Fuel rods 40 are bundled together so as to form the previously mentioned hexagonal nuclear fission modules 30. Although fuel rods 40 are adjacent to each other, fuel rods 40 are nonetheless maintained in a spaced-apart relationship by a wire wrapper 50 (see FIG. 7) that extends spirally along the length of each fuel rod 40, according to techniques known by persons of skill in the art of nuclear power reactor design. With particular reference to FIG. 1B, each fuel rod 40 has a plurality of nuclear fuel pellets 60 stacked end-to-end therein, which nuclear fuel pellets 60 are sealingly surrounded by a fuel rod cladding material 70. Nuclear fuel pellets 60 comprise the afore-mentioned fissile nuclide, such as uranium-235, uranium-233 or plutonium-239. Alternatively, nuclear fuel pellets 60 may comprise a fertile nuclide, such as thorium-232 and/or uranium-238 which will be transmuted during the fission process into the fissile nuclides mentioned immediately hereinabove. A further alternative is that nuclear fuel pellets 60 may comprise a predetermined mixture of fissile and fertile nuclides. More specifically, by way of example only and not by way of limitation, nuclear fuel pellets 60 may be made from an oxide selected from the group consisting essentially of uranium monoxide (UO), uranium dioxide (UO2), thorium dioxide (ThO2) (also referred to as thorium oxide), uranium trioxide (UO3), uranium oxide-plutonium oxide (UO—PuO), triuranium octoxide (U3O8) and mixtures thereof. Alternatively, nuclear fuel pellets 60 may substantially comprise uranium either alloyed or unalloyed with other metals, such as, but not limited to, zirconium or thorium metal. As yet another alternative, nuclear fuel pellets 60 may substantially comprise a carbide of uranium (UCx) or a carbide of thorium (ThCx). For example, nuclear fuel pellets 60 may be made from a carbide selected from the group consisting essentially of uranium monocarbide (UC), uranium dicarbide (UC2), uranium sesquicarbide (U2C3), thorium dicarbide (ThC2), thorium carbide (ThC) and mixtures thereof. As another non-limiting example, nuclear fuel pellets 60 may be made from a nitride selected from the group consisting essentially of uranium nitride (U3N2), uranium nitride-zirconium nitride (U3N2Zr3N4), uranium-plutonium nitride ((U—Pu)N), thorium nitride (ThN), uranium-zirconium alloy (UZr) and mixtures thereof. Fuel rod cladding material 70, which sealingly surrounds the stack of nuclear fuel pellets 60, may be a suitable zirconium alloy, such as ZIRCOLOY™ (trademark of the Westinghouse Electric Corporation), which has known resistance to corrosion and cracking. Cladding 70 may be made from other materials, as well, such as ferritic martensitic steels. As best seen in FIG. 1, reactor core 20 is disposed within a reactor pressure vessel 80 for preventing leakage of radioactive particles, gasses or liquids from reactor core 20 to the surrounding biosphere. Pressure vessel 80 may be steel, concrete or other material of suitable size and thickness to reduce risk of such radiation leakage and to support required pressure loads. In addition, there may be a containment vessel (not shown) sealingly surrounding parts of reactor 10 for added assurance that leakage of radioactive particles, gasses or liquids from reactor core 20 to the surrounding biosphere is prevented. Referring again to FIG. 1, a primary loop coolant pipe 90 is coupled to reactor core 20 for allowing a suitable coolant to flow through reactor core 20 in order to cool reactor core 20. Primary loop coolant pipe 90 may be made from any suitable material, such as stainless steel. It may be appreciated that, if desired, primary coolant loop pipe 90 may be made not only from ferrous alloys, but also from non-ferrous alloys, zirconium-based alloys or other structural materials or composites. The coolant carried by primary loop coolant pipe 90 may be a noble gas or mixture of noble gases. Alternatively, the coolant may be other fluids such as “light” water (H2O) or gaseous or supercritical carbon dioxide (CO2). As another example, the coolant may be a liquid metal. Such a liquid metal may be a lead (Pb) alloy, such as lead-bismuth (Pb—Bi). Further, the coolant may be an organic-based coolant, such as a polyphenyl or a fluorocarbon. In the exemplary embodiment disclosed herein, the coolant may suitably be a liquid sodium (Na) metal or sodium metal mixture, such as sodium-potassium (Na—K). As an example and depending on the particular reactor core design and operating history, normal operating temperature of a sodium-cooled reactor core may be relatively high. For instance, in the case of a 500 to 1,500 MWe sodium-cooled reactor with mixed uranium-plutonium oxide fuel, the reactor core outlet temperature during normal operation may range from approximately 510° Celsius (i.e., 950° Fahrenheit) to approximately 550° Celsius (i.e., 1,020° Fahrenheit). On the other hand, during a LOCA (Loss of Coolant Accident) or LOFTA (Loss of Flow Transient Accident) peak fuel cladding temperatures may reach about 600° Celsius (i.e. 1,110° Fahrenheit) or more, depending on reactor core design and operating history. Moreover, decay heat build-up during post-LOCA or post-LOFTA scenarios and also during suspension of reactor operations may produce unacceptable heat accumulation. In some cases, therefore, it is appropriate to control coolant flow to reactor core 20 during both normal operation and post accident scenarios. Moreover, the temperature profile in reactor core 20 varies as a function of location. In this regard, the temperature distribution in reactor core 20 may closely follow the power density spatial distribution in reactor core 20. It is known that the power density near the center of reactor core 20 is generally higher than near the periphery of reactor core 20, in the absence of a suitable neutron reflector or neutron breeding “blanket” surrounding the periphery of reactor core 20. Thus, it is to be expected that coolant flow parameters for nuclear fission modules 30 near the periphery of reactor core 20 would be less than coolant flow parameters for nuclear fission modules 30 near the center of reactor core 20, especially at the beginning of core life. Hence, in this case, it would be unnecessary to provide the same or uniform coolant mass flow rate to each nuclear fission module 30. As described in detail hereinbelow, a technique is provided to vary coolant flow to individual nuclear fission modules 30 depending on location of nuclear fission modules 30 in reactor core 20 and desired reactor operating results. Still referring to FIG. 1, the heat-bearing coolant generated by reactor core 20 flows along a coolant flow path 95 to an intermediate heat exchanger 100, for reasons described presently. The coolant flowing along coolant flow path 95 flows through intermediate heat exchanger 100 and into a plenum volume 105 associated with intermediate heat exchanger 100. After flowing into plenum volume 105, the coolant continues through primary loop pipe 90, as shown by a plurality of arrows 107. It may be appreciated that the coolant leaving plenum volume 105 has been cooled due to the heat transfer occurring in intermediate heat exchanger 100. A first pump 110 is coupled to primary loop pipe 90, and is in fluid communication with the reactor coolant carried by primary loop pipe 90, for pumping the reactor coolant through primary loop pipe 90, through reactor core 20, along coolant flow path 95, into intermediate heat exchanger 100, and into plenum volume 105. Referring again to FIG. 1, a secondary loop pipe 120 is provided for removing heat from intermediate heat exchanger 100. Secondary loop pipe 120 comprises a secondary “hot” leg pipe segment 130 and a secondary “cold” leg pipe segment 140. Secondary cold leg pipe segment 140 is integrally formed with secondary hot leg pipe segment 130 so as to form a closed loop that defines secondary loop pipe 120, as shown. Secondary loop pipe 120, which is defined by hot leg pipe segment 130 and cold leg pipe segment 140, contains a fluid, which suitably may be liquid sodium or a liquid sodium mixture. Secondary hot leg pipe segment 130 extends from intermediate heat exchanger 100 to a steam generator and superheater combination 143 (hereinafter referred to as “steam generator 143”), for reasons described momentarily. After passing through steam generator 143, the coolant flowing through secondary loop pipe 120 and exiting steam generator 143 is at a lower temperature than before entering steam generator 143 due to the heat transfer occurring within steam generator 143. After passing through steam generator 143, the coolant is pumped, such as by means of a second pump 145, along “cold” leg pipe segment 140, which terminates in intermediate heat exchanger 100. The manner in which steam generator 143 generates steam is generally described immediately hereinbelow. Referring yet again to FIG. 1, disposed in steam generator 143 is a body of water 150 maintained at a predetermined temperature and pressure. The fluid flowing through secondary hot leg pipe segment 130 will surrender its heat to body of water 150, which is at a lower temperature than the fluid flowing through secondary hot leg pipe segment 130. As the fluid flowing through secondary hot leg pipe segment 130 surrenders its heat to body of water 150, a portion of body of water 150 will vaporize to steam 160 according to the temperature and pressure within steam generator 143. Steam 160 will then travel through a steam line 170 which has one end thereof in vapor communication with steam 160 and another end thereof in liquid communication with body of water 150. A rotatable turbine 180 is coupled to steam line 170, such that turbine 180 rotates as steam 160 passes therethrough. An electrical generator 190, which is connected to turbine 180, such as by a rotatable turbine shaft 195, generates electricity as turbine 180 rotates. In addition, a condenser 200 is coupled to steam line 170 and receives the steam passing through turbine 180. Condenser 200 condenses the steam to liquid water and passes any waste heat to a heat sink, such as a cooling tower 210, which is associated with reactor 10. The liquid water condensed by condenser 200 is pumped along steam line 170 from condenser 200 to steam generator 143 by means of a third pump 220 interposed between condenser 200 and steam generator 143. Turning now to FIGS. 2, 3 and 4, there are shown in transverse cross section, exemplary configurations for reactor core 20. In this regard, nuclear fission modules 30 may be arranged to define a hexagonally-shaped configuration, generally referred to as 230, for reactor core 20. Alternatively, nuclear fission modules 30 may be arranged to define a cylindrically-shaped configuration, generally referred to as 240, for reactor core 20. As another alternative, nuclear fission modules 30 may be arranged to define a parallelpiped-shaped configuration, generally referred to as 250, for reactor core 20. In this regard, reactor core 250 has a first end 252 and a second end 254 for reasons provided hereinbelow. Referring to FIG. 5, regardless of the configuration chosen for reactor core 20, a plurality of spaced-apart, longitudinally extending and longitudinally movable control rods 260 are symmetrically disposed within a control rod guide tube or cladding (not shown), extending the length of a predetermined number of nuclear fission modules 30. Control rods 260, which are shown disposed in a predetermined number of the hexagonally-shaped nuclear fission modules 30, control the neutron fission reaction occurring in nuclear fission modules 30. Control rods 260 comprise a suitable neutron absorber material having an acceptably high neutron absorption cross-section. In this regard, the absorber material may be a metal or metalloid selected from the group consisting essentially of lithium, silver, indium, cadmium, boron, cobalt, hafnium, dysprosium, gadolinium, samarium, erbium, europium and mixtures thereof. Alternatively, the absorber material may be a compound or alloy selected from the group consisting essentially of silver-indium-cadmium, boron carbide, zirconium diboride, titanium diboride, hafnium diboride, gadolinium titanate, dysprosium titanate and mixtures thereof. Control rods 260 will controllably supply negative reactivity to reactor core 20. Thus, control rods 260 provide a reactivity management capability to reactor core 20. In other words, control rods 260 are capable of controlling or are configured to control the neutron flux profile across reactor core 20 and thus influence the temperature profile across reactor core 20. Referring to FIGS. 5A and 5B, alternative embodiments of nuclear fission module 30 are shown. It may be appreciated that nuclear fission module 30 need not be neutronically active. In other words, nuclear fission module 30 need not contain any fissile material. In this case, nuclear fission module 30 may be a purely reflective assembly or a purely fertile assembly or a combination of both. In this regard, nuclear fission module 30 may be a breeder nuclear fission module comprising nuclear breeding material or a reflective nuclear fission module comprising reflective material. Alternatively, in one embodiment, nuclear fission module 30 may contain fuel rods 40 in combination with nuclear breeding rods or reflector rods. For example, in FIG. 5A, a plurality of fertile nuclear breading rods 270 are disposed in nuclear fission module 30 in combination with fuel rods 40. Control rods 260 may also be present. The fertile nuclear breeding material in nuclear breeding rods 270 may be thorium-232 and/or uranium-238, as mentioned hereinabove. In this manner, nuclear fission module 30 defines a fertile nuclear breeding assembly. In FIG. 5B, a plurality of neutron reflector rods 274 are disposed in nuclear fission module 30 in combination with fuel rods 40. Control rods 260 may also be present. The reflector material may be a material selected from the group consisting essentially of beryllium (Be), tungsten (W), vanadium (V), depleted uranium (U), thorium (Th), lead alloys and mixtures thereof. Also, reflector rods 274 may be selected from a wide variety of steel alloys. In this manner, nuclear fission module 30 defines a neutron reflector assembly. Moreover, it may be appreciated by a person of ordinary skill in the art of nuclear in-core fuel management that nuclear fission module 30 may include any suitable combination of nuclear fuel rods 40, control rods 260, breeding rods 270 and reflector rods 274. FIG. 5C shows another embodiment of the previously mentioned reactor core 250. In FIG. 5C, a breeding blanket comprising a plurality of breeding nuclear fission modules 276 containing fertile material are disposed around an interior periphery of parallelpiped reactor core 250. The breeding blanket breeds fissile material therein. Returning to FIG. 4, regardless of the configuration selected for nuclear fission reactor core 20, the nuclear fission reactor core 20 may be configured as a traveling wave nuclear fission reactor core, such as exemplary reactor core 250. In this regard, a comparatively small and removable nuclear fission igniter 280, that includes a moderate isotopic enrichment of nuclear fissionable material, such as, without limitation, U-233, U-235 or Pu-239, is suitably located in reactor core 250. By way of example only and not by way of limitation, igniter 280 may be located near first end 252 that is opposite second end 254 of reactor core 250. Neutrons are released by igniter 280. The neutrons that are released by igniter 280 are captured by fissile and/or fertile material within nuclear fission modules 30 to initiate the fission chain reaction. Igniter 280 may be removed once the fission chain reaction becomes self-sustaining, if desired. Referring again to FIG. 4, igniter 280 initiates a three-dimensional, traveling deflagration wave or “burn wave” 290 having a width “x”. When igniter 280 releases its neutrons to cause “ignition”, burn wave 290 travels outwardly from igniter 280 near first end 252 and toward second end 254 of reactor core 250, so as to form the propagating burn wave 290. In other words, each nuclear fission module 30 is capable of accepting at least a portion of traveling burn wave 290 as burn wave 290 propagates through reactor core 250. Speed of the traveling burn wave 290 may be constant or non-constant. Thus, the speed at which burn wave 290 propagates can be controlled. For example, longitudinal movement of the previously mentioned control rods 260 (see FIG. 5) in a predetermined or programmed manner can drive down or lower neutronic reactivity of fuel rods 40 that are disposed in nuclear fission modules 30. In this manner, neutronic reactivity of fuel rods 40 that are presently being burned at the location of burn wave 290 is driven down or lowered relative to neutronic reactivity of “unburned” fuel rods 40 ahead of burn wave 290. This result gives the burn wave propagation direction indicated by an arrow 295. The basic principles of such a traveling wave nuclear fission reactor is disclosed in more detail in U.S. Patent Application Publication No. 2008/0123797 by Roderick A. Hyde, et al. and titled “Automated Nuclear Power Reactor for Long-Term Operation”, which application is assigned to the assignee of the present application, the entire disclosure of which is hereby incorporated by reference. Referring to FIGS. 6 and 7, there are shown upright adjacent hexagonally-shaped nuclear fission modules 30. Only three adjacent nuclear fission modules 30 are shown, it being understood that a greater number of nuclear fission modules 30 are present in reactor core 20. In addition, each nuclear fission module 30 comprises the plurality of the previously mentioned fuel rods 40. Each nuclear fission module 30 is mounted on a horizontally extending reactor core lower support plate 360. Reactor core lower support plate 360 extends across all nuclear fission modules 30. Reactor core lower support plate 360 has a counter bore 370 therethrough for reasons provided hereinbelow. Counter bore 370 has an open end 380 for allowing flow of coolant thereinto. Horizontally extending across a top portion or exit portion of each nuclear fission module 30 and removably connected thereto is a reactor core upper support plate 400 that caps each nuclear fission module 30. Reactor core upper support plate 400 also defines a plurality of flow slots 410 for allowing flow of coolant therethrough. As previously mentioned, it is important to control the temperature of reactor core 20 and the nuclear fission modules 30 therein, regardless of the configuration selected for reactor core 20. Proper temperature control is important for several reasons. For example, heat damage may occur to reactor core structural materials if the peak temperature exceeds material limits. Such peak temperatures may undesirably reduce the operational life of structures subjected to such peak temperatures by altering the mechanical properties of the structures, particularly those properties relating to thermal creep. Also, reactor power density is limited by the ability of core structural materials to withstand such high temperatures without damage. In addition, reactor 10 alternatively may be used to conduct tests, such as tests to determine effects of temperature on reactor materials. Controlling reactor core temperature is important for successfully conducting such tests. In addition, nuclear fission modules 30 residing at or near the center of reactor core 20 may generate more heat than nuclear fission modules 30 residing at or near the periphery of reactor core 20 in the absence of a neutron reflector or neutron breeding blanket surrounding the periphery of reactor core 20. Therefore, it would be inefficient to supply a uniform coolant mass flow rate across reactor core 20 because hotter nuclear fission modules 30 near the center of reactor core 20 would involve a higher coolant mass flow rate than nuclear fission modules 30 near the periphery of reactor core 20. The disclosure herein provides a technique to address these concerns. With reference to FIGS. 1, 6 and 7, first pump 110 and primary loop pipe 90 deliver reactor coolant to nuclear fission modules 30 along a coolant flow path or fluid stream indicated by flow arrows 420. The primary coolant then continues along coolant flow path 420 and through open end 380 that is formed in lower support plate 360. As described in more detail hereinbelow, the reactor coolant can be used to remove heat from or cool selected ones of nuclear fission modules 30 at the location of traveling burn wave 290. The nuclear fission module 30 may be selected, at least in part, on the basis of whether or not burn wave 290 is located, detected, or otherwise resides within or in the vicinity of the nuclear fission module 30, as described in more detail hereinbelow. Referring again to FIGS. 1, 6 and 7, in order to achieve the desired result of cooling the selected one of nuclear fission modules 30, an adjustable flow regulator subassembly 430 is coupled to nuclear fission module 30. Flow regulator subassembly 430 controls flow of the coolant in response to the location of burn wave 290 (see FIG. 4) relative to nuclear fission modules 30 and also in response to certain operating parameters associated with nuclear fission module 30. In other words, flow regulator subassembly 430 is capable of supplying or is configured to supply a relatively lesser amount of coolant to nuclear fission module 30 when a lesser amount of burn wave 290 (i.e., lesser intensity of burn wave 290) is present within nuclear fission module 30. On the other hand, flow regulator subassembly 430 is capable of supplying or is configured to supply a relatively greater amount of coolant to nuclear fission module 30 when a greater amount of burn wave 290 (i.e., greater intensity of burn wave 290) is present within nuclear fission module 30. Presence and intensity of burn wave 290 may be identified by heat generation rate, neutron flux level, power level or other suitable operating characteristic associated with nuclear fission module 30. Referring to FIGS. 7, 8, 8A, 8B, 8C, and 8D, adjustable flow regulator subassembly 430 extends through counter bore 370 for regulating flow of fluid stream 420 into nuclear fission module 30. It will be understood by a person of ordinary skill in the art that, in order to regulate flow of fluid stream 420, flow regulator subassembly 430 provides a controllable flow resistance. Flow regulator subassembly 430 comprises a generally cylindrical first or outer sleeve 450 having a plurality of first ligaments 460, which define respective ones of a plurality of axially spaced-apart first holes or first controllable flow apertures 470 radially distributed around outer sleeve 450. Outer sleeve 450 further comprises a first nipple 480 which may have a hexagonally-shaped transverse cross section for reasons provided hereinbelow. First nipple 480 defines a threaded internal cavity 500 for reasons provided hereinbelow. Referring again to FIGS. 7, 8, 8A, 8B, 8C and 8D, flow regulator subassembly 430 further comprises a generally cylindrical second or inner sleeve 530 that is threadably received into outer sleeve 450, as disclosed in more detail hereinbelow. In one embodiment, inner sleeve 530 may be integrally formed with nuclear fission module 30 during fabrication of fission module 30, such that inner sleeve 530 is a permanent portion of nuclear fission module 30. In another embodiment, inner sleeve 530 may be removably connected to nuclear fission module 30, such that inner sleeve 530 is readily separable from nuclear fission module 30 and hence not a permanent portion of nuclear fission module 30. In either embodiment, inner sleeve 530 comprises a plurality of second ligaments 540, which define respective ones of a plurality of axially spaced-apart second holes or second controllable flow apertures 550 radially distributed around inner sleeve 530. Inner sleeve 530 further comprises an externally threaded second nipple 560 sized to be threadably received into threaded internal cavity 500 of bottom portion 490 that belongs to outer sleeve 450. A top portion 570 of inner sleeve 530 includes a cap 580, which may or may not be permanently formed with nuclear fission module 30, as previously mentioned. An internal bore 590 extends through top portion 570, including through cap 580, for passage of the coolant therethrough. Coupled to cap 580 and fuel rods 40 may be a frusto-connical funnel portion 600 having an inner surface 605 in communication with internal bore 590 and the interior of canister 43 for allowing passage of the coolant from internal bore 590 and into canister 43 where fuel rods 40 reside. As previously mentioned, nuclear fission modules 30 are capable of having or are configured to have a temperature dependent reactivity change. Thus, flow control regulator subassembly 430 is at least partially configured to control temperature within nuclear fission module 30 by controlling coolant flow into nuclear fission module 30 in order to effect such a temperature dependent reactivity change. Referring now to FIGS. 8A and 8D, bottom portion 490 of outer sleeve 450 includes an anti-rotation configuration, generally referred to as 606, to prevent relative rotation of outer sleeve 450 with respect to inner sleeve 530. In this regard, outer sleeve 450 defines a plurality of grooves, such as grooves 607a and 607b, for matingly receiving respective ones of a plurality of tabs 608a and 608b integrally formed with inner sleeve 530. Thus, as outer sleeve 450 is rotated, inner sleeve 530 is prevented from rotating with respect to outer sleeve 450 due to the engagement of tabs 608a and 608b in grooves 607a and 607b, respectively. As best seen in FIG. 8E, first nipple 480 is rotatable relative to outer sleeve 450. In this regard, first nipple 480 includes an annular flange 608c that is slidably received in an annular slot 608d formed in outer sleeve 450. In this manner, first nipple 480 is freely slidably rotatable with respect to outer sleeve 450. First nipple 480 is freely slidably rotatable in either of the directions indicated by curved arrows 608e or 608f. Moreover, as first nipple 480 freely slidably rotates in one direction, such as in the direction of arrow 608e, threaded internal cavity 500 will threadably engage the external threads of second nipple 560. It may be appreciated that as the threads of internal cavity 500 threadably engage the external threads of second nipple 560, first nipple 480 will abut first sleeve 450, such as at surface 608g. As first nipple 480 abuts first sleeve 450, first sleeve 450 will upwardly translate or ascend along a longitudinal axis thereof in a direction indicated by a vertical arrow 608h. First sleeve 450 will upwardly translate or ascend only in the direction of arrow 608h due to presence of anti-rotation configuration 606. As first sleeve 450 upwardly translates or ascends a predetermined amount, first holes 470 will be progressively closed, covered, shut-off and otherwise blocked by second ligaments 540 of inner sleeve 530. Moreover, it may be appreciated that, as first sleeve 450 upwardly translates or ascends the predetermined amount, second holes 550 will be progressively closed, covered, shut off and otherwise blocked by first ligaments 460 of outer sleeve 450. Progressively closing, covering, shutting off and otherwise blocking first holes 470 and second holes 550 in this manner variably reduces flow of the coolant through first holes 470 and second holes 550. It may be appreciated that rotation of first nipple 480 in an opposite direction, such as in the direction of curved arrow 608f, causes first holes 470 and second holes 550 to be progressively opened, uncovered, revealed and otherwise unblocked for variably increasing flow of coolant through first holes 470 and second holes 550. Therefore, referring to FIGS. 7, 8, 8A, 8B, 8C, 8D, 8E, 9 and 10, flow control in nuclear fission module 30 is achieved, at least in part, by use of two distinct components, which are outer sleeve 450 and inner sleeve 530, as described presently. As previously mentioned, inner sleeve 530 may be integrally formed with nuclear fission module 30 when nuclear fission module 30 is first fabricated. However, if desired, inner sleeve may be formed separately from nuclear fission module 30, but connectable thereto, rather than being integrally formed with nuclear fission module 30 when nuclear fission module 30 is first fabricated. Inner sleeve 530 defines the plurality of second holes 550 to allow passage of the coolant into nuclear fission module 30. Outer sleeve 450 slides on top of inner sleeve 530 and has the corresponding plurality of first holes 470. Outer sleeve 450 and inner sleeve 530 are concentric and holes 470/550 are always aligned to match along the radial or rotational axis. Coolant flow is controlled by the relative positions of inner sleeve 530 and outer sleeve 450 in the axial or vertical direction. In this regard, FIG. 8B shows flow regulator subassembly 430 in a fully open configuration to fully allow fluid flow into nuclear fission module 30 and FIG. 8C shows flow regulator subassembly 430 in a fully closed configuration to fully block fluid flow into nuclear fission module 30. The engagement of tabs 608a and 608b into respective ones of grooves 607a and 607b restricts rotation of outer sleeve 450 relative to inner sleeve 530, as previously mentioned. This feature allows axial sliding of outer sleeve 450 on inner sleeve 530, but no relative rotation between outer sleeve 450 and inner sleeve 530. Fine adjustment of coolant flow is achieved by the progressive axial sliding of outer sleeve 450 relative to inner sleeve 530. Thus, rotation of first nipple 480 in direction 608e progressively opens flow regulator subassembly 430 and rotation of first nipple 480 in direction 608f progressively closes flow regulator subassembly 430 for achieving fine adjustment of holes 470/550 and thus fine adjustment of coolant flow. As best seen in FIG. 11, there may be a plurality of smaller flow regulator subassemblies, such as flow regulator subassemblies 609a and 609b, assigned to a single nuclear fission module 30. Assignment of the plurality of smaller flow regulator subassemblies 609a and 609b to a single nuclear fission module 30 provides an alternative configuration for providing coolant flow to nuclear fission module 30. In addition, assignment of the plurality of smaller flow regulator subassemblies 609a and 609b to an individual or single nuclear fission module 30 provides a possibility of substantially controlling temperature distribution within distinct portions of an individual or single nuclear fission fuel module 30. This is possible because fluid flow through each of the smaller flow regulator subassemblies 609a and 609b can be individually controlled. Referring to FIGS. 12, 13, 14, 15, and 16, there is shown flow regulator subassembly 430 in operative condition to adjust or regulate coolant fluid flow into nuclear fission module 30. Together, flow regulator subassembly 430 and a carriage subassembly 610 define a flow control assembly, generally referred to as 615, as disclosed more fully hereinbelow. In other words, flow control assembly 615 comprises flow regulator subassembly 430 and carriage subassembly 610. In this regard, carriage subassembly 610 is disposed underneath reactor core 20, such as underneath core lower support plate 360, and is capable of being coupled to or is configured to be coupled to flow regulator subassembly 430 for adjusting flow regulator subassembly 430. Adjustment of flow regulator subassembly 430 variably controls coolant flow into nuclear fission module 30, as mentioned hereinabove. Moreover, carriage subassembly 610 is capable of carrying outer sleeve 450 to nuclear fission module 30, if desired. Referring to FIGS. 13, 14, 15, and 16, the configuration of carriage subassembly 610 will now be described. Carriage subassembly 610 comprises an elongate bridge 620 spanning reactor core 20 for supporting a plurality of vertically movable socket wrenches 630 thereon. Each of socket wrenches 630 has a shaft 700 and is movably disposed in a socket well 635 for reasons disclosed hereinbelow. Connected to opposing ends of bridge 620 are a first bridge mover 640a and a second bridge mover 640b, respectively. Bridge movers 640a and 640b may be operable by means of a gear arrangement (not shown) driven by a motor (also not shown). Such a motor may be located externally to reactor core 20 to avoid the corrosive effects and heat of the coolant, such as liquid sodium, circulating through reactor core 20. Each of bridge movers 640a and 640b includes at least one wheel 650a and 650b, respectively, for allowing bridge movers 640a and 640b to simultaneously move along respective ones of transversely spaced-apart and parallel tracks 660a and 660b. Bridge movers 640a and 640b are capable of moving or are configured to move bridge 620 along tracks 660a and 660b in either of the directions indicated by arrow 663. Connected to each of tracks 660a and 660b may be a track support 665a and 665b, respectively, for supporting tracks 660a and 660b thereon. Referring to FIGS. 13, 14, 15, 16, 17, 18, and 19, socket wrenches 630 are configured to be vertically reciprocated in socket well 635 into engagement and out of engagement with first nipple 480 of outer sleeve 450. In one embodiment of carriage assembly 610, rows of socket wrenches 630 are configured to be driven by a lead screw arrangement, generally referred to as 670. Lead screw arrangement 670 has a lead screw 680 configured to threadably engage external threads 690 surrounding shaft 700 belonging to each socket wrench 630. Lead screw 680 may be driven by a mechanical drive system 705 comprising a mechanical linkage 707 coupled to lead screw 680. When mechanical linkage 707 drives lead screw 680, the lead screw 680 will turn or rotate shaft 700 due to the threaded engagement of lead screw 680 and the external threads 690 surrounding shaft 700. Turning or rotating shaft 700 will turn or rotate first nipple 480 a like amount when an hexagonally shaped recess 700a in an upper portion of shaft 700 engages hexagonally shaped first nipple 480, as shown. Referring to FIGS. 15 and 16, the manner in which each shaft 700 is selectively raised and lowered will now be described. In this regard, an externally threaded, elongate mechanical linkage extension 708 engages a first gear wheel 709 for rotating first gear wheel 709 in either of the directions indicated by curved arrows 709a and 709b. For example, as mechanical linkage extension 708 translates in one of the directions indicated by a double-headed arrow 709c, first gear wheel 709 will rotate in a first direction, such as in the direction of arrow 709a. On the other hand, as mechanical linkage extension 708 translates in an opposite direction indicated by double-headed arrow 709c, first gear wheel 709 will rotate in a second direction, such as in the direction of arrow 709b. As first gear wheel 709 rotates, such as in the direction of arrow 709a, an externally threaded centermost first rod 709d will also rotate a like amount because the external threads of first rod 709d threadably engage internal threads (not shown) formed through the center of first gear wheel 709. A second gear wheel 709e has internal threads (not shown) formed through the center thereof for threadably engaging the external threads of first rod 709d. Thus, as first rod 709d is rotated by first gear wheel 709, second gear wheel 709e will translate along first rod 709d due to the threaded engagement of first rod 709d with second gear wheel 709e. Second gear wheel 709e will translate along first rod 709d until the location of a predetermined one of shafts 700 is reached. It may be appreciated that the pitch of the external threads or gear teeth of second gear wheel 709e is such as not to create an interference with the pitch of the external threads surrounding shafts 700 so that translation of second gear wheel 709e along first rod 709e may proceed unimpeded. A third gear wheel 709f is also provided for reasons described presently. In this regard, third gear wheel 709f is coupled to an elongate second rod 709g and to an elongate third rod 709h disposed on either side of and adjacent to centermost first rod 709d. Third gear wheel 709f is driven by the previously mentioned mechanical linkage extension 708, which is movable from a first position of engagement with first gear wheel 709 to a second position of engagement with third gear wheel 709f. As third gear wheel 709f rotates, second rod 709g and third rod 709h will rotate about the longitudinal axis of first rod 709d for rotating second gear wheel 709e about the longitudinal axis of first rod 709d. As second gear wheel 709e rotates, the external threads of second gear wheel 709e will threadably engage the external threads of shaft 700 for vertically translating shaft 700. In this manner, socket wrench 630 is translated either upwardly or downwardly. It should be appreciated that mechanical linkage extension 708 may be replaced by a fourth gear wheel (not shown) or by a pulley belt assembly (also not shown). Referring to FIGS. 17, 18 and 19, in another embodiment of carriage assembly 610, socket wrenches 630 are individually rotatable and axially translatable by means of respective ones of a plurality of hermetically sealed, reversible, first electric motors 710 that are coupled to shafts 700. First electric motors 710 are hermetically sealed and may be gas cooled to protect first electric motors 710 from the corrosive effects and heat of the coolant, which may be liquid sodium or liquid sodium mixture. First electric motors 710 are configured to selectively, vertically move shafts 700. Motors 710 are reversible in the sense that rotors of motors 710 may be operated in a first direction or a second direction opposite the first direction for moving shafts 700 either upwardly or downwardly, respectively. Operation of either mechanical drive system 705 or motors 710 is suitably controlled by means of a controller or control unit 720 coupled thereto. Each motor 710 may be a custom designed direct current servomotor, such as may be available from ARC Systems, Incorporated located in Hauppauge, N.Y., USA. Controller 720 may be a custom designed motor controller, such as may be available from Bodine Electric Company located in Chicago, Ill., USA. According to another embodiment, socket wrenches 630 are individually movable by means of a radio transmitter-receiver arrangement that includes a plurality of hermetically sealed, gas cooled, reversible, second electric motors 730 that are individually operable by receipt of a radio frequency signal transmitted by a radio transmitter 740. Second electric motors 730 are hermetically sealed and may be gas cooled to protect second electric motors 730 from the corrosive effects and heat of the sodium coolant. A power supply for second electric motor 730 may be a battery or other power supply device (not shown). Second electric motors 730, that are configured to receive such a radio signal, and radio transmitter 740 may be a custom designed motor and transmitter that may be available from Myostat Motion Control, Incorporated located in Ontario, Canada. According to another embodiment, socket wrenches 630 are individually movable by means of a fiber optic transmitter-receiver arrangement, generally referred to as 742, having a plurality of fiber optic cables 745 in order to operate the reversible motor arrangement by light transmission. As best seen in FIG. 14, flow control assembly 615, and thus flow regulator subassembly 430, are capable of being operated according to or in response to an operating parameter associated with nuclear fission module 30. In this regard, at least one sensor 750 may be disposed in nuclear fission module 30 to sense status of the operating parameter. The operating parameter sensed by sensor 750 may be current temperature in nuclear fission module 30. Alternatively, the operating parameter sensed by sensor 750 may have been a previous temperature in nuclear fission module 30. In order to sense temperature, sensor 750 may be a thermocouple device or temperature sensor that may be available from Thermocoax, Incorporated located in Alpharetta, Ga. U.S.A. As another alternative, the operating parameter sensed by sensor 750 may be neutron flux in nuclear fission module 30. In order to sense neutron flux, sensor 750 may be a “PN9EB20/25” neutron flux proportional counter detector or the like, such as may be available from Centronic House, Surrey, England. As another example, the operating parameter sensed by sensor 750 may be a characteristic isotope in nuclear fission module 30. The characteristic isotope may be a fission product, an activated isotope, a transmuted product produced by breeding or other characteristic isotope. Another example is that the operating parameter sensed by sensor 750 may be neutron fluence in nuclear fission module 30. As well known in the art, neutron fluence is defined as the neutron flux integrated over a certain time period and represents the number of neutrons per unit area that passed during that time. As yet another example, the operating parameter sensed by sensor 750 may be fission module pressure, which may be a dynamic fluid pressure of approximately 10 bars (i.e., approximately 145 psi) for an exemplary sodium cooled reactor or approximately 138 bars (i.e., approximately 2000 psi) for an exemplary pressurized “light” water cooled reactor during normal operation. Alternatively, fission module pressure that is sensed by sensor 750 may be a static fluid pressure or a fission product pressure. In order to sense either dynamic or static fission module pressure, sensor 750 may be a custom designed pressure detector that may be available from Kaman Measuring Systems, Incorporated located in Colorado Springs, Colo. U.S.A. As another alternative, sensor 750 may be a suitable flow meter such as a “BLANCETT 1100 TURBINE FLOW METER”, that may be available from Instrumart, Incorporated located in Williston, Vt. U.S.A. In addition, the operating parameter sensed by sensor 750 may be determined by a suitable computer-based algorithm. A variety of algorithms can be implemented, including those such as the ideal gas law, PV=nRT, or known algorithms that produce signals indicative of pressure or temperature from direct or indirect measurement of other properties, such as flows, temperatures, electrical properties, or other. According to yet another example, the operating parameter may be operator initiated action. That is, flow regulator subassembly 430 is capable of being modified in response to any suitable operating parameter determined by a human operator. Further, flow regulator subassembly 430 is capable of being modified in response to an operating parameter determined by a suitable feedback control. Also, flow regulator subassembly 430 is capable of being modified in response to an operating parameter determined by an automated control system. Moreover, flow regulator subassembly 430 is capable of being modified in response to a change in decay heat. In this regard, decay heat decreases in the “tail” of burn wave 290 (see FIG. 4). Detection of the presence of the tail of burn wave 290 is used to decrease coolant flow rate over time to account for this decrease in decay heat found in the tail of burn wave 290. This is particularly the case when nuclear fission module 30 resides behind burn wave 290. In this case, flow regulator subassembly 430 accounts for changes in decay heat output of nuclear fission module 30 as the distance of nuclear fission module 30 from burn wave 290 changes. Sensing status of such operating parameters can facilitate suitable control and modification of flow control assembly 615 operation and thus suitable control and modification of temperature in reactor core 20. Referring to FIGS. 14, 15, 17, 18 and 19, it should be understood from the description hereinabove that flow regulator subassembly 430 is reconfigurable according to a predetermined input to controllers 720 and 740, so that controllers 720 and 740 in combination with flow regulator subassembly 430 suitably control fluid flow. That is, the predetermined input to controllers 720 and 740 is a signal produced by the previously mentioned sensor 750. For example, the predetermined input to controllers 720 and 740 may be a signal produced by the previously mentioned thermocouple or temperature sensor. Alternatively, the predetermined input to controllers 720 and 740 may be a signal produced by the previously mentioned fluid flow meter. As another alternative, the predetermined input to controllers 720 and 740 may be a signal produced by the previously mentioned neutron flux detector. As another example, signals received by controllers 720 and 740 may have been processed by reactor control systems (not shown). For example, the signals produced by such a reactor control system may come from a meter or detector and get processed either by a computer or operator in a reactor control room and then go out to carriage subassembly 610, so as to move bridge 620 and socket wrenches 630 to operate flow regulator subassembly 430. Referring to FIGS. 4, 10, and 14, it may be understood by a person of skill in the art that, based on the teachings herein, flow control assembly 615 can be capable of controlling or modulating flow of the coolant according to when traveling burn wave 290 arrives at and/or departs from nuclear fission module 30. Also, flow control assembly 615 is capable of controlling or modulating flow of the coolant according to when traveling burn wave 290 is proximate to or in the vicinity of nuclear fission module 30. Flow control assembly 615 is also capable of controlling or modulating flow of the coolant according to the previously mentioned width “x” of burn wave 290. Arrival and departure of burn wave 290, as burn wave 290 travels through nuclear fission module 30, is detected by sensing any of the previously mentioned operating parameters. For example, flow control assembly 615 is capable of controlling or modulating flow of the coolant according to heat generation rate sensed in nuclear fission module 30. It should be apparent to those skilled in the art that, in some cases, an input signal alone may control modification of flow control assembly 615 and the associated fluid flow in nuclear fission module 30. Referring to FIGS. 14 and 15, and as previously mentioned, flow control assembly 615 is operated to provide variable fluid flow to a selected one of nuclear fission modules 30. Nuclear fission module 30 is selected on the basis of the desired value for the operating parameter (e.g., temperature) in nuclear fission module 30 compared to the actual value of the operating parameter that is sensed in nuclear fission module 30. As described in more detail presently, fluid flow to nuclear fission module 30 is adjusted to bring the actual value for the operating parameter into substantial agreement with the desired value for the operating parameter. To achieve this result, bridge 620 that belongs to carriage subassembly 630 is caused to travel along tracks 660a and 660b by simultaneously actuating bridge movers 640a and 640b. As bridge 620 travels along tracks 660a and 660b, the bridge 620 will travel underneath core lower support plate 360. Bridge 620 eventually stops its travel at a predetermined location underneath core lower support plate 360 based on the actual value of the operating parameter sensed by sensors 750 in nuclear fission module 30 compared to the desired value of the operating parameter for nuclear fission module 30, as described in more fully presently. Activation and extent of travel of bridge movers 640a and 640b may be controlled by a suitable controller, such as by controllers 720 or 740. In this regard, controllers 720 or 740 will stop the travel of bridge 620 based on location of the selected one of the plurality of nuclear fission modules 30. As mentioned hereinabove, the nuclear fission module 30 to be adjusted can be selected on the basis of whether or not there is substantial agreement between the actual value of the operating parameter sensed by sensor 750 and the value of the operating parameter desired for nuclear fission module 30. Next, a selected one of the plurality of hexagonal socket wrenches 630 is caused to move vertically upwardly to matingly engage hexagonal first nipple 480. After engagement of socket wrench 630 with first nipple 480, shaft 700 is caused to rotate in order to rotate socket wrench 630. Shaft 700 is caused to rotate either by means of the previously mentioned lead screw arrangement 670, first electric motors 710, or second electric motors 730 that are coupled to controllers 720 or 740. Referring to FIGS. 7, 8, 8A, 8B, 8C, 8D, 8E, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18 and 19, after engagement with first nipple 480, rotation of socket wrench 630 in a first direction causes first or outer sleeve 450 to rotate in the same first direction. As outer sleeve 450 rotates, outer sleeve 450 will axially slidably ascend along the exterior of inner sleeve 530 due to the threaded engagement of first nipple 480 belonging to outer sleeve 450 and second nipple 560 belonging to inner sleeve 530. As outer sleeve 450 slides upwardly along inner sleeve 530, first ligaments 460 of outer sleeve 450 will progressively close, cover, shut-off and otherwise block second holes 550 of inner sleeve 530 and second ligaments 540 of inner sleeve 530 will simultaneously progressively close, cover, shut-off and otherwise block first holes 470 of outer sleeve 530. Progressively closing, covering, shutting-off and otherwise blocking first holes 470 and second holes 550 variably reduces flow of the coolant through first holes 470 and second holes 550. In this case, second holes 550 and first holes 470 may have been previously aligned for allowing full flow of coolant therethrough. Alternatively, second holes 550 and first holes 470 may have been previously partially aligned for allowing partial flow of coolant therethrough. Referring again to FIGS. 7, 8, 8A, 8B, 8C, 8D, 8E, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18 and 19, after engagement with first nipple 480, rotation of socket wrench 630 in a second direction opposite the first direction causes first or outer sleeve 450 to rotate in the second direction. As outer sleeve 450 rotates, outer sleeve 450 will axially slidably descend along the exterior of inner sleeve 530 due to the threaded engagement of first nipple 480 belonging to outer sleeve 450 and second nipple 560 belonging to inner sleeve 530. As outer sleeve 450 slides downwardly along inner sleeve 530, first ligaments 460 of outer sleeve 450 will progressively open, uncover, reveal and otherwise unblock second holes 550 of inner sleeve 530 and second ligaments 540 of inner sleeve 530 will simultaneously progressively open, uncover, reveal and otherwise unblock first holes 470 of outer sleeve 530. Progressively opening, uncovering, revealing and otherwise unblocking first holes 470 and second holes 550 variably increases flow of the coolant through first holes 470 and second holes 550. In this case, second holes 550 and first holes 470 may have been previously misaligned for restricting or disallowing flow of coolant therethrough. Alternatively, second holes 550 and first holes 470 may have been previously partially misaligned for partially restricting or partially disallowing flow of coolant therethrough. Thus, use of flow control assembly 615, which includes flow regulator subassembly 430 and carriage subassembly 610, achieves variable coolant flow on a module-by-module (i.e., fuel assembly-by-fuel assembly) basis. This allows coolant flow to be varied across reactor core 20 according to the location of burn wave 290 or the non-uniform temperature distribution in reactor core 20. Illustrative Methods Illustrative methods associated with exemplary embodiments of a nuclear fission reactor and flow control assembly will now be described. Referring to FIGS. 20A-20S, illustrative methods are provided for operating a nuclear fission reactor. Turning now to FIG. 20A, an illustrative method 760 of operating a nuclear fission reactor starts at a block 770. At a block 780, the method comprises producing at least a portion of a traveling burn wave at a location relative to a nuclear fission module. At a block 790, a flow control assembly is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. The method stops at a block 800. In FIG. 20B, an illustrative method 810 of operating a nuclear fission reactor starts at a block 820. At a block 830, at least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module. At a block 840, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 850, a flow regulator subassembly is operated. The method stops at a block 860. In FIG. 20C, another illustrative method 870 of operating a nuclear fission reactor starts at a block 880. At a block 890, at least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module. At a block 900, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. A flow regulator subassembly is operated at a block 910. At a block 920, the flow regulator subassembly is operated according to an operating parameter associated with the nuclear fission module. The method stops at a block 930. In FIG. 20D, a further illustrative method 940 of operating a nuclear fission reactor starts at a block 950. At a block 960, at least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module. At a block 970, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. A flow regulator subassembly is operated at a block 980. At a block 990, the flow regulator subassembly is modified in response to an operating parameter associated with the nuclear fission module. The method stops at a block 1000. In FIG. 20E, another illustrative method 1010 of operating a nuclear fission reactor starts at a block 1020. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1030. At a block 1040, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. A flow regulator subassembly is operated at a block 1050. At a block 1060, the flow regulator subassembly is reconfigured according to a predetermined input to the flow regulator subassembly. The method stops at a block 1070. In FIG. 20F, still another illustrative method 1080 of operating a nuclear fission reactor starts at a block 1090. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1100. At a block 1110, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1120, a flow regulator subassembly is operated. At a block 1130, a controllable flow resistance is achieved. The method stops at a block 1140. In FIG. 20G, an illustrative method 1150 of operating a nuclear fission reactor starts at a block 1160. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1170. At a block 1180, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1190, a flow regulator subassembly is operated. At a block 1200, a second sleeve is inserted into a first sleeve, the first sleeve having a first hole and the second sleeve having a second hole alignable with the first hole. The method stops at a block 1210. In FIG. 20H, another illustrative method 1220 of operating a nuclear fission reactor starts at a block 1230. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1240. At a block 1250 a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1260, a flow regulator subassembly is operated. At a block 1270 a carriage subassembly that is coupled to the flow regulator subassembly is operated. The method stops at a block 1280. In FIG. 20I, an additional illustrative method 1290 of operating a nuclear fission reactor starts at a block 1300. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1310. At a block 1320, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1330, a flow regulator subassembly is operated. At a block 1340, a temperature sensor is coupled to the nuclear fission module and the flow regulator subassembly. The method stops at a block 1350. In FIG. 20J, a further illustrative method 1360 of operating a nuclear fission reactor starts at a block 1370. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1380. At a block 1390, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1400, flow of the fluid is controlled in response to the location relative to the location of the nuclear fission module by operating the flow control assembly according to when the burn wave arrives at the location relative to the location of the nuclear fission module. The method stops at a block 1410. In FIG. 20K, still another illustrative method 1420 of operating a nuclear fission reactor starts at a block 1430. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1440. At a block 1450, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1460, flow of the fluid is controlled in response to the location relative to the nuclear fission module by operating the flow control assembly according to when the burn wave departs from the location relative to the nuclear fission module. The method stops at a block 1470. In FIG. 20L, another illustrative method 1480 of operating a nuclear fission reactor starts at a block 1490. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1500. At a block 1510, a flow control assembly that is coupled to the nuclear fission module is modulated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1520, flow of the fluid is controlled in response to the location relative to the nuclear fission module by operating the flow control assembly according to when the burn wave is proximate to the location relative to the nuclear fission module. The method stops at a block 1530. In FIG. 20M, an illustrative method 1540 of operating a nuclear fission reactor starts at a block 1550. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1560. At a block 1570, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1580, flow of the fluid is controlled according to a width of the burn wave. The method stops at a block 1590. In FIG. 20N, an illustrative method 1600 of operating a nuclear fission reactor starts at a block 1610. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1620. At a block 1630, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1640, flow of the fluid is controlled by operating the flow control assembly according to a heat generation rate in the nuclear fission module. The method stops at a block 1650. In FIG. 20O, an illustrative method 1660 of operating a nuclear fission reactor starts at a block 1670. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1680. At a block 1690, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1700, flow of a fluid is controlled by operating the flow control assembly according to a temperature in the nuclear fission module. The method stops at a block 1710. In FIG. 20P, an illustrative method 1720 of operating a nuclear fission reactor starts at a block 1730. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1740. At a block 1750, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1760, flow of the fluid in controlled by operating the flow control assembly according to a neutron flux in the nuclear fission module. The method stops at a block 1770. In FIG. 20Q, an illustrative method 1780 of operating a nuclear fission reactor starts at a block 1790. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1800. At a block 1810, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1820, at least a portion of the traveling burn wave is produced at a location relative to a nuclear fission fuel assembly. The method stops at a block 1830. In FIG. 20R, an illustrative method 1840 of operating a nuclear fission reactor starts at a block 1850. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1860. At a block 1870, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1880, at least a portion of the traveling burn wave is produced at a location relative to a fertile nuclear breeding assembly. The method stops at a block 1890. In FIG. 20S, an illustrative method 1900 of operating a nuclear fission reactor starts at a block 1910. At least a portion of a traveling burn wave is produced at a location relative to a nuclear fission module at a block 1920. At a block 1930, a flow control assembly that is coupled to the nuclear fission module is operated to modulate flow of a fluid in response to the location relative to the nuclear fission module. At a block 1940, at least a portion of the traveling burn wave is produced at a location relative to a neutron reflector assembly. The method stops at a block 1950. Referring to FIGS. 21A-21H, illustrative methods are provided for assembling a flow control assembly for use in a nuclear fission reactor. Turning now to FIG. 21A, an illustrative method 1960 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 1970. At a block 1980, a flow regulator subassembly is received. The method stops at a block 1990. In FIG. 21B, another illustrative method 2000 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 2010. At a block 2020, a carriage subassembly is received. The method stops at a block 2030. In FIG. 21C, another illustrative method 2040 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 2050. A flow regulator subassembly is received at a block 2060. A first sleeve having a first hole is received at a block 2070. At a block 2080, a second sleeve is inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole, and the first sleeve being configured to rotate for rotating the first hole into alignment with the second hole. At a block 2090, a carriage subassembly is coupled to the flow regulator subassembly. The method stops at a block 2100. In FIG. 21D, yet another illustrative method 2110 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 2120. A flow regulator subassembly is received at a block 2130. At a block 2140, a first sleeve is received having a first hole. At a block 2150, a second sleeve is inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole. At a block 2160, a carriage subassembly is coupled to the flow regulator subassembly. At a block 2170, the carriage subassembly is coupled to the flow regulator subassembly so that the carriage subassembly carries the flow regulator subassembly to the fuel assembly. The method stops at a block 2180. In FIG. 21E, a further illustrative method 2190 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 2200. A flow regulator subassembly is received at a block 2210. At a block 2220, a first sleeve is received having a first hole. At a block 2230, a second sleeve is inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole. At a block 2240, a carriage subassembly is coupled to the flow regulator subassembly. At a block 2250 the carriage subassembly is coupled to the flow regulator subassembly so that the carriage subassembly is driven by a lead screw arrangement. The method stops at a block 2260. In FIG. 21F, an illustrative method 2270 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 2280. A flow regulator subassembly is received at a block 2290. A first sleeve having a first hole is received at a block 2300. At a block 2310, a second sleeve is inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole, and the first sleeve being configured to rotate for rotating the first hole into alignment with the second hole. At a block 2320, a carriage subassembly is coupled to the flow regulator subassembly. At a block 2330, the carriage subassembly is coupled so that the carriage subassembly is driven by a reversible motor arrangement. The method stops at a block 2340. In FIG. 21G, an illustrative method 2350 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 2360. A flow regulator subassembly is received at a block 2370. A first sleeve having a first hole is received at a block 2380. At a block 2390, a second sleeve is inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole, and the first sleeve being configured to rotate for rotating the first hole into alignment with the second hole. At a block 2400, a carriage subassembly is coupled to the flow regulator subassembly. At a block 2410, the carriage subassembly is coupled so that the carriage subassembly is at least partially controlled by a radio transmitter-receiver arrangement operating the reversible motor arrangement. The method stops at a block 2415. In FIG. 21H, an illustrative method 2420 of assembling a flow control assembly for use in a nuclear fission reactor starts at a block 2430. A flow regulator subassembly is received at a block 2440. A first sleeve having a first hole is received at a block 2450. At a block 2460, a second sleeve is inserted into the first sleeve, the second sleeve having a second hole alignable with the first hole, and the first sleeve being configured to rotate for rotating the first hole into alignment with the second hole. At a block 2470, a carriage subassembly is coupled to the flow regulator subassembly. At a block 2480, the carriage subassembly is coupled so that the carriage subassembly is at least partially controlled by a fiber optic transmitter-receiver arrangement operating the reversible motor arrangement. The method stops at a block 2490. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken as limiting. Moreover, those persons skilled in the art will appreciate that the foregoing specific exemplary processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those persons skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. Therefore, what are provided are a nuclear fission reactor, flow control assembly, methods therefor and a flow control assembly system. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. For example, a horizontally disposed orifice plate may be substituted for the flow regulator subassembly, the orifice plate having a plurality of orifices therethrough. A plurality of individually actuatable shutters would be associated with respective ones of the orifices, the shutters being capable of progressively closing and opening the orifices for regulating or modulating flow of coolant to the nuclear fission module. In addition, it may be appreciated from the teachings herein that, unlike the devices disclosed in the prior art patents cited hereinabove, the flow control assembly and system of the present disclosure dynamically change the amount of the fluid flow, avoids reliance on different and precisely constituted neutron-induced growth properties of structural materials for controlling fluid flow, and can be dynamically varied during reactor operation, as needed. Moreover, the various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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abstract | Aspects of the invention relate to several methods to deposit and regenerate target materials in neutron generators and similar nuclear reaction devices. In situ deposition and regeneration of a target material reduces tube degradation of the nuclear reaction device and covers impurities on the surface of the target material at the target location. Further aspects of the invention include a method of designing a target to generate neutrons at a high efficiency rate and at a selected neutron energy from a neutron energy spectrum. |
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claims | 1. A lithographic apparatus comprising:a radiation system constructed to provide a beam of radiation from radiation emitted by a radiation source, the radiation system comprisinga contaminant trap configured to trap material emanating from the radiation source, the contaminant trap comprising a static part and a rotating part, the rotating part comprising a contaminant engaging surface arranged in the path of the radiation beam that receives the material emanating from the radiation source during propagation of the radiation beam in the radiation system, anda liquid tin cooling system constructed to cool the contaminant trap with liquid tin;an illumination system configured to condition the radiation beam;a support constructed to support a patterning device, the patterning device being configured to impart the radiation beam with a pattern in its cross-section to form a patterned radiation beam;a substrate table constructed to hold a substrate; anda projection system configured to project the patterned radiation beam onto a target portion of the substrate. 2. An apparatus according to claim 1, wherein the liquid tin cooling system is arranged to condition the temperature of the contaminant trap. 3. An apparatus according to claim 1, wherein the liquid tin cooling system comprises a closed liquid tin circuit arranged inside the static part of the contaminant trap. 4. An apparatus according to claim 3, wherein the liquid tin cooling system comprises a liquid tin supply channel inside the static part of the contaminant trap, the supply channel extending to the rotating part of the contaminant trap for supplying the liquid tin towards an external surface of the rotating part, the external surface comprising the contaminant engaging surface. 5. An apparatus according to claim 4, wherein the contaminant engaging surface is disposed on a foil, and wherein the liquid tin cooling system further comprises a return path along a leading edge of the foil comprised in the rotating part of the contaminant trap. 6. An apparatus according to claim 4, wherein the contaminant engaging surface is disposed on a foil, and wherein the liquid tin cooling system further comprises a return path embedded in the foil comprised in the rotating part of the contaminant trap. 7. An apparatus according to claim 1, wherein the liquid tin cooling system comprises a semi-open liquid tin circuit constructed to directly cool the rotating part of the contaminant trap. 8. An apparatus according to claim 1, wherein the liquid tin cooling system comprises an exterior supply channel having a spray end arranged to spray the rotating part of the contaminant trap. 9. An apparatus according to claim 8, wherein the contaminant engaging surface is disposed on a foil, and wherein the spray end is arranged near the foil comprised in the rotating part of the contaminant trap. 10. An apparatus according to claim 1, further comprising a gas inlet and a heating element, both arranged near the contaminant trap. 11. An apparatus according to claim 1, further comprising a radical or plasma generating unit. 12. An apparatus according to claim 1, wherein an exterior surface of the contaminant trap comprises a top layer having a low oxidation rate. 13. An apparatus according to claim 12, wherein the top layer comprises gold. 14. An apparatus according to claim 1, wherein the contaminant engaging surface is disposed on a foil, the foil having a segment that is substantially porous, and wherein a liquid tin supply channel ends in the porous segment of the foil. 15. A radiation system constructed to provide a beam of radiation from radiation emitted by a radiation source, the radiation system comprising:a contaminant trap configured to trap material emanating from the radiation source, the contaminant trap comprising a static part and a rotating part, the rotating part comprising a contaminant engaging surface arranged in the path of the radiation beam that receives the material emanating from the radiation source during propagation of the radiation beam in the radiation system; anda liquid tin cooling system constructed to cool the contaminant trap with liquid tin. 16. A device manufacturing method comprising:trapping material emanating from a radiation source using a contaminant trap comprising a static part and a rotating part, the rotating part comprising a contaminant engaging surface by arranging the surface in a radiation beam emitted by the radiation source;cooling the contaminant trap with liquid tin;conditioning the radiation beam;imparting the radiation beam with a pattern in its cross-section using a patterning device to form a patterned radiation beam; andprojecting the patterned radiation beam onto a target portion of a substrate. 17. A method according to claim 16, further comprising collecting liquid tin that is dropped into a chamber in which the contaminant trap is arranged, and reusing collected liquid tin to cool the contaminant trap. 18. A method according to claim 16, further comprising regenerating tin liquid in a cooling circuit of a liquid tin cooling system. 19. A method according to claim 16, further comprising pre-treating a contaminant trap exterior surface for improved surface wetting characteristics. 20. A method according to claim 19, wherein said pre-treating comprises heating the exterior surface. 21. A method according to claim 20, wherein said heating the exterior surface is performed in a hydrogen atmosphere. 22. A method according to claim 19, wherein said pre-treating comprises introducing radicals or a plasma near the contaminant trap. 23. A method according to claim 22, wherein said plasma is an oxygen plasma. 24. A method according to claim 19, wherein said pre-treating comprises coating the exterior surface with a top layer having a low oxidation rate. 25. A method according to claim 24, wherein said top layer comprises gold. 26. A method according to claim 24, wherein the top layer has a solubility in liquid tin of less than about 0.05%. 27. A method according to claim 26, wherein the top layer has a solubility in liquid tin of less than about 0.005%. 28. A radiation generating method comprising:trapping material emanating from a radiation source using a contaminant trap comprising a static part and a rotating part, the rotating part comprising a contaminant engaging surface by arranging the surface in a radiation beam emitted by the radiation source; andcooling the contaminant trap with liquid tin. |
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description | This application is a Division of U.S. patent application Ser. No. 14/666,890, filed on Mar. 24, 2015 which claims the benefit of the filing date of U.S. Provisional Application No. 61/972,551, filed Mar. 31, 2014, the content of both of which is incorporated by reference in their entirety. Radiation particle power converters can convert energy from a radioactive source that emits high-energy electrons, e.g., beta particles, into electrical energy. The power converter can convert the energy of the high-energy electrons to electrical energy, i.e., current, by collecting electron-hole pairs that are formed by the high-energy electrons that are incident upon a semiconductor material of the power converter. One such power converter includes a radiation-emitting radioisotope and a plurality of semiconductor substrates. Each of the plurality of semiconductor substrates includes a junction for converting nuclear radiation particles to electrical energy, e.g., a p-n junction. The junction collects electron-hole pairs that are created within the semiconductor material as a result of interaction with the nuclear radiation particles. Specifically, when a radiation particle of sufficient energy impacts the semiconductor material, electrons in the semiconductor material are excited into a conduction band of the semiconductor material, thereby creating electron-hole pairs. Electrons formed on an n side of a p-n junction are generally prevented from crossing the p-n junction due to the electric field that is created in a depletion zone, while the corresponding holes are swept across the p-n junction by the electric field. Electrons formed on the p side of the p-n junction are swept across the junction by the electric field while the corresponding holes are prevented from crossing the junction by the electric field. When the semiconductor material is connected to a load, electrons formed on the n side of the junction or are swept across the junction from the p side are further swept via an anode through a circuit connected to the power converter. The electrons that flow through the circuit then flow into the p side via a cathode, where they can recombine with holes formed as part of the original electron-hole pairs. In general, the present disclosure provides several embodiments of a nuclear radiation particle power converter. In one aspect, the present disclosure provides one embodiment of a power converter that includes first and second electrodes; a three-dimensional current collector disposed between the first and second electrodes and electrically coupled to the first electrode; and a charge carrier separator disposed on at least a portion of a surface of the three-dimensional current collector. The power converter further includes a hole conductor layer disposed on at least a portion of the charge carrier separator and electrically coupled to the second electrode; and nuclear radiation-emitting material disposed such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon the charge carrier separator. In another aspect, the present disclosure provides one embodiment of a method that includes forming a three-dimensional current collector between first and second electrodes, where the three-dimensional current collector is electrically coupled to the first electrode; forming a charge carrier separator on at least a portion of a surface of the three-dimensional current collector; and forming a hole conductor layer on at least a portion of the charge carrier separator, where the hole conductor layer is electrically coupled to the second electrode. The method further includes forming nuclear radiation-emitting material proximate the charge carrier separator such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon the charge carrier separator. These and other aspects of the present disclosure will be apparent from the detailed description below. In no event, however, should the above summaries be construed as limitations on the claimed subject matter, which subject matter is defined solely by the attached claims, as may be amended during prosecution. In general, the present disclosure provides several embodiments of a nuclear radiation particle power converter. In one or more embodiments, the power converter can include nuclear radiation-emitting material that emits nuclear radiation particles whose energy can be converted to electrical energy, i.e., current, by the power converter. This nuclear radiation-emitting material can, e.g., emit alpha particles (a nuclear particle that includes two protons and two neutrons, e.g., a nucleus of a helium atom) via alpha decay, or beta particles (a high-energy electron) via beta decay. Although the power converters described in the present disclosure may be configured to convert any suitable nuclear radiation particle to electrical energy, the disclosure will focus on power converters for converting energy from beta particles into electrical energy, generally referred to as “betavoltaic power converters.” Energy from the beta particles may be converted by the power converter using a charge carrier separator (e.g., quantum dots) that in one or more embodiments facilitates the separation of electron-hole pairs created by high energy electrons that are generated from the nuclear radiation-emitting material that decays by beta particle emission. In one or more embodiments, the power converter can include a charge carrier separator disposed on at least a portion of a surface of a three-dimensional current collector that is disposed between first and second electrodes. A hole conductor layer can be disposed on at least a portion of the charge carrier separator. And in one or more embodiments, nuclear radiation-emitting material can be disposed such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon the charge carrier separator. For example, FIGS. 1A-B are schematic cross-section views of one embodiment of a power converter 10, where FIG. 1B is an enlarged view of a portion of the power converter 10 of FIG. 1A. The power converter 10 includes a first electrode 12 and a second electrode 20. A three-dimensional current collector 30 is disposed between the first and second electrodes 12, 20 and, in one or more embodiments, electrically coupled to the first electrode. The power converter 10 can also include a charge carrier separator 40 (FIG. 1B) disposed, in one or more embodiments, on at least a portion of a surface 32 of the three-dimensional current collector 30. A hole conductor layer 50 can be disposed, e.g., on at least a portion of the charge carrier separator 40. In one or more embodiments, the hole conductor layer 50 is electrically coupled to the second electrode 20. Power converter 10 can also include nuclear radiation-emitting material 60 (FIG. 1B). In one or more embodiments, the nuclear radiation-emitting material 60 can be disposed in any suitable location relative to the charge carrier separator 40 such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon the charge carrier separator. For example, in one or more embodiments, the nuclear radiation material 60 can be disposed within at least one of the three-dimensional current collector 30, charge carrier separator 40, and the hole conductor layer 50. In one or more embodiments, the nuclear radiation material 60 can be formed such that a counter electrode 70 is positioned between the nuclear radiation material and the hole conductor layer 50 as is further described herein. The first and second electrodes 12, 20 can take any suitable shape or shapes and include any suitable materials, e.g., metals, conductive polymers, other suitable electrical conductors, or combinations thereof. In one or more embodiments, the materials for the first electrode 12 and the three-dimensional current collector 30 can be selected such that the work functions of the materials prevent schottky barrier semiconductor/conductor interfaces from being formed. In one or more embodiments, the materials of the first electrode 12 and the three-dimensional current collector 30 can be selected and configured such that they form a schottky barrier or an ohmic contact. In one or more embodiments, the materials for the second electrode 20 and the hole conductor layer 50 and/or the counter electrode 70 can be selected such that the work functions of the materials prevent schottky barrier semiconductor/conductor interfaces from being formed. In one or more embodiments, the materials for the second electrode 20 and the hole conductor layer 50 and/or the counter electrode 70 can be selected and configured such that they form a schottky barrier or alternately form an ohmic contact. In one or more embodiments, the first and second electrodes 12, 20 can electrically couple the power converter 10 to other devices using any suitable techniques. Further, either of the first or second electrodes 12, 20 can be positive or negative depending upon the application in which the power converter 10 is utilized. Disposed between the first and second electrodes 12, 20 is the three-dimensional current collector 30. As used herein, a three-dimensional current collector is a structure or device that includes one or more surfaces that provide the collector with an extent in three dimensions and that is configured to receive or transmit a current on or through the one or more surfaces of the collector. In one or more embodiments, the three-dimensional current collector 30 can be constructed such that it provides a heterostructure with the charge carrier separator 40. In one or more embodiments, the three-dimensional current collector 30 is electrically coupled to the first electrode 12. In one or more embodiments, the three-dimensional current collector 30 is electrically coupled to the second electrode 20. In one or more embodiments, the three-dimensional current collector 30 can be formed on the first electrode 12. In one or more embodiments, the collector 30 can be formed on the second electrode 20. In one or more embodiments, one or more intervening layers can be disposed between the three-dimensional current collector 30 and one or both of the first and second electrodes, 12, 20 that provides added functionality, e.g., a conductive layer, an adhesion layer, etc. The three-dimensional current collector 30 can include any suitable material or materials. In one or more embodiments, the collector 30 can include a porous material. Further, in one or more embodiments, the collector 30 can include a high bandgap semiconductor material, e.g., TiO2, SnO2, ZnO, WO3, Nb2O5, Ta2O5, BaTiO3, SrTiO3, ZnTiO3, CuTiO3, and combinations thereof. In one or more embodiments, the collector 30 can include graphite, graphene, C60, C70 and combinations thereof. Further, in one or more embodiments, the collector 30 can include a porous sintered TiO2 material, a porous Ti/TiO2 material, etc. The three-dimensional current collector 30 can take any suitable shape or shapes. In one or more embodiments, the collector 30 can include nanorods, nanotubes, nanowires, nanocrystalline structures, metal foam, graphene foam, and combinations thereof. In one or more embodiments, the collector 30 can include a lithographically-patterned structure or other ordered structure such as those that can be manufactured using, e.g., 3D printing techniques. And the surface 32 of the collector 30 can take any suitable shape or shapes. In general, the three-dimensional current collector 30 can, in one or more embodiments, maximize a surface area of surface 32 for any given volume. The power converter 10 can also include a charge carrier separator 40. The separator 40 can be disposed in any suitable location. For example, in the embodiment illustrated in FIGS. 1A-B, the separator 40 is disposed on at least a portion of the surface 32 of the three-dimensional current collector 30. The charge carrier separator 40 can be disposed on any suitable portion of the surface 32 of the three-dimensional current collector, e.g., the entire surface 32. In one or more embodiments, the charge carrier separator 40 can be disposed within the hole conductor layer 50 such that it provides a three-dimensional structure but still is in contact with the three-dimensional current collector 30. In one or more embodiments, the charge carrier separator 40 and the three-dimensional current collector 30 can form a heterostructure. The charge carrier separator 40 can include any suitable material or materials. In one or more embodiments, the separator 40 can include an oxide. In one or more embodiments, the separator 40 can be an oxide of the material used to form the three-dimensional current collector 30, e.g., TiO2. In one or more embodiments, the charge carrier separator 40 can include nanocrystals. As used herein, the term “nanocrystal” refers to nanostructures that are substantially monocrystalline. A nanocrystal has at least one region or characteristic dimension with a dimension of less than about 500 nm, and down to on the order of less than about 1 nm. The terms “nanocrystal,” “nanodot,” “dot,” and “quantum dot” are readily understood by the ordinarily skilled artisan to represent like structures and are used herein interchangeably. The present disclosure also encompasses the use of polycrystalline or amorphous nanocrystals. Typically, the region of characteristic dimension will be along the smallest axis of the structure. Nanocrystals can be substantially homogenous in material properties, or in some embodiments, can be heterogeneous. The nanocrystals can be produced using any suitable technique or techniques. The nanocrystals for use in the present disclosure can also include any suitable material or materials, including an inorganic material, and more suitably an inorganic conductive or semiconductive material. Suitable semiconductor materials can include any type of semiconductor, including group II-VI, group III-V, group IV-VI and group IV semiconductors. Suitable semiconductor materials can include, but are not limited to, Si, Ge, Sn, Se, Te, B, C (including diamond), P, BN, BP, BAs, AlN, AlP, AlAs, AlSb, GaN, GaP, GaAs, GaSb, InN, InP, InAs, InSb, AlN, AlP, AlAs, AlSb, GaN, GaP, GaAs, GaSb, ZnO, ZnS, ZnSe, ZnTe, CdS, CdSe, CdTe, HgS, HgSe, HgTe, BeS, BeSe, BeTe, MgS, MgSe, GeS, GeSe, GeTe, SnS, SnSe, SnTe, PbO, PbS, PbSe, PbTe, CuF, CuCl, CuBr, CuI, Si3N4, Ge3N4, Al2O3, (Al, Ga, In)2 (S, Se, Te)3, Al2CO, and combinations thereof. In one or more embodiments, the semiconductor nanocrystals can include a dopant such as a p-type dopant or an n-type dopant. The nanocrystals useful in the present disclosure can also include II-VI or III-V semiconductors. Examples of II-VI or III-V semiconductor nanocrystals include any combination of an element from Group II, such as Zn, Cd and Hg, with any element from Group VI, such as S, Se, Te, Po, of the Periodic Table; and any combination of an element from Group III, such as B, Al, Ga, In, and Tl, with any element from Group V, such as N, P, As, Sb and Bi, of the Periodic Table. In one or more embodiments, the nanocrystals can include core-shell structures that are obtained by adding organometallic precursors containing the shell materials to a reaction mixture containing the core nanocrystal. In this case, rather than a nucleation-event followed by growth, the cores act as the nuclei, and the shells grow from their surfaces. The temperature of the reaction is kept low to favor the addition of shell material monomers to the core surface, while preventing independent nucleation of nanocrystals of the shell materials. Surfactants in the reaction mixture are present to direct the controlled growth of shell material and ensure solubility. A uniform and epitaxially grown shell is obtained when there is a low lattice mismatch between the two materials. Exemplary materials for preparing core-shell nanocrystals can include, but are not limited to, Si, Ge, Sn, Se, Te, B, C (including diamond), P, Co, Au, BN, BP, BAs, AlN, AlP, AlAs, AlSb, GaN, GaP, GaAs, GaSb, InN, InP, InAs, InSb, AlN, AlP, AlAs, AlSb, GaN, GaP, GaAs, GaSb, ZnO, ZnS, ZnSe, ZnTe, CdS, CdSe, CdTe, HgS, HgSe, HgTe, BeS, BeSe, BeTe, MgS, MgSe, GeS, GeSe, GeTe, SnS, SnSe, SnTe, PbO, PbS, PbSe, PbTe, CuF, CuCl, CuBr, CuI, Si3N4, Ge3N4, Al2O3, (Al, Ga, In)2 (S, Se, Te)3, Al2CO, and combinations thereof. Exemplary core-shell luminescent nanocrystals include, but are not limited to, (represented as Core/Shell), CdSe/ZnS, InP/ZnS, PbSe/PbS, CdSe/CdS, CdTe/CdS, CdTe/ZnS, as well as others. In one or more embodiments, the nanocrystals can include functionalized ligands to promote adhesion to the surface 32 of current collector 30 or to additional material of the charge carrier separator 40 as is further described herein. The ligands may also be used to facilitate bonding to the nuclear radiation-emitting material 60 as is also further described herein. The ligands may also facilitate charge transfer from the nanocrystals to at least one of the charge carrier separator 40, the current collector 30, and the hole conductor layer 50. In one or more embodiments, the charge carrier separator 40 can be a layer that is formed or deposited onto the at least a portion of the surface 32 of three-dimensional current collector 30. In one or more embodiments, the charge carrier separator 40 can include two or more layers. In one or more embodiments, the charge carrier separator 40 can include other types of structures, e.g., quantum wells, PN junctions, PIN junctions, schottky junctions, and perovskite structures. And in one or more embodiments, the charge carrier separator 40 can include two or more types of materials, e.g., an oxide layer combined with nanocrystals, two or more different types of nanocrystals, one or more quantum wells combined with nanocrystals, etc., as is further described herein. The power converter 10 can also include the hole conductor layer 50. In one or more embodiments, the hole conductor layer 50 is disposed on at least a portion of the charge carrier separator 40. Further, in one or more embodiments, the hole conductor layer 50 is disposed on at least a portion of the surface 32 of the three-dimensional current collector 30. And in one or more embodiments, the hole conductor layer 50 can be disposed on at least a portion of the charge carrier separator 40 and at least a portion of the surface 32 of the three-dimensional current collector 30. In one or more embodiments, the hole conductor layer 50 can be disposed on at least a portion of the surface 32 of the three-dimensional current collector 30, and the charge carrier separator 40 can be disposed within the hole conductor layer such that it is electrically coupled with the current collector 30. In one or more embodiments, the hole conductor layer 50 is electrically coupled to the second electrode 20 to provide an electrical pathway for one or both of the electrons and holes emitted by the charge carrier separator 40. The hole conductor layer 50 can include any suitable material or materials. For example, in one or more embodiments, the hole conductor layer 50 can include any suitable p-type semiconductor material, e.g., CsSnI3, ZnO, CuSCN, doped or undoped graphene, hole conducting hole transport medium (i.e., hole conductor, hole transport medium) such as PTAA or PEDOT, and liquid redox shuttles such as I-/I3-. The term “p-type” or “p-doped” as used in this disclosure refer to a semiconductor material that includes a dopant that provides for excess holes to act as positive, or “p-type,” mobile charge carriers. In one example, a p-type dopant can accept an electron from the semiconductor material. The p-type semiconductor material may also be made of intrinsic p-type material. Further, the terms “n-doped” or “n-type” as they are used in this disclosure refer to a semiconductor material that includes a dopant that provides for excess electrons to act as negative, or “n-type,” mobile charge carriers. In one example, an n-type dopant can donate one or more valence electrons to a semiconductor material. The n-type semiconductor material may also be made of intrinsic n-type material. The power converter 10 can also include nuclear radiation-emitting material 60. The material 60 can be disposed in any suitable location. For example, in one or more embodiments, the material 60 can be disposed proximate the charge carrier separator 40 to minimize losses in particle energy. As used herein, the phrase “proximate the charge carrier separator” means that the nuclear radiation-emitting material is disposed such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon the charge carrier separator. For example, in one or more embodiments, the material 60 can be disposed within at least one of the three-dimensional current collector 30, charge carrier separator 40, and hole conductor layer 50 (as illustrated in FIG. 1B). In one or more alternative embodiments, the material 60 can be disposed such that a counter electrode 70 is between the material 60 and the charge carrier separator 40 (see, e.g., power converter 600 of FIG. 6). The nuclear radiation-emitting material 60 can include any suitable material or materials. In one or more embodiments, the nuclear radiation-emitting material 60 can include a plurality of radiation-emitting radioisotopes, e.g., tritium 3H, 60Co, 63Ni, 90Sr, 99Tc, 127Cs and combinations thereof. And the material 60 can emit any suitable type of particles, e.g., alpha, beta, gamma, x-ray, etc. In one or more embodiments, the power converter 10 can include optional counter electrode 70 disposed between the hole conductor layer 50 and second electrode 20. The counter electrode 70 can be electrically coupled to the hole conductor layer 50. In one or more embodiments, the counter electrode 70 can electrically couple the hole conductor layer 50 and the second electrode 20. In one or more embodiments, the counter electrode 70 can be in contact with the second electrode 20. In one or more embodiments, the counter electrode can be electrically coupled to the second electrode 20 through a conductive adhesive 72 (FIG. 1A). Further, in one or more embodiments, the counter electrode 70 can be electrically coupled to the hole conductor layer 50 through a conductive adhesive (not shown). And in one or more embodiments, the counter electrode 70 can serve as the second electrode, thereby replacing second electrode 20. The counter electrode 70 can include any suitable material or materials, e.g., Au, Pt, graphene, a metallic material, a conducting polymer, a semiconductor, or combinations thereof. The power converter 10 can include any other suitable layer or layers. For example, in one or more embodiments, the power converter 10 can include one or more absorption layers for absorbing nuclear radiation particles that are emitted by the nuclear radiation-emitting material 60 to prevent the release of nuclear radiation particles from the power converter. Such one or more absorbing layers may also absorb bremsstrahlung (x-rays) resulting from the deceleration of nuclear radiation particles emitted by the nuclear radiation-emitting material 60. While not wishing to be bound by any particular theory, the power converter 10 can provide a current to a device or system that is electrically coupled to the converter by converting energy from radioactive decay of the nuclear radiation-emitting material 60 into electrical energy. For example, in reference to FIG. 1B, the nuclear radiation-emitting material 60 can emit one or more nuclear radiation particles 62, e.g., an electron, along with an antineutrino 64 as the material 60 decays. The particle 62 can generate electron/hole pairs 44/46 in the charge carrier separator 40 through, e.g., impact ionization. One or more liberated electrons 44 or excitons can be injected into a conduction band of the three-dimensional current collector 30. The collector 30 can direct these liberated electrons 44 to the first electrode 12 (FIG. 1A) before the liberated electrons 44 can recombine with their associated holes 46. The hole conductor layer 50 can fill the holes 46, i.e., electron vacancies, in the charge carrier separator 40 by replenishing the separator 40 with electrons. This replenishing of electrons can aid in preventing recombination of electron-hole pairs 44/46 before the liberated electrons 44 can be collected by the collector 30. In one or more embodiments, the electrons 44 can be absorbed by the three-dimensional current collector 30 more quickly than the electrons can recombine with the associated holes 46, thereby also helping to prevent the electron-hole pairs from recombining. As mentioned herein, the power converters of the present disclosure can include any suitable material or combination of materials for the charge carrier separator. For example, FIG. 2 is a schematic cross-section view of a portion of one of embodiment of a power converter 200. The power converter 200 is similar in many aspects to power converter 10 illustrated in FIGS. 1A-B. All of the design considerations and possibilities regarding the power converter 10 of FIGS. 1A-B apply equally to the power converter 200 of FIG. 2. One difference is that power converter 200 includes a charge carrier separator 240 that includes a first material 242 disposed on at least a portion of a surface 232 of a three-dimensional current collector 230, and a second material 244. In the embodiment illustrated in FIG. 2, the second material 244 is disposed on at least a portion of the first material 242. In one or more embodiments, the second material 244 can be disposed within the first material 242, or spaced apart from the first material 242. Charge carrier separator 240 can include any suitable material or combination of materials. For example, in one or more embodiments, the first material 242 can include an oxide, e.g., an oxide of the material utilized for the three-dimensional current collector 230, and the second material 244 can include quantum dots as described herein. While not wishing to be bound by any particular theory, the power converter 200 can convert energy from nuclear radiation-emitting material 260 into electrical energy by, e.g., impact ionization. For example, nuclear radiation-emitting material 260 can emit nuclear radiation particle 262 (e.g., an electron) and an antineutrino 264. The particle 262 is incident upon at least one of the first material 242 and the second material 244 of the charge carrier separator 240. The impact of the particle 262 on one or both of materials 242, 244 can cause the formation of at least one electron-hole pair 246/248. The electrons 246 from the pairs 246/248 can be directed by the three-dimensional current collector 230 to an electrode (not shown) as is further described herein. As mentioned herein, the power converters of the present disclosure can include nuclear radiation-emitting material that can be disposed in any suitable location such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon a charge carrier separator. For example, FIG. 3 is a schematic cross-section view of a portion of another embodiment of a power converter 300. All of the design considerations and possibilities regarding the power converters 10 and 200 of FIGS. 1A-B and FIG. 2 apply equally to the power converter 300 of FIG. 3. One difference is that power converter 300 includes nuclear radiation-emitting material 360 disposed in three-dimensional current collector 330. Any suitable technique or combination of techniques can be utilized to provide nuclear radiation-emitting material 360 within three-dimensional current collector 330. The nuclear radiation-emitting material 360 is disposed such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material 360 is incident upon at least one of first material 342 and second material 344 of charge carrier separator 340. Further, in one or more embodiments, nuclear radiation-emitting material can be disposed within charge carrier separator material. For example, FIG. 4 is a schematic cross-section view of a portion of a power converter 400. All of the design considerations and possibilities regarding the power converters 10 and 200 as illustrated in FIGS. 1A-B and FIG. 2 apply equally to the power converter 400 of FIG. 4. As illustrated in FIG. 4, nuclear radiation emitting material 460 is disposed within second material 444 of charge carrier separator 440. Any suitable technique or combination of techniques can be utilized to dispose nuclear radiation-emitting material within second material 444. As previously described herein, nuclear radiation-emitting material 460 is disposed within second material 444 such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon at least one of first material 442 and second material 444 of the charge carrier separator 440. In one or more embodiments, the charge carrier separator 440 can be disposed on at least a portion of a surface of three-dimensional current collector 430. Nuclear radiation-emitting material can also be disposed within a counter electrode. For example, FIG. 5 is a schematic cross-section view of a portion of a power converter 500. The power converter 500 can include any suitable power converter, e.g., power converters 10 and 200. All of the design considerations and possibilities regarding power converters 10 and 200 of FIGS. 1A-B and FIG. 2 apply equally to power converter 500 of FIG. 5. As illustrated in FIG. 5, nuclear radiation-emitting material 560 is disposed within counter electrode 570 such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon at least one of first material 542 and second material 544 of charge carrier separator 540. Nuclear radiation-emitting material 560 can be disposed within counter electrode 570 using any suitable technique or combination of techniques. In one or more embodiments, the charge carrier separator 540 can be disposed on at least a portion of a surface of three-dimensional current collector 530. Nuclear radiation-emitting material of the power converters described herein can also be disposed such that a counter electrode is between the nuclear radiation-emitting material and the charge carrier separator. For example, FIG. 6 is a schematic cross-section view of a portion of a power converter 600. All of the design considerations and possibilities regarding the power converters 10 and 200 as illustrated in FIGS. 1A-B and FIG. 2 apply equally to power converter 600 of FIG. 6. As illustrated in FIG. 6, nuclear radiation-emitting material 660 is disposed in matrix 680 adjacent counter electrode 670. The nuclear radiation-emitting material 660 is disposed such that the counter electrode 670 is between the nuclear radiation-emitting material and a charge carrier separator 640. The nuclear radiation-emitting material 660 is disposed such that at least one nuclear radiation particle emitted by the nuclear radiation-emitting material is incident upon at least one of first material 642 and second material 644 of the charge carrier separator 640. The nuclear radiation-emitting material 660 and matrix 680 can be disposed on the counter electrode 670. Alternatively, one or more intervening layers can be disposed between the nuclear radiation-emitting material 660 and matrix 680, and the counter electrode 670. The matrix 680 can include any suitable material or materials, e.g., polymers such as paraffin, liquids such as water, solids such as metals and oxides, etc. Any suitable technique or combination of techniques can be utilized to produce the power converters described herein. Referring to FIGS. 1A-B, the three-dimensional current collector 30, in one or more embodiments, can be formed between the first and second electrodes 12, 20. The collector 30 can be formed using any suitable technique or combination of techniques, e.g., sintering, pressing, electrophoresis, anodic growth, cathodic reduction, etching, photolithography, 3D printing, or combinations thereof. As mentioned herein, the collector 30 can be formed on either of the first and second electrodes 12, 20. In one or more embodiments, the three-dimensional current collector 30 is electrically coupled to the first electrode 12. The charge carrier separator 40 can be formed on at least a portion of the surface 32 of the three-dimensional current collector 30 using any suitable technique or combination of techniques, e.g., anodization, drop coating, atomic layer deposition (ALD), sequential ionic layer adsorption, reaction (SILAR), chemical bath deposition (CBD) or combinations thereof. The three-dimensional current collector 30 and the charge carrier separator 40 can form a heterostructure. The hole conductor layer 50 can be formed on at least a portion of the charge carrier separator 40. In one or more embodiments, the hole conductor layer 50 is formed using any suitable technique or combination of techniques, e.g., drop coating, atomic layer deposition (ALD), sequential ionic layer adsorption and reaction (SILAR), chemical bath deposition (CBD) or combinations thereof. The nuclear radiation-emitting material 60 can be formed proximate the charge carrier separator 40. In one or more embodiments, the nuclear radiation-emitting material 60 can be formed within or attached to at least one of the three-dimensional current collector 30, the charge carrier separator 40, and the hole conductor layer 50 using any suitable technique or combination of techniques, e.g., exposure to a gaseous form of the material 60 (e.g., tritium gas) under pressure and/or heat, electroplating, synthesis, or combinations thereof. In one or more embodiments, the nuclear radiation-emitting material 60 can be formed within the three-dimensional current collector 30 by depositing tritiated paraffin wax onto the current collector. In one or more embodiments, the optional counter electrode 70 can be formed between the hole conductor layer 50 and the second electrode 20 such that the counter electrode electrically couples the hole conductor layer 50 and the second electrode 20. Any suitable technique or combination of techniques can be utilized to form the counter electrode 70, e.g., drop coating (if liquid), atomic layer deposition (ALD), etc. In one or more embodiments, where the nuclear radiation-emitting material 60 is formed such that the counter electrode 70 is between the material 60 and the charge carrier separator 40, the material can be formed, e.g., on the counter electrode using any suitable technique or combination of techniques. For example, in one or more embodiments, the nuclear radiation-emitting material 60 can be formed on the counter electrode 70 by depositing tritiated paraffin wax onto the counter electrode. The various elements of the power converter can be formed in any suitable order. For example, in one or more embodiments, the counter electrode 70 can be formed first, followed by the hole conductor layer 50, the charge separator 40, current collector 30, and the first and second electrodes 12, 20. In one or more embodiments, it may be advantageous to dispose the nuclear radiation-emitting material 60 in the desired location at the end of the process to help prevent the material 60 from degrading other elements of the power converter. The various embodiments of power converters described herein can be utilized as a current source for any suitable devices or systems. For example, FIG. 7 is a schematic cross-section of one embodiment of an implantable medical device system 700 that includes a power converter 710 and an implantable medical device 714. The power converter 710 can include any suitable power converter described herein, e.g., power converter 10 of FIGS. 1A-B. The implantable medical device 714 can include any suitable medical device, e.g., electrocardiogram (ECG) monitors, sensors (such as glucose, pressure), implantable pulse generators (IPGs) (e.g., pacemakers), implantable cardioverter defibrillators (ICDs), etc. Although not shown, the power convertor 710 can be electrically coupled to the implantable medical device 714 using any suitable technique or combination of techniques. All references and publications cited herein are expressly incorporated herein by reference in their entirety into this disclosure, except to the extent they may directly contradict this disclosure. Illustrative embodiments of this disclosure are discussed and reference has been made to possible variations within the scope of this disclosure. These and other variations and modifications in the disclosure will be apparent to those skilled in the art without departing from the scope of the disclosure, and it should be understood that this disclosure is not limited to the illustrative embodiments set forth herein. Accordingly, the disclosure is to be limited only by the claims provided below. |
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claims | 1. An extreme ultraviolet light generation apparatus used in combination with a laser system, the apparatus comprising:a chamber provided with at least one inlet port and configured for introducing a laser beam outputted from the laser system into the chamber;a target supply unit provided to the chamber and configured for supplying a target material to a predetermined region inside the chamber, where the target material is irradiated with the laser beam;a gas introduction unit configured for introducing an etching gas including hydrogen gas into the chamber;a pressure detector configured for detecting pressure inside the chamber;a mass flow controller configured for controlling a flow rate of the etching gas;an exhaust unit configured for discharging gas inside the chamber, anda controller configured for controlling at least one of the mass flow controller and the exhaust unit based on a value detected by the pressure detector. 2. The apparatus of claim 1, wherein the pressure inside the chamber is approximately at or above 0.5 Pa and is approximately at or below 13 Pa. 3. The new apparatus according to claim 2, further comprising:a magnetic field generation unit configured for generating a magnetic field around the predetermined region; andan ion collection unit disposed in a direction of a line of magnetic force of the magnetic field and configured for collecting an ion which is generated when the target material is irradiated with the laser beam and flows along the line of magnetic force. 4. The apparatus of claim 3, further comprising:a magnetic field intensity controller configured for controlling intensity of the magnetic field;a power source configured for supplying current to the magnetic field generation unit; andan ion sensor disposed close to the at least one optical element and configured for detecting an ion. 5. The apparatus of claim 3, wherein the magnetic flux density of the magnetic field is approximately at or above 0.35 tesla and is approximately at or below 2 tesla. 6. The apparatus of claim 3, further comprisingis a collector mirror configured for collecting extreme ultraviolet light emitted as the target material is irradiated with the laser beam inside the chamber. 7. The apparatus of claim 6, wherein:the magnetic field generation unit includes a coil to which current is supplied and a magnetic core which extends toward the predetermined region from the coil, anda part of the magnetic core extends into an obscuration region of the extreme ultraviolet light collected by the collector mirror. 8. The apparatus of claim 7, whereina coating film is formed on a surface of the magnetic core. 9. The apparatus of claim 7, wherein:the magnetic core is cylindrical with one end thereof being open, andthe ion collection unit is disposed at the other end of the magnetic core. 10. The apparatus of claim 9, wherein a coating film is formed on a surface of the ion collection unit. 11. An extreme ultraviolet light generation apparatus used in combination with a laser system, the apparatus comprising:a chamber provided with at least one inlet port and configured for introducing a laser beam outputted from the laser system into the chamber;a target supply unit provided to the chamber and configured for supplying a target material to a predetermined region inside the chamber, where the target material is irradiated with the laser beam;a gas introduction unit configured for introducing an etching gas mending hydrogen gas into the chamber;a pressure detector configured for detecting pressure inside the chamber;a mass flow controller configured for controlling a flow rate of the etching gas;an exhaust unit configured for discharging gas inside the chamber;a controller configured for controlling at least one of the mass flow controller and the exhaust unit based on a value detected by the pressure detector;a magnetic field generation unit configured for generating a magnetic field around the predetermined region; andan ion collection unit disposed in a direction of a line of magnetic force of the magnetic field and configured for collecting an ion which is generated when the target material is irradiated with the laser beam and flows along the line of magnetic force. 12. The apparatus of claim 11, further comprisinga free radical source configured for turning the etching gas into a free radical. 13. The apparatus of claim 12, whereinthe pressure inside the chamber is approximately at or above 0.5 Pa and is approximately at or below 13 pa. |
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062326139 | claims | 1. An angular pumped and emitting capillary (APEC) discharge light source, comprising: a first electrode having a gas inlet port; a partially insulating capillary attached to one side of the electrode, having an inlet bore in fluid connection to the gas inlet port of the first electrode, and an outlet bore; and a second electrode facing the outlet bore of the insulating capillary, wherein applying a voltage between the electrodes causes light to be emitted from the outlet bore of the capillary, and debris formed within the capillary are prevented from exiting the outlet bore of the capillary by the second electrode. EUV light. visible light. ultraviolet light. vacuum ultraviolet light. an emitting region having an inwardly angled tapered outlet bore. a portion facing the outlet bore and having an outwardly tapered surface, so that the radiation is emitted in an angular direction, and the second electrode blocks any debris generated within the capillary from being expelled into collecting optics. a diameter would range between approximately 0.5 mm and approximately 2.5 mm and an overall length range between approximately 1 mm and approximately 10 mm. a gap spacing between the second electrode and the outlet bore region less than an interior bore diameter of the capillary. a tapered surface having an outer diameter larger than an interior bore diameter of the capillary. a semiconducting material. a transparent window about the capillary for allowing wavelengths of at least 100 nm to be generated. a first electrode having a gas inlet port; a partially insulating capillary attached to one side of the electrode, having an inlet bore in fluid connection to the gas inlet port of the first electrode, and an outlet bore; and a second electrode adjacent to the capillary; and a collector means facing the outlet bore of the insulating capillary, wherein applying a voltage between the electrodes causes light to be emitted from the outlet bore of the capillary, and debris formed within the capillary are collected by the collector means. a first electrode; a partially insulating capillary attached to one side of the first electrode, having an outlet bore with an emitting region; and a second electrode adjacent to the capillary; through-hole means in the capillary for directing inlet gas into the capillary for providing a higher pressure and increasing emission flux in the emitting region; and a collector-blocking means facing the outlet bore of the insulating capillary, wherein applying a voltage between the electrodes causes light to be emitted from the emitting region of the capillary, and debris formed within the capillary are collected by the collector means. a first electrode having a conduit for introducing an emitting gas therethrough; a second electrode spaced apart by a gap to the first electrode, wherein applying a voltage to the first electrode and the second electrode forms an emitting region in the gap so that the second electrode serves to block and collect electrode debris material; and means for pumping gas away from the emitting region so as to provide a low pressure for transmission of EUV flux. 2. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, wherein the light emitted includes: 3. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, wherein the light emitted includes: 4. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, wherein the light emitted includes: 5. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, wherein the light emitted includes: 6. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, wherein the outlet bore of the capillary includes: 7. The angular pumped and emitting capillary (APEC) discharge light source of claim 6, wherein the second electrode includes: 8. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, wherein the inlet and the outlet bore of the capillary includes: 9. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, further including: 10. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, further includes: 11. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, wherein the capillary includes: 12. The angular pumped and emitting capillary (APEC) discharge light source of claim 1, further comprising: 13. An angular pumped and emitting capillary (APEC) discharge light source, comprising: 14. The angular pumped and emitting capillary discharge light source of claim 13, wherein the second electrode and the collector means are one and the same. 15. The angular pumped and emitting capillary discharge light source of claim 13, wherein the second electrode and the collector means are different components. 16. An angular pumped and emitting capillary (APEC) discharge light source, comprising: 17. The angular pumped and emitting capillary discharge light source of claim 16, wherein the second electrode and the collector-blocking means are one and the same. 18. The angular pumped and emitting capillary discharge light source of claim 16, wherein the second electrode and the collector-blocking means are different components. 19. A discharge device for generating a discharge plasma for forming EUV flux, from two electrodes without using a capillary between the two electrodes, comprising: 20. The discharge device for generating the discharge plasma of claim 19, wherein the gas emitting gas is introduced at a pressure of at least approximately 0.1 Torr, and the emitting gas is chosen from one of xenon, helium, neon, argon and krypton. |
claims | 1. A method for fabricating a compound zone plate comprising:placing a first zone plate frame comprising a first zone plate over a second zone plate frame comprising a second zone plate;placing microbeads and an adhesive mechanically between the first zone plate frame and the second zone plate frame; andaligning the first zone plate to the second zone plate prior to hardening of the adhesive by using the microbeads to facilitate inplane relative movement between the first zone plate and the second zone plate. 2. A method as claimed in claim 1, wherein the step of aligning the first zone plate to the second zone plate comprises:bonding the first zone plate frame to a base frame; andbonding the second zone plate frame to the base frame, wherein the base frame is a reference plane used for aligning the first zone plate relative to the second zone plate. 3. A method for fabricating a compound zone plate comprising:placing a first zone plate frame comprising a first zone plate over a second zone plate frame comprising a second zone plate;placing microbeads and an adhesive mechanically between the first zone plate frame and the second zone plate frame; andaligning the first zone plate to the second zone plate prior to hardening of the adhesive by transmitting x-rays through the first zone plate and the second zone plate and detecting the x-rays and positioning the first zone plate relative to the second zone plate in response to the detected x-rays. 4. A method as claimed in claim 3, wherein the x-rays are detected with a spatially resolved detector. 5. A method for fabricating a compound zone plate comprising:placing a first zone plate frame comprising a first zone plate over a second zone plate frame comprising a second zone plate;placing microbeads and an adhesive mechanically between the first zone plate frame and the second zone plate frame;aligning the first zone plate to the second zone plate prior to hardening of the adhesive; andaligning a third zone plate to the first zone plate and the second zone plate. 6. A method as claimed in claim 5, further comprising aligning a fourth zone plate to the first zone plate, the second zone plate, and the third zone plate. 7. A method for fabricating a compound zone plate comprising:placing a first zone plate frame comprising a first zone plate over a second zone plate frame comprising a second zone plate;placing microbeads and an adhesive mechanically between the first zone plate frame and the second zone plate frame;aligning the first zone plate to the second zone plate prior to hardening of the adhesive; andwherein different sized microbeads are used in the adhesive between a base frame and the first zone plate frame relative to the adhesive between the base frame and the second zone plate frame to control a relative spacing of the zone plate frames with respect to each other. 8. A method for fabricating a compound zone plate comprising:placing a first zone plate frame comprising a first zone plate over a second zone plate frame comprising a second zone plate;placing microbeads and an adhesive mechanically between the first zone plate frame and the second zone plate frame;aligning the first zone plate to the second zone plate prior to hardening of the adhesive;placing a third zone plate frame comprising a third zone plate over the second zone plate frame; andplacing microbeads and adhesive mechanically between the second zone plate frame and the third zone plate frame. 9. A method as claimed in claim 8, further comprising:bonding a fourth zone plate frame comprising a fourth zone plate to a base frame and then placing the base frame on the third zone plate frame; andplacing microbeads and adhesive mechanically between the base frame and the third zone plate frame, wherein the base frame is a reference plane used for aligning the zone plates relative to each other. 10. A method for fabricating a compound zone plate comprising:bonding a first zone plate frame comprising a first zone plate over a second zone plate frame comprising a second zone plate; wherein the first zone plate frame and the second zone plate frame are bonded on a base frame and wherein the second zone plate frame comprises an optical port and the first zone plate frame is situated at least partially within the optical port of the second zone plate frame. 11. A method as claimed in claim 10, further comprising using a spacer between the first zone plate frame and the base frame. 12. A method as claimed in claim 10, further comprising fabricating the second zone plate frame on a membrane, and fabricating an optical port into the first zone plate frame to produce an unsupported membrane over the optical port. 13. A method as claimed in claim 12, further comprising positioning the first zone plate laterally within the optical port. 14. A method as claimed in claim 10, further comprising bonding a third zone plate frame comprising a third zone plate to the second zone plate frame. 15. A method as claimed in claim 14, further comprising inserting a spacer between the second zone plate frame and the third zone plate frame. 16. A method as claimed in claim 14, further comprising bonding a fourth zone plate frame to an additional base frame and subsequently bonding the additional base frame to the third zone plate frame. 17. A method for fabricating a compound zone plate comprising:bonding a first zone plate frame including a first zone plate over a base frame;placing a second zone plate frame including a second zone plate over the base frame;inserting a spacer between the base frame and the second zone plate frame, wherein the spacer includes microbeads; andaligning the first zone plate to the second zone plate by using the microbeads to facilitate inplane relative movement between the first zone plate and the second zone plate. 18. A method for fabricating a compound zone plate comprising:bonding a first zone plate frame comprising a first zone plate over a base frame;bonding a second zone plate frame comprising a second zone plate over the base frame;positioning and holding the base frame with respect to the first and second zone plate frames with an x, y axis positioner; andaligning the first zone plate to the second zone plate by using the microbeads to facilitate inplane relative movement between the first zone plate and the second zone plate. 19. A method as claimed in claim 18, wherein the aligning of the first zone plate to the second zone plate is based on feedback from x-rays that are transmitted through the first zone plate and the second zone plate. |
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claims | 1. A reuse method of eutectic salt waste including nuclides comprising the steps of:a) producing oxychloride, or oxide and e of nuclides through injection and heating of oxygen into the eutectic salt waste including the nuclides;b) performing a layer separation into a precipitate layer of oxide, oxychloride, or oxide and oxychloride and an eutectic salt layer formed on the upper of the precipitate layer by using free settling;c) separating and collecting the precipitate layer and the eutectic salt layer, respectively;d) recollecting the eutectic salt by distilling the separated precipitate layer and condensing the distilled eutectic salt; ande) reusing the eutectic salt collected from step (c) and the eutectic salt recollected in step (d) in an electro refining process of nuclear fuel. 2. The method as set forth in claim 1, wherein oxide, oxychloride, or oxide and oxychloride of the nuclides of the step a) are produced at a temperature of 600 to 700° C. 3. The method as set forth in claim 1, wherein the separation and collection of the step c) uses siphons. 4. The method as set forth in claim 1, wherein the distillation of the step d) is distilled at a temperature of 800 to 1000° C. under a pressure of 0.1 to 100 torr. |
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description | This application claims priority to Korean Patent Application No. 10-2014-0006212, filed on Jan. 17, 2014 the disclosure of which is incorporated herein by reference in its entirety. Field Apparatuses and methods consistent with exemplary embodiments relate to a manipulator, and more particularly, to an articulated manipulator capable of moving a tool such as an inspection device, a processing device, or a welding device to a desired position for inspection or repair of a defect portion in a limited place. Description of the Related Art A steam generator constituting a nuclear steam supply system of a nuclear power plant is a device which generates dry steam when high-temperature primary coolant supplied from a reactor exchanges heat with secondary coolant outside heat transfer tubes while flowing along the heat transfer tubes inside the steam generator. A primary system of a nuclear power plant is typically configured by a reactor, a steam generator, a coolant circulation pump, a pressurizer, and the like. Here, the reactor has a control rod and a fuel assembly therein, and heat energy generated in the reactor is transferred to the outside of the reactor through a primary coolant. The primary coolant is circulated through heat transfer tubes via the steam generator. The steam generator is supplied with secondary coolant which exchanges heat with the primary coolant. The secondary coolant exchanges heat with the high-temperature and high-pressure primary coolant to be vaporized. Steam generated by the steam generator is transferred to a turbine through a pipe so as to drive a generator and generate electric energy. FIG. 1 is a view illustrating an example of a typical steam generator. Referring to FIG. 1, a steam generator 10 has a plurality of heat transfer tubes (several thousand or more heat transfer tubes) 11 therein, and a tube sheet 12 and a channel head 13 are formed at a lower portion of the steam generator 10. An inner portion of the channel head 13 is partitioned by a partition plate 14, and an inlet 15 and an outlet 16 are formed on an outer peripheral surface of the channel head 13 so as to respectively communicate with spaces partitioned thereby. Primary coolant discharged from a reactor is introduced into the inner space through the inlet 15 of the channel head 13, exchanges heat with secondary coolant through the heat transfer tubes 11 via the tube sheet 12, and is then discharged to the outlet 16. Accordingly, the tube sheet 12 is formed with through-holes (not shown) communicating with the respective heat transfer tubes 11. A large number of defects due to corrosion or external impact tend to be generated at welding portions of the partition plate 14 after the steam generator 10 has been operated for a long time. Thus, the defects have to be inspected and repaired in order to prevent safety accidents. In this case, the inspection and repair are mainly performed by mechanical devices which are remotely controlled from the outside, because of a danger of residual radioactivity and also because of the small space within the channel head 13. Therefore, there is a need for a manipulator capable of accurately locating an inspection device or a repair device at a defect portion in order to improve accuracy of inspection and repair. FIG. 2 is a cross-sectional view illustrating various welding portions in the channel head. As shown in FIG. 2, the channel head 13 is provided with welding portions which are respectively formed between the partition plate 14 and the channel head 13, between the partition plate 14 and the tube sheet 12, and between the partition plate 14 and a stub runner 17. The manipulator needs to move the inspection device or the repair device to various positions corresponding to the respective welding portions. In addition, the inner space of the channel 13 is formed in various sizes according to the standard of the steam generator 10. Accordingly, in order for the manipulator to be applied to the various-sized work spaces, the number of rotary shafts of or shape of the manipulator needs to be actively changed and the shape of the manipulator has to be easily changed for improvement of workability. Accordingly, in view of the above-mentioned problems, one or more exemplary embodiments provide an articulated manipulator capable of accurately locating an inspection device or a repair device at a defect portion so as to inspect or repair defects generated in a limited space as in a channel head or pipe of a steam generator. One or more exemplary embodiments provide an articulated manipulator capable of actively corresponding according to positions of defect generation portions or shapes of spaces requiring inspection or repair and easily performing maintenance, by assembling or disassembling a plurality of rotation modules. Other objects and advantages of the exemplary embodiments can be understood by the following description. Also, it is obvious to those skilled in the art that the objects and advantages can be realized by the means as claimed and combinations thereof. In accordance with an exemplary embodiment, an articulated manipulator includes a base plate having a length that is longer than a width thereof, a movable unit slidably coupled to the base plate and configured to slide in a direction of the length of the base plate, a rotatable unit rotatably coupled to an upper side of the movable unit and configured to rotate along a plane which is parallel to a plane on which the base plate is disposed, and a rotation unit rotatably coupled to one side of the rotatable unit. The rotation unit may comprise a plurality of rotation modules which are configured to be assembled together in a modular manner. Each of the rotation modules may include a coupling block coupled to one side of the rotatable unit or one side of another adjacent rotation module, and a rotation member rotatably coupled to the coupling block. Each of the rotation modules may further include a rotary motor provided at one side of the rotation member to rotate the rotation member relative to the coupling block. Each of the rotation modules may further include an arm coupled to one side of the rotation member. At least one support member having an adjustable length by an actuation cylinder may be provided at an upper side of the base plate. A feed motor may be provided at one side of the base plate, a lead screw may be provided in a longitudinal direction of the base plate, and the movable unit may be screwed to the lead screw to slide by driving of the feed motor. A pair of guide rails may be installed on the base plate and the movable unit may slide along the guide rails. A drive motor which rotates the rotatable unit relative to the movable unit may be provided at one side of the rotatable unit. The rotatable unit may horizontally rotate with respect to the movable unit, and the rotation unit may rotate in a direction perpendicular to a rotational direction of the rotatable unit. An end effector may be provided at one side of the rotation unit. A processing device having a reaction measurement sensor may be coupled to the end effector. A support plate may be vertically coupled to one side of the base plate. In accordance with another exemplary embodiment, an articulated manipulator includes a base plate having a length that is longer than a width thereof, a pair of guide rails installed on the base plate parallel to and along the length thereof, a movable unit slidably installed on the guide rails and configured to rotate along a plane which is parallel to a plane on which the base plate is disposed, a rotatable unit rotatably coupled to an upper side of the movable unit and configured to rotate along a plane which is parallel to a plane on which the base plate is disposed, and a first rotation module which is vertically and rotatably coupled to one side of the rotatable unit. The first rotation module may include a first coupling block coupled to one side of the rotatable block, a first rotation member rotatably coupled to one side of the first coupling block, and a first rotary motor provided at one side of the first rotation member to rotate the first rotation member relative to the first coupling block. The first rotation module may further include a first arm coupled to one side of the first rotation member. A first coupling portion which is protrusively formed at one side of the first coupling block may be coupled to one side of the rotatable unit. The articulated manipulator may further include a second rotation module which is vertically and rotatably coupled to one side of the first rotation module. The second rotation module may include a second coupling block coupled to one side of the first rotation module, a second rotation member rotatably coupled to one side of the second coupling block, and a second rotary motor provided at one side of the second rotation member to rotate the second rotation member relative to the second coupling block. The second rotation module may further include a second arm coupled to one side of the second rotation member. A second coupling portion which is protrusively formed at one side of the second coupling block may be coupled to a first coupling plate provided at an end of a first arm of the first rotation module. The articulated manipulator may further include a third rotation module which is vertically and rotatably coupled to one side of the second rotation module. The third rotation module may include a third coupling block coupled to one side of the second rotation module, a third rotation member rotatably coupled to one side of the third coupling block, and a third rotary motor provided at one side of the third rotation member to rotate the third rotation member relative to the third coupling block. A third coupling portion which is protrusively formed at one side of the third coupling block may be coupled to a second coupling plate provided at an end of a second arm of the second rotation module. The articulated manipulator may further include a fourth rotation module which is vertically and rotatably coupled to one side of the third rotation module. The fourth rotation module may include a fourth coupling block coupled to one side of the third rotation module, a fourth rotation member rotatably coupled to one side of the fourth coupling block, and a fourth rotary motor provided at one side of the fourth rotation member to rotate the fourth rotation member relative to the fourth coupling block. A fourth coupling portion which is protrusively formed at one side of the fourth coupling block may be coupled to one side of the third rotation member of the third rotation module. An end effector may be provided at the other side of the fourth rotation member. A processing device having a reaction measurement sensor may be coupled to the end effector. At least one support member having an adjustable length by an actuation cylinder may be provided at an upper side of the base plate. A support plate may be vertically coupled to one side of the base plate. In accordance with another exemplary embodiment, an articulated manipulator includes a base plate having a length that is longer than a width thereof, a pair of guide rails installed on the base plate parallel to and along the length thereof, a movable unit installed on the base plate so as to be slidable along the guide rails, a rotatable unit coupled to an upper portion of the movable unit to be rotated by a drive motor, and a rotation unit coupled to one side of the rotatable unit to be rotated by a rotary motor, wherein a rotary shaft of the rotatable unit is perpendicular to a rotary shaft of the rotation unit. The rotation unit may be configured by assembly of a plurality of rotation modules. Each of the rotation modules may include a coupling block coupled to one side of the rotatable unit or one side of another adjacent rotation module, and a rotation member rotatably coupled to one side of the coupling block to be rotated by the rotary motor. Each of the rotation modules may further include an arm coupled to one side of the rotation member. At least one rotary shaft of the plurality of rotation modules may be perpendicular to a rotary shaft of another rotation module. An end effector may be provided at an end of a rotation module which is lastly assembled in the plurality of rotation modules. A processing device having a reaction measurement sensor may be coupled to the end effector. At least one support member having an adjustable length by an actuation cylinder may be provided at an upper side of the base plate. A support plate may be vertically coupled to one side of the base plate. It is to be understood that both the foregoing general description and the following detailed description are exemplary and explanatory and are intended to provide further explanation of the invention as claimed. Hereinafter, an articulated manipulator according to exemplary embodiments will be described in more detail with reference to the accompanying drawings. In the description, the thickness of each line or the size of each component illustrated in the drawings may be exaggerated for convenience of description and clarity. In addition, terms to be described later are terms defined in consideration of functions of the exemplary embodiments, and these may vary with the intention or practice of a user or an operator. Therefore, such terms should be defined based on the entire content disclosed herein. In addition, the following embodiments are for the purpose of describing the components set forth in the appended claims only and are not intended to limit the spirit and scope of the invention. More particularly, various variations and modifications are possible in concrete constituent elements of the embodiments, and it is to be understood that differences relevant to the variations and modifications fall within the spirit and scope of the present disclosure defined in the appended claims. Although an example in which an articulated manipulator is installed in a channel head to inspect and repair a welding portion of a partition plate is described in the following embodiment, the articulated manipulator may be used in any limited place, such as in a pipe of a nuclear power plant, in addition to the examples herein. FIG. 3 is a perspective view illustrating an articulated manipulator according to an exemplary embodiment. FIG. 4 is a schematic view illustrating the articulated manipulator according to the exemplary embodiment. As shown in FIGS. 3 and 4, an articulated manipulator 100 according to an embodiment of the present invention includes a base plate 200, a movable block or unit 300 which is slidably coupled on the base plate 200, a rotatable block or unit 400 which is rotatably coupled on the movable block 300, and a rotation unit 500 which is rotatably coupled to one side of the rotatable block 400. The base plate 200 supports the movable block 300, the rotatable block 400, and the rotation unit 500, and is of a rectangular plate shape having a length longer than a width. A pair of guide rails 210 is installed on an upper surface of the base plate 200, and the movable block 300 is installed on the guide rails 210 of the base plate 200 to move in a longitudinal direction of the base plate 200. In this case, an elongated lead screw 220 is installed between the pair of guide rails 210 in the longitudinal direction of the base plate 200. One end of the lead screw 220 is coupled to a feed motor 230 installed to one side of the upper surface of the base plate 200. That is, the lead screw 220 axially rotates when the feed motor 230 is driven. The movable block 300 slides along the guide rails 210 in the longitudinal direction of the base plate 200, and is moved by an operation of the feed motor 230. To this end, both sides of a bottom surface of the movable block 300 are formed with rail portions (not shown) coupled to the guide rails 210, and the lead screw 220 is screwed to the movable block 300 so as to pass through a center thereof. Accordingly, when the lead screw 220 axially rotates during driving of the feed motor 230, the movable block 300 screwed to the lead screw 220 slides along the guide rails 210. A support plate 600 is vertically coupled to one side of a bottom surface of the base plate 200. The support plate 600 is configured such that the base plate 200 is not moved and is securely supported during installation of the articulated manipulator 100, and one surface of the support plate 600 may be adhered to and/or supported by a partition plate 14 (see FIG. 2) of a channel head 13 (see FIG. 2). That is, the support plate 600, coupled to one side of the bottom surface of the base plate 200, may be adhered to and/or supported by one side of the partition plate 14, and the other side of the base plate 200 is supported by an inner wall of the hemispherical channel head 13, so that the base plate 200 is securely installed within the channel head 13 (see FIG. 5). In this case, at least one support member 700 is provided on the upper surface of the base plate 200. A total of four support members 700 are preferably provided one by one at respective corner portions of the upper surface of the base plate 200. One end of each support member 700 is coupled to the associated corner portion of the upper surface of the base plate 200 and the other end thereof extends upward. The support members 700 press a tube sheet 12 (see FIG. 2) at an upper portion of the channel head 13 to securely support the base plate 200 by reaction thereof. One side of each support member 700 is provided with an actuation cylinder 710 capable of adjusting a length of the support member 700 and a plurality of jaws 720 spaced apart from each other extend from respective upper ends of the support members 700. A cylinder rod of the actuation cylinder 710 may be elongated and shortened by hydraulic or pneumatic pressure or driving of a motor. When the support members 700 are elongated upward by the operation cylinders 710 during installation of the base plate 200, the jaws 720 are inserted into the through-holes (not shown) of the tube sheet 12 communicating with heat transfer tubes 11 (see FIG. 2) to push up against the tube sheet 12. In this case, the reaction force applied to the base plate 200 through the support members 700 presses the base plate 200 and the support plate 600 toward the inner wall of the channel head 13. Thus, the base plate 200 and the support plate 600 are securely supported in a state in which edge portions thereof are laid on the inner wall of the channel head 13. In this case, all of the jaws 720 at the upper ends of the support members 700 may be inserted into the same through-hole of the tube sheet 12, or the respective jaws 720 may be inserted into separate through-holes. Standards such as the number and diameters of the jaws 720 and a spaced distance between the jaws 720 may be properly selected as necessary. The rotatable block 400 is rotatably coupled to an upper side of the movable block 300. The rotatable block 400 is coupled to the movable block 300 in a dovetail manner, and rotates relative to the movable block 300 in a horizontal direction on the movable block 300. For example, referring to an x-y-z coordinate system shown in FIG. 3, the movable block 300 slides in a direction parallel with an x-axis and the rotatable block 400 rotates along a plane parallel with an x-y plane. To this end, a drive motor 410 is installed to an upper side of the rotatable block 400 and a rotary shaft of the rotatable block 400 is coupled to an end of the drive motor 410. That is, the rotatable block 400 rotates in the horizontal direction on the movable block 300 by driving of the drive motor 410. The rotation unit 500 is rotatably coupled to one side of the rotatable block 400. In this case, a rotational direction of the rotation unit 500 is a direction perpendicular to a rotational direction of the rotatable block 400. For example, the rotation unit 500 rotates along a plane parallel with a z-y plane or a plane parallel with a z-x plane. The rotation unit 500 is configured by assembly of at least one rotation module. The number of rotation modules for configuring the rotation unit 500 may be properly selected according to work conditions such as an installation place of the articulated manipulator 100 and a position of a defect portion. For example, when the installation place of the articulated manipulator 100 is narrow or the defect portion is close to the base plate 200, the rotation unit 500 required for the purpose may be sufficiently configured by fewer rotation modules. However, when the installation place of the articulated manipulator 100 is wide or the defect portion is far away from the base plate 200, there is a need to configure the rotation unit 500 having more rotary shafts by assembly of more rotation modules in order for an end effector at an end of the rotation unit 500 to be located close to the defect portion. Although FIGS. 3 and 4 show an example of configuring the rotation unit 500 by assembly of four rotation modules, the present invention is not limited thereto. For example, the rotation unit 500 may also be configured by one rotation module or may also be configured by assembly of five or more rotation modules, as necessary. One of the characteristics of the present disclosure is to actively configure the articulated manipulator 100 according to a particular environment of an installation place, a defect portion, and a type of work by assembly of the proper number of rotation modules as necessary. In addition, the articulated manipulator 100 may be easily transported, assembled, and disassembled since the movable block 300, the rotatable block 400, and the rotation unit 500 are assembled in a modular manner. In addition, since the plurality of rotation modules having the respective rotary shafts is mutually coupled by bolting in the rotation unit 500, a worker may easily change the number of rotary shafts configuring the rotation unit 500 according to a work condition such as an installation space. Hereinafter, the articulated manipulator 100 of the present invention will be described with reference to the embodiment shown in FIGS. 3 and 4. A first rotation module 510 is rotatably coupled to one side of the rotatable block 400. In this case, the first rotation module 510 includes a first coupling block 511 coupled to one side of the rotatable block 400, a first rotation member 512 rotatably coupled to one side of the first coupling block 511, and a first rotary motor 513 provided at one side of the first rotation member 512. The first coupling block 511 is of a circular block shape as a whole, and a first coupling portion 511a is protrusively formed at one side of the first coupling block 511. The first coupling portion 511a is coupled to one side of the rotatable block 400 by bolting. The first rotation member 512 has a cut portion formed at one side thereof so as to surround both side surfaces of the first coupling block 511, and is of a circular block shape as a whole. The first coupling block 511 is received in the cut portion of the first rotation member 512 and the first coupling portion 511a is formed to protrude to the outside of the cut portion. The first rotary motor 513 is installed to one surface of the first rotation member 512 and a rotary shaft of the first rotary motor 513 is coupled to the first rotation member 512. Accordingly, the first coupling block 511 rotates in the cut portion of the first rotation member 512 when the first rotary motor 513 is operated. In this case, the rotary shaft of the first rotary motor 513 is installed in a direction parallel with the x-axis with respect to the first rotation member 512, and the first rotation member 512 rotates along the plane parallel with the z-y plane by operation of the first rotary motor 513. A first arm 514 extending in a direction perpendicular to the rotary shaft of the first rotation member 512 is coupled to one side of an outer peripheral surface of the first rotation member 512, as necessary. A first coupling plate 514a having an extended width is formed for coupling with another rotation module (for instance, a second rotation module) adjacent to an end of the first arm 514. Meanwhile a second rotation module 520 is formed in the same shape as the first rotation module 510. That is, the second rotation module 520 includes a second coupling block 521 configured to have a second coupling portion 521a protrusively formed at one side thereof, a second rotation member 522 has a cut portion formed at one side thereof so as to surround both side surfaces of the second coupling block 521 and being rotatably coupled thereto, a second rotary motor 523 provided at one surface of the second rotation member 522, and a second arm 524 configured to extend from one side of the second rotation member 522 and have a second coupling plate 524a formed at an end of the second arm 524. The first rotation module 510 is assembled to the second rotation module 520 by bolting the second coupling portion 521a of the second rotation module 520 to the first coupling plate 514a of the first rotation module 510. A rotary shaft of the second rotary motor 523 is installed in the direction parallel with the x-axis with respect to the second rotation member 522, and the second rotation member 522 rotates along the plane parallel with the z-y plane with respect to the second coupling block 521 by operation of the second rotary motor 523. A third rotation module 530 includes a third coupling block 531 configured to have a third coupling portion 531a protrusively formed at one side thereof, a third coupling block 531 rotatably coupled to a cut portion of the third rotation member 532, and a third rotary motor 533 provided at one surface of the third rotation member 532. The second rotation module 520 is assembled to the third rotation module 530 by bolting the third coupling portion 531a of the third rotation module 530 to the second coupling plate 524a of the second rotation module 520. In this case, a rotary shaft of the third rotary motor 533 is installed in the direction parallel with the x-axis with respect to the third rotation member 532, and the third rotation member 532 rotates along the plane parallel with the z-y plane with respect to the third coupling block 531 by operation of the third rotary motor 533. A fourth rotation module 540 includes a fourth coupling block 541 configured to have a fourth coupling portion 541a protrusively formed at one side thereof, a fourth coupling block 541 rotatably coupled to a cut portion of the fourth rotation member 542, and a fourth rotary motor 543 provided at one surface of the fourth rotation member 542. The third rotation module 530 is assembled to the fourth rotation module 540 by bolting the fourth coupling portion 541a of the fourth rotation module 540 to one side of the third rotation member 532 of the third rotation module 530. That is, the fourth coupling portion 541a of the fourth rotation module 540 is bolted to one side of the third rotation member 532 without a separate third arm. In this case, a rotary shaft of the fourth rotary motor 543 is installed in a direction perpendicular to the x-axis with respect to the third rotary motor 533, and thus the fourth rotation member 542 rotates in a direction perpendicular to the rotational direction of the third rotation member 532. For example, in the embodiment shown in FIG. 3, the third rotation member 532 rotates along the plane parallel with the z-y plane, whereas the fourth rotation member 542 rotates along the plane parallel with the z-x plane. Of course, the plane along which the fourth rotation member 542 rotates is varied according to rotation of the third rotation member 532. The other side of the fourth rotation member 542, which is an opposite side at which the third rotation member 532 is provided, has an end effector 544 as a fastening portion to which an auxiliary device, for example, an inspection device such as a CCD camera (not shown), a processing device such as an end mill 545 (see FIG. 5), or a welding device such as a torch (not shown) is fastened. That is, the end effector 544 is provided at an end of a rotation module which is lastly assembled when the rotation unit 500 is configured by assembly of the plurality of rotation modules. The last rotation module is preferably configured to rotate in a direction perpendicular to the rotational directions of the other rotation modules. For example, when the rotation unit 500 is configured by two rotation modules in order to be used in a limited space, a worker may easily remove an unnecessary rotation module by releasing bolts fastened to the rotation module. In this case, the rotary shaft of the second rotation module 520 may be arranged in a direction perpendicular to the rotary shaft of the first rotation module 510. As an alternative example, when the rotation unit 500 is configured by three rotation modules, the rotary shaft of the third rotation module 530 may be arranged in a direction perpendicular to the rotary shafts of the first and second rotation modules 510 and 520. Meanwhile, when the rotation unit 500 is configured to have more rotary shafts by an increase of the number of rotation modules, the number of rotary shafts of the rotation unit 500 may be easily increased in such a manner that one side of an existing rotation module is disassembled by releasing of bolts and another prepared rotation module is assembled between existing rotation modules or to one side of an existing rotation module by bolting. FIG. 5 is a view illustrating a use state of the articulated manipulator according to an exemplary embodiment. FIG. 5 shows an example of repairing a defect of a welding portion of the partition plate installed in the channel head of the steam generator using the articulated manipulator of the present invention. Hereinafter, an operation of the articulated manipulator 100 according to the exemplary embodiment will be described with reference to FIGS. 3 to 5. First, the articulated manipulator 100 is installed at a place requiring inspection or repair work of a defect portion. In this case, the articulated manipulator 100 is transported into the channel head 13 through a manway (not shown) thereof and installed in the channel head 13. The articulated manipulator 100 may also be transported into the channel head 13 after assembly thereof is completed at an external location, and may also be immediately assembled in the channel head 13 in a manner of adding or subtracting the number of rotation modules as necessary. The articulated manipulator 100 is installed in such a manner that the support plate 600 is adhered to and/or supported by one side of the partition plate 14, the opposite edge of the base plate 200 is supported by the inner wall of the channel head 13, and then the jaws 720 provided at the upper ends of the support members 700 are inserted into the through-holes of the tube sheet 12 by driving of the actuation cylinders 710. When the support members 700 are extended by the actuation cylinders 710, the upper ends of the support members 700 press and push up against the bottom surface of the tube sheet 12. Since the tube sheet 12 is securely fixed to the upper side of the channel head 13, the lower end of the support plate 600 is pressed toward the bottom surface of the channel head 13 by the reaction thereof and the opposite edge of the base plate 200 is pressed toward the inner wall of the channel head 13 so that the base plate 200 is securely supported. Although the example in which the support plate 600 is vertically installed to the partition plate 14 has been described in the embodiment shown in FIG. 5, the installation form of the base plate 200 including the support plate 600 may be freely changed according to a type of work such as positioning, inspection, or repair of the defect portion. For example, the support plate 600 may also be horizontally installed to the partition plate 14. Next, the end effector 544 at the end of the rotation unit 50 is located at the defect portion. In this case, the position of the end effector 544 is displaced in such a manner that the movable block 300 linearly moves along the base plate 200, the rotatable block 400 horizontally rotates on the movable block 300, and each rotation module vertically rotates. The position displacement of the end effector 544 is remotely controlled by a controller (not shown) at the outside of the steam generator, and various motors required for the position displacement is driven by power supply through a power cable (not shown) connected to the outside. In order to grasp the progress of defect inspection or repair work of the welding portion, a CCD camera may be installed to the end effector 544. In this case, the CCD camera is preferably configured such that coordinate data is extractable through a vision image processing function, and the controller calibrates a position coordinate using the extracted coordinate data during the inspection or repair work. In addition, when machining (for instance, grinding) of the defect portion is required to repair the welding portion, a processing device such as an end mill 545 may be installed to the end effector 544. In this case, since the base plate 200 is securely supported in the channel head 13 by the reaction transferred through the support members 700, it may be possible to enhance accuracy and efficiency of the work by preventing movement of the base plate 200 due to the reaction generated during processing even when the processing device is installed to the end effector 544 such that the defect portion is machined. A reaction sensor (not shown) is preferably installed at one side of the processing device or one side of the end effector 544. In this case, when the reaction applied during machining exceeds a preset value, it may be possible to prevent damage of the steam generator by controlling the end effector 544 or the processing device such that the feed speed of the end effector 544 moving along the welding portion or the rotational speed of the processing device such as the end mill 545 is reduced. The articulated manipulator 100 of the exemplary embodiment may be utilized at inspection and repair work of sealed welding portions of the heat transfer tubes 11 (see FIG. 1), and may be used to locate an inspection device or a repair device at a desired point in a limited place such as the pipe of the nuclear power plant in which access of the work is limited. In addition, since each component is configured in a modular manner to be easily assembled and disassembled, it may be possible to cope with conditions such as an installation place, a position of a defect portion, and a type of required work by actively changing the overall configuration of the articulated manipulator 100. That is, the various rotation modules may be easily assembled to or disassembled from each other by fastening or releasing of bolts in a state in which the respective rotation modules for configuring the rotation unit 500 are separately manufactured and prepared. Therefore, the articulated manipulator 100 having the proper number of rotary shafts may be configured according to an installation space. In addition, the articulated manipulator 100 is linearly moved by the movable block 300, horizontally rotated by the rotatable block 400, and vertically rotated by the plurality of rotation modules which is independently and rotatably operated. Therefore, the end effector 544 may perform the accurate inspection and repair work while smoothly moving along the curved surface. As is apparent from the above description, an articulated manipulator according to exemplary embodiments may be easily moved and installed in a limited place such as the inside of a channel head or pipe of a steam generator. In addition, it may be possible to locate an inspection device or a repair device at various positions in different spaces by selectively assembling a plurality of rotation modules as necessary, and the manipulator may be easily transported, assembled, and disassembled since joints thereof are assembled in a modular manner by bolting. In addition, since any failed rotation module of the rotation modules may be replaced with another prepared rotation module, maintenance is easily performed. In addition, since a base plate is securely supported by a support member, repair work may be accurately performed by preventing movement of a processing tool when a defect portion is processed. Furthermore, a welding portion may be prevented from being damaged during the repair work by controlling an operation of the processing tool according to a processing reaction measured by a reaction sensor. While the present disclosure has been described with respect to the specific embodiments, it will be apparent to those skilled in the art that various changes and modifications may be made without departing from the spirit and scope of the invention as defined in the following claims. |
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051981815 | abstract | Particles including fusible nuclei and electrons are contained in a fusion reaction vessel having a conductive length. The particles individually have a mass and a velocity, and are resonated by a weak magnetic field applied to the vessel at a magnetic flux density set according to a relation equating the gravitational energy of the particles with the electromagnetic energy of the applied magnetic field. The magnetic field can be applied in addition to much stronger confinement and heating magnetic fields. The flux density B of the applied field is calculated according to the equation mc.sup.2 =Bvl coulomb, where m=the mass of a particle, c=the velocity of light, v=the mean orbital velocity of the earth about the sun, and l=the plasma circumference. At least two distinct particle masses are contained in the plasma, and distinct magnetic fields can be applied to the plasma at flux densities calculated according to said equation mc.sup.2 =Bvl coulomb for at least said two distinct masses. The field is applied using electromagnet coils substantially encompassing the plasma body, in particular by poloidal coils encompassing the toroidal confinement vessel of a Tokamak type reactor. |
abstract | A waste treatment process includes containing a reactant metal alloy (210) in a reactant alloy container (202) substantially isolated from oxygen gas. The reactant metal alloy includes at least one chemically active alkaline metal and at least one radiation absorbing metal. After heating the reactant alloy (210) in the reactant alloy container (202) to a desired operating temperature, a waste material including radioactive isotopes to be alloyed is introduced into the molten alloy, preferably below the surface of the alloy. Non-radioactive compounds in the waste material react with metals in the reactant alloy (210) to produce useful halogen salts and other materials. The metal radioactive isotopes in the waste material are alloyed with the alkaline metal and radiation absorbing metals to create a storage product for long term storage. |
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049833502 | abstract | The present invention seeks to obtain reliable detection of the fall of a control cluster into the core of a pressurized water nuclear reactor without giving rise to the risk of an erroneous detection causing an untimely emergency reactor stop. Four acquisition units (U1, U2, U3, U4) corresponding to the four quadrants of the reactor core receive neutron flux signals (F1), heat flux signals (P1), and cluster position signals (Z1A, Z1B, . . . ) from the various quadrants and process them to generate corresponding alarm signals. These units, together with a combination circuit (14, 18, 22), provide a cluster fall signal (24) in the presence of at least two alarm signals including at least one cluster position alarm signal. The invention is applicable to electricity power generation. |
claims | 1. An optical system, comprising:an illumination system having:a field plane in which a mask is located;a conjugated plane, conjugated to said field plane, and having a field stop situated therein, wherein said conjugated plane is situated in a light path from a light source to said field plane, before said field plane, so that light from said light source traverses through said conjugated plane;a first optical element having a first raster element in said light path before said conjugated plane;a second optical element having a second raster element in said light path after said first optical element; andat least a first mirror, in said light path, after said conjugated plane, for imaging a field in said conjugated plane into an image in said field plane. 2. The illumination system of claim 1, further comprising:a second mirror for imaging said field into said image, wherein light rays travel along a light path from said conjugated plane to said field plane; andan arc-shaped field in said field plane, whereby said arc-shaped field has a radial direction in a middle of said arc-shaped field defining a scanning direction,wherein said first mirror and said second mirror are arranged in said light path in such a position and having such a shape, that an edge sharpness of said arc-shaped field is smaller than 5 mm in said scanning direction,wherein said light rays are impinging on said first mirror and said second mirror with incidence angles less than or equal to 30° or greater than or equal to 60° relative to a surface normal of said first mirror and second mirror. 3. The illumination system according to claim 2, wherein said edge sharpness is smaller than 2 mm. 4. The illumination system according to claim 2, wherein the edge sharpness is smaller than 1 mm. 5. The illumination system according to claim 2, wherein said incidence angles relative to said surface normal are ≦20° or ≧70°. 6. The illumination system according to claim 2, wherein said illumination system includes an object field, and wherein said object field is arc-shaped. 7. The illumination system according to claim 2, wherein said first mirror and said second mirror are part of an imaging system and the imaging system has a magnification ratio unequal to 1. 8. The illumination system according to claim 7, wherein said imaging system is a non-centered system. 9. The illumination system according to claim 2, wherein said first mirror and said second mirror are part of an imaging system and said imaging system comprises an exit pupil and an aperture stop that is located on or close to a plane conjugated to said exit pupil. 10. The illumination system according to claim 2, wherein said first mirror and said second mirror are part of an imaging system, said imaging system comprises an exit pupil, and said first mirror is positioned close to a plane conjugated to the exit pupil. 11. The illumination system according to claim 2, wherein said first mirror and said second mirror are aspheric mirrors. 12. The illumination system according to claim 2, wherein said first mirror is a concave mirror having a nearly hyperbolic form or a nearly elliptic form, and wherein said first mirror defines a first axis of rotation. 13. The illumination system according to claim 12, wherein said second mirror is a concave mirror having a nearly hyperbolic form or a nearly elliptic form, and wherein said second mirror defines a second axis of rotation. 14. The illumination system according to claim 13,wherein said first axis of rotation and said second axis of rotation subtend to an angle γ, andwherein said first mirror and said second mirror define a first magnification for a chief array traveling through a center of said field and a center of an exit pupil, a second magnification for an upper COMA ray traveling through said center of the field and an upper edge of said exit pupil and a third magnification for a lower COMA ray traveling through said center of the field and a lower edge of said exit pupil,whereby the angle γ between said first axis of rotation and said second axis of rotation is chosen so that said first, said second and said third magnification is nearly identical. 15. The illumination system according to claim 2, wherein said first mirror and said second mirror have a used area, and wherein said used area is arranged off-axis with an axis of rotation of said first mirror and said second mirror. 16. The illumination system of claim 2, wherein at least one of said first mirror or said second is an aspheric mirror. 17. The illumination system according to claim 1, further comprising a field-forming optical component in said light path, after said conjugated plane, that shapes said image. 18. The illumination system according to claim 1, further comprising a field-forming optical component in said light path, before said conjugated plane, that forms said field. 19. The illumination system according to claim 17, wherein said field forming optical component is situated in a light path from the light source to said field plane downstream from said conjugated plane to said field plane. 20. The illumination system according to claim 1, wherein said first mirror is part of a multi-mirror system. 21. The illumination system according to claim 1, wherein said first raster element is a field facet and said second raster element is a pupil facet for directing said light from said light source to said conjugated plane. 22. An EUV-projection exposure unit for microlithography, comprising:the illumination system according to claim 1, wherein said mask is arranged on a carrier system in said field plane; anda projection objective for imaging said mask onto a light-sensitive object on a carrier system. 23. The EUV-projection exposure unit according to claim 22, wherein the unit is designed as a scanning system. 24. A process for producing microelectronic devices with the EUV-projection exposure unit according to claim 23. 25. The illumination system of claim 1, wherein said source emits light having a wavelength of less than or equal to 193 nm. |
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summary | ||
abstract | A method and apparatus for testing semiconductors according to various aspects of the present invention comprises a test system comprising composite data analysis element configured to analyze data from more than one dataset. The test system may be configured to provide the data in an output report. The composite data analysis element suitably performs a spatial analysis to identify patterns and irregularities in the composite data set. The composite data analysis element may also operate in conjunction with a various other analysis systems, such as a cluster detection system and an exclusion system, to refine the composite data analysis. The composite may also be merged into other data. |
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046831134 | summary | BACKGROUND OF THE INVENTION The present invention relates to a nuclear fuel assembly and, more particularly, to a nuclear fuel assembly suitable for use for a boiling water reactor. The reactor core of a boiling water reactor is charged with a plurality of nuclear fuel assemblies each of which is constituted by a channel box, lower tie plate, upper tie plate and a multiplicity of fuel rods. The fuel rods arranged in the form of a bundle are held at their upper and lower ends by the upper and lower tie plates. The bundle of the fuel rods is disposed in a channel box secured to the upper tie plate. Each fuel rod is charged with a multiplicity of fuel pellets (UO.sub.2 pellets). In each fuel assembly, there are several fuel rods having UO.sub.2 pellets which contain gadolinea as a burnable poison. There also are two water rods disposed in the central region of the nuclear fuel assembly. In general, a boiling water reactor exhibits a void distribution in the vertical or axial direction. Due to the variation of void reactivity along the axis, the boiling water reactor shows such a power distribution that the peak of the power is shifted to the lower side along the axis. In order to attain a flat axial power distribution by obviating the axially downward shifting of the power peaking, it has been proposed to use a fuel assembly having different degrees of enrichment at the upper and lower regions thereof. One of such a fuel assembly is disclosed in the specification of the U.S. Pat. No. 4,229,258. In this fuel assembly, some of the fuel rods arranged in the peripheral region thereof have different degrees of enrichment at their upper and lower regions. More specifically, the upper region of each of such fuel rod has an enrichment which is about 15% higher than that in the lower region thereof. In recent years, various studies have been made for the development of fuel assemblies suitable for higher burn-up, i.e., fuel assemblies which can be burnt up to a high degree. Such a fuel is obtained by arranging fuel rods rich in fissile material, i.e. fuel rods having high enrichment, in the vicinity of the channel box having a high density of thermal neutron flux. Japanese Patent Laid-Open No. 26292/1983 discloses a fuel assembly which can be burnt up to a high degree in accordance with the theory disclosed in U.S. Pat. No. 4,229,258. SUMMARY OF THE INVENTION Accordingly, an object of the invention is to provide a fuel assembly which is improved such that the difference in the power level between different cross-sections along the axis is minimized so as to flatten the power distribution in the axial direction of the fuel assembly. Another object of the invention is to provide a fuel assembly having a simple construction constituted by a fewer number of kinds of the fuel rod. Still another object of the invention is to provide a fuel assembly capable of improving the fuel economy. To these ends, according to the invention, there is provided a fuel assembly having first fuel rods each containing a burnable poison over the substantial region in the axial direction thereof, and second fuel rods containing no burnable poison, each of said first fuel rods having such an enrichment distribution that the mean enrichment over the most part of the upper region thereof is higher than the mean enrichment over the most part of the lower region thereof, while each of said second fuel rods has a substantially uniform enrichment distribution over the entire axial region thereof. |
abstract | A nuclear fuel rod plenum spring assembly that has a spacer affixed to the lower end of the ground torsion spring. The spacer has a substantially flat surface on its underside that presses against the upper surface of the upper fuel pellets to spread the load of the spring over the top surface of the upper most fuel pellet. |
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043893694 | summary | BACKGROUND OF THE INVENTION The invention described herein relates to nuclear reactor fuel assemblies and more particularly to a fuel assembly grid designed to improve reactor performance and to be manufactured at a cost less than conventional grids. Commercial nuclear reactors used for generating electric power include a core composed of a multitude of fuel assemblies which generate heat used for electric power generation purposes. Each fuel assembly includes an array of fuel rods and control rod guide tubes held in spaced relationship with each other by grids of egg-crate configuration spaced along the fuel assembly length. The fuel rods may be approximately 0.5 inch in diameter and about 14 feet long thus requiring a number of supporting grids along their length. Each grid includes interwoven Inconel or Zircaloy straps which form multiple cells, each cell having springs on two adjacent walls and a pair of projections or dimples on each of the other two walls forming a cell. The springs laterally impress resistive forces on each fuel rod in the assembly. Although this fuel assembly design performs exceptionally well in a nuclear reactor, one disadvantage inherent in the design is that the the inwardly projecting springs and dimples occasionally mar or score the surface of fuel rods during the time they are being pulled into the fuel assembly grids. In carrying out this fuel rod loading operation, the grids are held immovably in position while a longitudinal steel rod attached to the end of a fuel rod pulls it axially through the aligned openings or cells in the grids. As the rod engages the springs and dimples in the grid cells, their edges engage the exposed relatively soft surface of the moving fuel rod and, in some cases, score its surface sufficiently deep as to cause the rod to fall outside established fuel rod surface specifications. Also, the grid strap material from which the springs and dimples are formed has a high neutron capture cross section, particularly when made of Inconel. Although annealed Zircaloy is not as deleterious, to some extent it adversely affects reactor performance and efficiency. The best balance between material stiffness and low neutron capture cross section should be reached for efficiency purposes. SUMMARY OF THE INVENTION Briefly stated, in accordance with the teachings of this invention, the above disadvantages are overcome by providing an improved design of fuel assembly grid wherein the grid body comprises two separate sets of interwoven straps, each formed to egg-crate configuration and positioned vertically with respect to each other. One set of interwoven straps is placed on top the other and both sets are held against longitudinal or radial movement by a peripheral strap to which the ends of the straps are welded. The square openings thus formed in the juxtaposed sets of straps are aligned with each other to receive fuel rods which extend axially therethrough. Each cell which comprises the aligned square openings on the strap sets, contains a pair of spaced dimples on each of two adjacent straps facing each cell and are arranged to provide lateral support to a fuel rod. The force applied laterally to a fuel rod in each cell is provided by a spring arranged to extend diagonally across the grid with a portion in each cell designed to engage a side of a fuel rod opposite from the dimples. It will occur to those skilled in the art that the height of the grids can be reduced by virtue of the elimination of spring members heretofore found in the middle of each grid strap. |
description | The present invention relates to a target device for a neutron generating device, an accelerator-excited neutron generating device, and a beam coupling method thereof. Heat exchange between a cooling medium and a solid target material is main factor that restricts the development of the neutron generating device. With the continual increase in the beam power of the accelerator, the solid target has been unable to meet the requirements for operation of the target. The current solid targets generally operate under the conditions of less than one MW beam coupling. The present invention has been made in view of the above problems. An object of the present invention is to provide a target device for a neutron generating device, an accelerator-excited neutron generating device, and a beam coupling method thereof, thereby solving the technical problem of for example selection of a movable thermal medium and removing of heat when beam bombardment is carried out. Another object of the present invention is to provide a beam coupling method of an accelerator-excited neutron generating device, thereby solving for example the technical problem that the existing target devices cannot be used in a high-power beam for a long period of time. In accordance with an aspect of the present invention, there is provided a target device for a neutron generating device, the target device comprising a plurality of solid particles serving as a target body; and a target body reaction chamber for accommodating the solid particles. In accordance with an aspect of the present invention, the solid particles have at least one of a spherical shape, an ellipsoidal shape, and a polyhedral shape. In accordance with an aspect of the present invention, the solid particle comprises anyone of tungsten, a tungsten alloy, uranium, a uranium alloy, uranium ceramics, thorium, a thorium alloy, and thorium ceramics. In accordance with an aspect of the present invention, the target body reaction chamber has an injection conduit defining an injection opening and a discharge conduit defining a discharge opening, and the solid particles are injected into the target body reaction chamber through the injection opening and moved out of the target body reaction chamber through the discharge opening. In accordance with an aspect of the present invention, a ratio of a diameter of the target body reaction chamber to a particle diameter of the plurality of solid particles is in a range of 5:1-30:1; and/or a ratio of a caliber of the injection conduit to the diameter of the target body reaction chamber is in a range of 1:1-1:10; and/or a ratio of a caliber of the discharge conduit to the diameter of the target body reaction chamber is in a range of 1:1-1:10. In accordance with an aspect of the present invention, there is provided an accelerator-excited neutron generating device comprising a target body reaction chamber for accommodating solid particles serving as a target body; and a solid particle conveying device for injecting the solid particles into the target body reaction chamber. In accordance with an aspect of the present invention, the accelerator-excited neutron generating device further comprises a cooling device, wherein the solid particles are cooled by the cooling device after the solid particles are moved out of the target body reaction chamber, and then the solid particles are injected into the target body reaction chamber by the solid particle conveying device. In accordance with an aspect of the present invention, the accelerator-excited neutron generating device further comprises a sorting device configured such that those of the solid particles that conform to a predetermined standard are selected from the solid particles by the sorting device after the solid particles are moved out of the target body reaction chamber, and then injected into the target body reaction chamber. The predetermined standard may be a particle size of the solid particles, other quality indexes, or the like. The indexes can be detected by means of various detection devices. In accordance with an aspect of the present invention, the solid particles have at least one of a spherical shape, an ellipsoidal shape, and a polyhedral shape. In accordance with an aspect of the present invention, the accelerator-excited neutron generating device further comprises a buffer chamber disposed at at least one of a solid particle injection opening and a solid particle discharge opening for temporarily storing the solid particles. In accordance with an aspect of the present invention, the solid particle conveying device is configured to circulate the solid particles from an inside of the target body reaction chamber through an outside of the target body reaction chamber to the inside of the target body reaction chamber while the beam is applied to the solid particles. In accordance with an aspect of the present invention, the accelerator-excited neutron generating device further comprises a cooling device and a sorting device, wherein the solid particles which are being circulated and situated outside the target body reaction chamber are cooled by the cooling device, and those of the solid particles that conform to a predetermined standard are selected by the sorting device from the solid particles which are being circulated and situated outside the target body reaction chamber. In accordance with an aspect of the present invention, there is provided a beam coupling method for an accelerator-excited neutron generating device, the method comprising: injecting solid particles serving as a target body into a target body reaction chamber, and applying a beam to the solid particles. In accordance with an aspect of the present invention, circulating the solid particles from an inside of the target body reaction chamber through an outside of the target body reaction chamber to the inside of the target body reaction chamber while the beam is applied to the solid particles. In accordance with an aspect of the present invention, the solid particles which are being circulated and situated outside the target body reaction chamber are processed. In accordance with an aspect of the present invention, the processing comprises cooling the solid particles and selecting those of the solid particles that conform to a predetermined standard from the solid particles. In accordance with an aspect of the present invention, the beam coupling method may be a low-power beam coupling method for an accelerator-excited neutron generating device, or a high-power beam coupling method for an accelerator-excited neutron generating device. According to some embodiments of the present invention, the present invention can solve the exiting technical problem of removing heat from a target device during high-power beam bombardment, thereby achieving the advantages of high-efficiency heat exchange, a long life time, a good stability and a wide application range. Other features and advantages of the application will be set forth in part in the description which follows and, in part, will be obvious from the description, or may be learned by practice of the application. The present invention will be further described in detail by means of the accompanying drawings and embodiments. The reference numerals in the drawings are listed as follows: 11: target body reaction chamber; 12: solid particle injection section; 13: solid particle discharge section; 14: proton beam input conduit; 15: beam scanning and security protecting device; and 16: solid particle. The preferable embodiments of the present invention will now be described with reference to the accompanying drawings. It should be understood that the preferable embodiments set forth herein are only used to describe and explain the present invention and the present invention is not limited to the embodiments. Device Embodiment As shown in FIGS. 1-3, an accelerator-excited neutron generating device 100 according to an embodiment of the present invention comprises a first subsystem S1, a second subsystem S2, a third subsystem S3, a fourth subsystem S4, and a fifth subsystem S5. The first subsystem S1 is an accelerator system for providing an accelerator incident beam. The distribution of the beam may be set to a uniform distribution, a Gaussian-like unimodal distribution or bimodal distribution, or the like according to requirements. The energy level of the particle beam may be in a range of 100 MeV to several GeV. The beam intensity is dependent upon a neutron intensity required for a neutron source. The beam intensity may be any value between 0.1 mA and several hundred mA. The first subsystem S1 may comprise components such as a beam bending magnet, and a beam vacuum protection system. Any known accelerator system may be adopted as the accelerator system serving as the first subsystem S1. The second subsystem S2 is an target device 100A for the neutron generating device (shown in FIGS. 4 and 5). The target device 100A composes a target body reaction chamber 11 and solid particles 16. The target body consisting of solid particles 16 is bombarded with the incident beam so that primary spallation reaction and cascade reaction occurs in the reaction chamber 11 to generate a great deal of neutrons. At one end of the target body reaction chamber 11, a solid particle injection section 12 is disposed. The solid particle injection section 12 may be coupled with the third subsystem S3 for temporarily storing the solid particles 16 so that the solid particles 16 will be inputted from the third subsystem S3 into the solid particle injection section 12. At the other end of the target body reaction chamber 11, a solid particle discharge section 13 is disposed. The solid particle discharge section 13 may be coupled to the fourth subsystem S4. The fourth subsystem S4 is used for temporarily storing the solid particles 16. The third subsystem S3 may comprise a solid particle conveying device and a sorting device. In the third subsystem S3, the solid particles will be inspected and will be separated according to conditions. The fourth subsystem S4 is a system for storing the solid particles temporarily and may comprise a solid particle conveying device, a heat exchanger and the like. In the fourth subsystem, the solid particles will be cooled by the heat exchanger. The fifth subsystem S5 is a neutron applying system and can receive neutrons transmitted from the second subsystem S2 and perform various possible applications of the neutrons such as power generation. Particles of the incident beam may be selected from particles such as protons, deuterium ions, tritium ions, and helium ions. An incident direction of the beam is the same as an axial direction of the chamber. The main part of the target body is the solid particles. A metal material such as tungsten or a tungsten alloy, or uranium, a uranium alloy or uranium ceramics, or thorium, thorium alloy or thorium ceramics may be selected for the main material of the solid particles according to requirements for spallation neutrons to generate more neutrons. The solid particles have at least one of a spherical shape, an ellipsoidal shape, and a polyhedral shape. The solid particles may have any other appropriate shape. As shown in FIGS. 3-5, the accelerator-excited neutron generating device 100 according to the embodiment comprises the target body reaction chamber 11 for accommodating the solid particles, an accelerator system and security protecting device 15, a proton beam input conduit 14 connected between a central portion of the target body reaction chamber 11 and a central portion of the accelerator system and security protecting device 15, and an off-line processing system C cooperating with the target body reaction chamber 11. The off-line processing system C comprises a solid particle conveying device for injecting the solid particles into the target body reaction chamber. As shown in FIGS. 1-2, the third subsystem S3 and the fourth subsystem S4 shown in FIG. 1 constitute the off-line system C shown in FIG. 2. The system is used for circulating the solid particles 16 from an inside of the target body reaction chamber through an outside of the target body reaction chamber to the inside of the target body reaction chamber. The off-line processing system C may further comprise a housing CH. As shown in FIGS. 3-5, the solid particle conveying device comprises a first solid particle conveying device C1 disposed substantially horizontally. The solid particles 16 may be discharged from the target body reaction chamber 11 to fall onto the first solid particle conveying device C1. The off-line processing system C may further comprise a cooling device 30. The solid particles are cooled by the cooling device after the solid particles are moved out of the target body reaction chamber, and then the solid particles are injected into the target body reaction chamber by the solid particle conveying device. For example, the cooling device 30 is an air blower to cool the solid particles 16. In addition, the first solid particle conveying device C1 may also be submerged in liquid, thereby cooling the solid particles 16 while the solid particle conveying device C1 conveys the solid particles 16. Furthermore, the off-line processing system C may further comprise a sorting device, such as a drum screen, for screening out cracked solid particles. The solid particle 16 may be discharged from the target body reaction chamber 11 and fall into the sorting device. The sorting device is configured such that those of the solid particles that conform to a predetermined standard are selected from the solid particles by the sorting device after the solid particles are moved out of the target body reaction chamber 11, and then injected into the target body reaction chamber 11. The predetermined standard may be a particle size of the solid particles, other quality indexes, or the like. The indexes can be detected by means of various detection devices. In addition, the cooling device 30 and the sorting device may be disposed at any appropriate portions of the off-line processing system C. If the target body reaction chamber 11 is not uprightly disposed, for example, the target body reaction chamber 11 is horizontal disposed, an upright conveyer such as a screw conveyer or a pickup device (for example, a mechanical arm or a robot) may be used to load the solid particles 16 to the first solid particle conveying device C1 or the sorting device. In addition, in some embodiments, the neutron generating device may not include the cooling device and the sorting device. For example, heat may dissipate from the solid particles on the conveying device, and less cracked or fragmentized particles are generated in the solid particles. In addition, when the neutron generating device operates at a low power, the neutron generating device may not include the cooling device. As shown in FIGS. 2-3, the solid particle conveying device further comprises a second solid particle conveying device C3 as an upright lifting and conveying device. The second solid particle conveying device C3 comprises a screw conveyer C31 for upright conveying the solid particles, a horizontal conveying device C32, and a pneumatic lift C33. The solid particle conveying device further comprises a third solid particle conveying device C5. The first solid particle conveying device C1, the horizontal conveying device C32, and the third solid particle conveying device C5 may be any appropriate conveyers such as a chute for conveying the solid particles with gravity, a scraper conveyor, or a belt conveyor. In addition, the solid particles may be discharged from the target body reaction chamber 11 with gravity and may fall directly onto the first solid particle conveying device C1. The target body reaction chamber 11 may have a substantially cylindrical shape or any other appropriate shape. For example, the target body reaction chamber may have an elliptical cross section. An included angle between an axis of the target body reaction chamber 11 and a horizontal plane may be larger than 0 degree and less than or equal to 90 degrees. In other words, the target body reaction chamber may be disposed generally horizontally or may be disposed vertically. A ratio of an axial or longitudinal size of the target body reaction chamber 11 to a radial size (diameter) or a transverse size of the target body reaction chamber is in a range of 3:1-10:1, and can be adjusted according to the beam intensity. A ratio of a diameter or particle diameter of each of the solid particles to the radial size (diameter) or the transverse size of the target body reaction chamber is in a range of 5:1-30:1. A ratio of a caliber of an injection conduit 12 of each of a plurality of solid particle injection sections to the radial size (diameter) or the transverse size of the target body reaction chamber is in a range of 1:1-1:10. A ratio of a caliber of a discharge conduit 13 of a solid particle discharge section to the radial size (diameter) or the transverse size of the target body reaction chamber is in a range of 1:1-1:10. The material of the solid particles can be selected from at least anyone of tungsten, a tungsten alloy, uranium, a uranium alloy, uranium ceramics, thorium, a thorium alloy, and thorium ceramics. The diameter or particle size of the solid particle is between an order of submillimeter and an order of several centimeters. The specific diameter or particle size of the solid particle is determined according to the radial size of the target body reaction chamber. The accelerator system is used to control a transverse distribution of the beam bombarding the solid particles. The accelerator system scans in a predetermined manner so that heat is distributed uniformly. The security protecting device 15 is used for preventing leakage of pressurized gas in the direction of the accelerator in an accidental state and ensures that the accelerator beam conduit is in a vacuum range for operation. As shown in FIGS. 2-3, the off-line system C comprises a buffer chamber C7 disposed at a corresponding injection opening 12 of the solid particle injection section. The sorting system is configured such that before the solid particles are injected, they are inspected and separated by the sorting system according to preset conditions. The qualified solid particles are conveyed to the buffer chamber C7. For example, the solid particles having a predetermined particle size is selected and the cracked solid particles are screened out. The screw conveyer C31 has the function of both sorting and dust removing and the pneumatic lift C33 has the function of both sorting and dust removing. The buffer chamber C7 may have substantially a funnel shape and is used for temporarily storing the solid particles. In the accelerator-excited neutron generating device according to the above embodiment, the target body in the target body reaction chamber 11 is composed of the solid particles. When the target body is bombarded with the beam, it will generate a high-flux neutron flow. The material of the solid particles is selected from tungsten, a tungsten alloy, uranium, a uranium alloy, uranium ceramics, thorium, a thorium alloy, thorium ceramics, and other materials having required characteristic. The characteristic of the material of the solid particles is high-temperature resistance, high heat capacity, high neutron yield and the like. The solid particles are placed in the target body reaction chamber. The solid particles enter the solid particle injection section through the buffer chamber and are injected into the reaction chamber from the injection section. The solid particle discharge section functions to move the solid particles in which high-power energy is deposited from the reaction chamber and convey the solid particles to the heat exchange system for removing the high-power deposited energy. The heat exchange system for the solid particles functions to remove the high-power deposited energy from the solid particles. The solid particle conveying device and the sorting system are disposed between the solid particle injection section and the heat exchange system. In the solid particle conveying device and the sorting system, the solid particles will be inspected and will be separated according to conditions. The beam input conduit functions to provide an input conduit for the beam from the accelerator and guide the beam into the device. Particles of the incident beam may be particles such as protons, deuterium ions, tritium ions, or helium ions. An incident direction of the beam is the same as an axial direction of the chamber. In the beam scanning and security protecting device, the beam scanning device can control a transverse distribution of the beam bombarding the solid particles, and scan in a predetermined manner so that heat is distributed uniformly; and the security protecting device can prevent leakage of pressurized gas in the direction of the accelerator in an accidental state of the device and ensure that the accelerator beam conduit is in a vacuum range for operation. At both ends of the target body reaction chamber, the solid particle injection section and the solid particle discharge section are disposed, respectively. The target body reaction chamber is provided with a plurality of solid particle injection section ports or interfaces. A user can select the number of the ports or interfaces to carry out connection according to specific requirements. The buffer chamber C7 may be an annular chamber fitted over an outer periphery of the beam conduit. The injection openings of the target body reaction chamber 11 may be initially closed. The third solid particle conveying device C5 such as a belt conveyor first injects the solid particles into the buffer chamber C7. The solid particles are poured into the buffer chamber C7 so that the level of the solid particles reaches a certain height, and then the injection opening of the target body reaction chamber 11 is opened. The number of the opened injection openings of the target body reaction chamber 11 is determined according to parameters such as the beam intensity and can be controlled during operation. After the injection opening of the target body reaction chamber 11 is opened, the solid particles fall into the injection opening of the target body reaction chamber 11 under the action of gravity, or under the action of gravity and by being driven with a gas. For example, a gas blowing opening may be disposed above the buffer chamber C7 so that the solid particles fall into the injection opening of the target body reaction chamber 11. The speed at which the solid particles are injected and discharged is determined by means of the number of the opened injection openings, thereby ensuring that a predetermined amount of the solid particles are retained in the target body reaction chamber 11. Referring to FIGS. 1, in a specific example of the present invention, an body of the accelerator-excited neutron generating device comprises the target body reaction chamber 11, the solid particle injection section 12, the solid particle discharge section 13, the proton beam input conduit 14, and the beam scanning and security protecting device 15. The target body reaction chamber 11 is connected at a lower portion to the solid particle discharge section 13, and at an upper portion to the solid particle injection section 12. In addition, the proton beam input conduit 14 is also connected to the solid particle injection section 12 and the central portion of the target body reaction chamber 11, and the proton beam input conduit 14 is connected to the beam scanning and security protecting device at a position away from the target body reaction chamber 11. The target device is provided with a plurality of solid particle injection section ports or interfaces. A user can select the number of the ports or interfaces to carry out connection according to specific requirements for injecting the solid particles into the target body reaction chamber. The primary spallation reaction and cascade reaction of the injected solid particles with the beam occur in the target body reaction chamber 11 to generate a great deal of neutrons. The target may be disposed generally horizontally or may be disposed vertically in use. Furthermore, an included angle between an axis of the target body reaction chamber and a horizontal plane is larger than 0 degree and less than or equal to 90 degrees. Referring to FIG. 5, the target body is composed of the target body reaction chamber 11 and the solid particle 16. After the solid particles 16 are injected through the solid particle injection section 12, the primary spallation reaction and cascade reaction of the injected solid particles with the beam occur in the target body reaction chamber 11 to generate a great deal of neutrons. The beam scanning and security protecting device 15 of the target system is mounted to an upper end of the target body reaction chamber, and the beam scanning device can control a transverse distribution of the beam bombarding the solid particles, and performs beam scanning in a predetermined manner so that heat is distributed uniformly. The security protecting device is used for protection of the accelerator system. When gas leaks from the target body in an accidental operating condition, the pressurized gas will be led to the security protecting device through an inlet of the security protecting device. In the security protecting device, the gas is ionized and cooperates with an upper pressure difference system to protect the accelerator system effectively. The beam scanning device is composed of a quadruple magnet system or a deflecting electrode system and can operate at a frequency from tens to hundreds Hz. In addition, the scanning frequency is improved as far as possible. The interface of the beam scanning device is connected to the security protecting device through a conduit. Material of the security protecting device is composed of high-temperature resistant solid material. The beam scanning and security protecting device is integrated. Embodiment on High-power Beam Coupling Method According to an embodiment, based on the accelerator-excited neutron generating device in the above embodiments, the present invention provides a high-power beam coupling method. The method comprises the steps of: injecting solid particles serving as a target body into a target body reaction chamber, and applying a beam to the solid particles. The solid particles are circulated from an inside of the target body reaction chamber through an outside of the target body reaction chamber to the inside of the target body reaction chamber while the beam is applied to the solid particles. The solid particles which are being circulated and situated outside the target body reaction chamber may be processed. The processing comprises cooling the solid particles and selecting those of the solid particles that conform to a predetermined standard from the solid particles. The present invention provides a high-power beam coupling method comprising the following steps. At step S10, the solid particles serving as the target body enter the solid particle injection section through the buffer chamber and are injected into the reaction chamber from the injection section, for example, by gravity or pneumatically. At step S20, nuclear reaction (primary spallation reaction and cascade reaction) of the solid particles with the beam occur in the target body reaction chamber to release a great deal of neutrons. At step 30, high-power energy is deposited in the solid particles by the high-power beam. At step 40, the solid particles in which the high-power energy is deposited are moved out of the reaction chamber through the solid particle discharge section and are conveyed to the heat exchange system for the solid particles so that the high-power deposited energy is also removed from the reaction chamber. At step 50, in the heat exchange system for the solid particles, the high-power deposited energy is removed from the solid particles. Specifically, when the solid balls pass through the sorting system, the solid balls are cooled by gas so that the temperature of the solid balls is decreased. Furthermore, a liquid heat exchange system may also be disposed after the sorting system for further heat exchange so that the temperature of the solid particles can be reduced to a room temperature. At step 60, before the solid particles are injected, the solid particles are inspected through the conveying device and the sorting system and are separated according to conditions, and the qualified solid particles are conveyed to the buffer chamber. Specifically, the cracked solid particles are screened out from the solid balls, discharged from the solid particle discharge section, through a horizontal mechanical conveying belt (or plate) having different holes, while the solid particles are cooled by gas. After that, the sorting and dust removing for the solid particles are further carried out through the screw conveyer and the pneumatic lift. Finally, the qualified solid particles are conveyed into the buffer chamber at the upper end of the target system through a continuous mechanical lift. With the above circulating process, the accelerator-excited neutron generating device can be coupled with the high-power beam. The circulating speed is determined by the beam power and the rate of heat exchange. The more the beam power, the more the circulating speed. Referring to the above device embodiment and the description relevant to FIGS. 1-5 for the structure and characteristic of the accelerator-excited neutron generating device, they are not described here again for the purpose of brevity. The above high-power beam coupling method mainly comprises the steps of injecting the solid particles into the reaction chamber (the target body reaction chamber) through the injection section (the solid particle injection section); generating neutrons by reaction of the solid particles of the target body with the beam; moving the solid particles in which high-power energy is deposited from the reaction chamber through the discharge section (the solid particle discharge section); and removing the high-power energy from the solid particles through a heat exchange system. The beam input conduit (the proton beam input conduit) is an input channel for the accelerator beam. The security protecting device can prevent leakage of pressurized gas in the direction of the accelerator in an accidental state of the accelerator-excited neutron generating device. The off-line processing of the accelerator-excited neutron generating device enables a system including the accelerator-excited neutron generating device to be maintained conveniently and components of the system to have a high technical maturity in the commercial manufacture, thereby reducing difficulty in manufacturing and application of the accelerator-excited neutron generating device. As described above, the accelerator-excited neutron generating device and the high-power beam coupling method according to the above embodiments of the present invention can be applied in the application field of accelerator neutron source, such as a nuclear power energy source system (including transmutation of nuclear waste and breeding of nuclear material), production of isotopes, treating of cancer with neutrons, and irradiation of material. The accelerator-excited neutron generating device and the beam coupling method have the characteristic of off-line processing. For the application with strict radioactivity requirements, tungsten or tungsten alloy may be selected for the material of the solid particles. With the simple off-line processing, simple operational maintenance can be achieved. For the application with low radioactivity requirements, uranium, a uranium alloy, uranium ceramics, thorium, a thorium alloy, or thorium ceramics may be selected for the material of the solid particles. In this case, facility for operational maintenance of the neutron generating device will be complicated. The off-line processing also helps to reduce complicacy. The accelerator-excited neutron generating device and the high-power beam coupling method according to the above embodiments of the present invention have been proposed for the difficult problems such as limiting factors on heat exchange and proportionality coefficient between cooling medium and solid target material in the existing solid target device; and corrosion of the target due to high-temperature liquid metals, damage of structural material of a beam window, stability of two-phase fluid, and difficulty in retaining vacuum of an accelerator beam conduit, which are confronted with by the liquid target. The present invention provides a spallation neutron generating device (accelerator-excited neutron generating device) which is operable under the action of a high-power beam. The accelerator-excited neutron generating device also improves the life time of the target device under the action of a high-power beam. The accelerator-excited neutron generating device is simple in structure, its components are technically mature in the industry, and difficulty in manufacturing the components is greatly reduced. The actual application of the accelerator-excited neutron generating device to a relevant field can be developed. It can be seen from the above contents that in the present invention, a target device for a neutron generating device is configured to include a plurality of solid particles serving as a target body; and a target body reaction chamber for accommodating the solid particles. Thereby, the target body of the target device can be conveniently replaced. In addition, the solid particles are circulated from an inside of the target body reaction chamber through an outside of the target body reaction chamber to the inside of the target body reaction chamber. As a result, the processing of the solid particles such as cooling, inspection and screening of the solid particles can be performed outside the target body reaction chamber. In addition to the above structures, the present invention can adopt any other appropriate conveying devices such as devices for conveying bulk material and articles and industrial robots, any other appropriate cooling devices such as devices for cooling work pieces in the industry, and any other appropriate inspection devices. Therefore, the present invention is not limited to the above embodiments. It should be noted that the above embodiments are only preferable embodiments of the present invention, and the present invention is not limited to the present invention. While the present invention has been described in detail with reference to the above embodiments, it will be understood by those skilled in the art that modifications to the technical solutions of the above embodiments, or equivalent substitutions for some of the technical features of the above embodiments may be made. All of the modifications, equivalent substitutions and improvements made without departing from the principles and spirit of the present invention should fall within the protection scope of the present invention. |
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043953816 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The confinement enclosure with a suction means for leaks according to FIG. 1 comprises a wall consisting of a general frame 1 and an elevated structure 2, all made from reinforced prestressed concrete; the reinforcements are not shown, nor is the nuclear reactor which the enclosure contains. The thickness of the wall of the building is, for example, of the order of 1/20 to 1/10 of the radius. The drainage layer 2a shown by a broken line consists of juxtaposed zones or sub-groups such as 3, 3', 3", each ending in a collector 4, 4', 4"; filters such as 6a, 6b, 6c are provided inside the enclosure and are connected to the network 2a by ventilators such as 21a, 21b, 21c which ventilators place the drainage network under higher pressure compared with the pressure existing in the enclosure itself. The system is therefore a damming containment rather than a drainage system. If no accidents occur, the ventilators function normally at reduced power and, using devices (not shown), the preferential leakage zones can be detected so that they can be repaired. If an accident occurs, the power of the ventilators is increased in order to maintain the network 2a under overpressure relative to the inside and thus prevent any pollution from entering; thus, leaks towards the outside are from a gas which has been filtered previously. If the ventilators 21a, 21b, 21c break down, this filtering still occurs to a large extent, in that the filters 6a, 6b, 6c are much more permeable than the wall separating the interior of the enclosure from the drainage network 2a. The confinement enclosure with the water dam according to FIG. 2 is on the whole identical to that described above, but differs from it in that each drainage region is connected to a water reservoir such as 8 inside the enclosure, at a higher level than the said region. Thus, in the case of an accident with an internal build-up of pressure, the damming water pressure is automatically adjusted. For the upper region of the enclosure, the reservoir may be arranged in a turret 9 which cannot be guaranteed leaktight by the damming liquid itself but which may be specially resistant to prevent any local increase in the risks; the reservoir may also be mounted inside the enclosure and an automatic pump be used to complete the loading. The function of the damming water is to provide a liquid protective wall and to be the first substance to escape, instead of the polluted internal medium, if serious cracking occurs. It also brings about an important additional advantage in that it continuously moistens the concrete in its central part and thus prevents it from shrinking and cracking and keeps it leaktight. The water used is advantageously treated with basic additives to prevent deterioration of the concrete and corrosion of the reinforcements in the case of leakages circulating continuously, but every effort should, of course, be made to ensure that the walls are constructed so as to be perfectly leaktight and to repair them if there is any local deterioration, which is easy to detect by the patches of dampness. In some countries, it may also be advisable to incorporate anti-freeze additives in the water. In normal operation, monitoring the level of the reservoirs is a means of assessing the degree of leaktightness obtained; in the case of accidents, these reservoirs may be topped up with uncontaminated emergency cooling water from the reactor to prevent any radioactivity leak into the atmosphere. Periodic tests for leaktightness could be carried out by putting the networks under high pressure. In the example shown in FIG. 2, the drainage network 2a extends to the frame 1 of the enclosure 2 and communicates, via the duct 4"', with a corresponding water reservoir 8. Such protection of the frame 1 proves very useful in minimising the consequences of a serious accident such as the fusion of the core of a nuclear reactor, when the reinforced concrete frame is the chief obstacle in the path of the molten material coming from the core and tank. In fact, in an accident of this kind, the molten materials first of all progress downwards, spreading out laterally, decomposing the hydrates and liberating water, with reduction of this water by the metals, resulting in the production of hydrogen, and the carbonates with the release of carbon dioxide. A thermal front in the concrete precedes and promotes the chemical attack owing to the thermal stresses, which cause intensive cracking. When this front reaches the region of the channels, which are under water pressure, the cracking causes upward leaks and these cool the concrete, whilst the water evaporates and the vapour, having come into contact with the molten mass, escapes in the form of bubbles, thus causing the concrete to cool. The descent of the molten materials may thus stop slightly above the bed of channels. Naturally, the pitch and diameter of these channels must be made optimal, as must the water pressure, and several beds of channels of different diameters may be provided beneath the tank. FIG. 3 shows, in section, a vertical wall 20 equipped with a drainage system according to the invention. Vertical channels 10 are shown, which appear to be easier to construct, but the channels could also be horizontal or peripheral. They are produced during casting of the concrete, in known manner, by arranging in the formwork rubber tubes inflated with water and held in a straight line by internal steel rods, to resist any deviations caused by forces occurring during the positioning of the concrete; these rods and tubes can readily be removed after the concrete has set, at heights of more than 10 m. A set of vertical tubes is connected, by means of an upper collector 11 with a connecting tube 12, to the corresponding reservoir (not shown) in the case of the embodiment according to FIG. 2 and to the corresponding filter (not shown) in the case of the embodiment according to FIG. 1. The tight seal required round the tube 12 where it passes through the inner part of the wall 20 may be obtained using known devices such as guard discs 13 combined with an injection of cement mortar between the discs. Vertical prestressing reinforcements such as 14 and horizontal ones such as 15 and steel connecting means such as 16 and 17 which prevent in particular any widthways flaking of the wall are provided in the wall. The network of drainage or topping-up channels may be placed either in the ordinary concrete used to build the wall or in a special permeable layer 18, notably consisting of concrete with a large proportion of cavities. Permeable inclusions arranged in horizontal strips such as 19 which link the vertical channels and consist of permeable tubes filled with gravel may also be provided in the concrete during casting. The arrangement of all the channels and possible additional permeable devices should be such that, in the event of the wall 20 cracking due to an overload, there is no risk of anything passing directly through the wall without coming into contact with the drainage network. In a wall 2 m thick, there could be a network consisting of channels 0.04 m in diameter at a spacing of 0.5 m. Therefore, a confinement enclosure according to the invention when compared with known enclosures has the advantages of being very easy to produce and virtually leaktight in the case of any accidents which can be foreseen, whilst giving a resistance to extreme loads which is as great as the mass of materials used will permit. The features described for the wall with the drainage or topping-up network incorporated in it also extend to cover various other embodiments, especially in underground construction. The invention is not limited to the embodiments described and represented hereinbefore and various modifications can be made thereto without passing beyond the scope of the invention. |
summary | ||
050769987 | claims | 1. A method for monitoring the power output of a nuclear reactor int he low power range, comprising: detecting neutrons produced by the reactor and producing, at the end of each of a succession of equal measuring intervals, a representation of the number of neutrons detected during a time period preceding the end of the respective measuring interval; and producing a power output level indication having a value associated with each measuring interval by a digital operation in which the indication value associated with a preceding measuring interval is altered according to a function of the representation produced during said detecting step for the present measuring interval, wherein said step of producing a power output level indication comprises: determining the difference between the value of the representation produced for the present measuring interval and the value of the power output level indication associated with a preceding measuring interval; providing a selected multiplying factor having a value less than unity; forming a representation of the value of the product of the difference determined in said determining step and the selected multiplying factor; and adding the product value representation to the power output level indication associated with the preceding measuring interval in order to produce the power output level indication associated with the present invention. detecting means for detecting neutrons produced by the reactor and producing, at the end of each of a succession of equal measuring intervals, a representation of the number of neutrons detected during a time period preceding the end of the respective measuring interval; and digital means connected for producing a power output level indication having a value associated with each measuring interval by a digital operation in which the indication value associated with a preceding time interval is altered according to a function of the representation produced by said detecting means for the present measuring interval, wherein said digital means comprises: means for determining the difference between the value of the representation produced for the present measuring interval and the value of the power output level indication associated with a preceding measuring interval; means connected for providing a selected multiplying factor having a value less than unity; means connected for forming a representation of the value of the product of the difference determined by said means for determining and the selected multiplying factor; and means connected for adding the product value representation to the power output level indication associated with the preceding measuring interval in order to produce the power output level indication associated with the present interval. 2. A method as defined in claim 1 wherein said step of providing a selected multiplying factor comprises giving the multiplying factor a value inversely proportional to at least the value of the power output level indication associated with a preceding measuring interval. 3. A method as defined in claim 2 wherein the value of the multiplying factor is inversely proportional to the sum of value of the power output level indication associated with the proceeding measuring interval and the value of the representation produced for the present measuring interval. 4. A method as defined in claim 3 further comprising generating a reactor power output level rate of change indication having a value associated with each measuring interval by: deriving, during each measuring interval, a power level change representation having a value which is a function of the difference between the values of the power output level indications associated with the present measuring interval and with a preceding measuring interval; and producing a reactor power output level rate of change indication having a value associated with each measuring interval by modifying the rate of change indication value associated with a preceding measuring interval according to a function of the power level change representation value associated with the present measuring interval. 5. A method as defined in claim 5 wherein the function of the difference in said deriving step is the difference between the logs of the values of the power output level indications, divided by the duration of one measuring interval. 6. A method as defined in claim 1 further comprising generating a reactor power output level rate of change indication having a value associated with each measuring interval by: deriving, during each measuring interval, a power level change representation having a value which is a function of the difference between the values of the power output level indications associated with the present measuring interval and with a preceding measuring interval; and producing a reactor power output level rate of change indication having a value associated with each measuring interval by modifying the rate of change indication value associated with a preceding measuring interval according to a function of the power level change representation value associated with the present measuring interval. 7. A method as defined in claim 6 wherein the function of the difference in said deriving step is the difference between the logs of the values of the power output level indications, divided by the duration of one measuring interval. 8. A device for monitoring the power output of a nuclear reactor in the low power range, comprising: 9. A device as defined in claim 8 wherein said means for providing a selected multiplying factor comprises means for giving the multiplying factor a value inversely proportional to at least the value of the power output level indication associated with a preceding measuring interval. 10. A device as defined in claim 9 wherein the value of the multiplying factor is inversely proportional to the sum of value of the power output level indication associated with the proceeding measuring interval and the value of the representation produced for the present measuring interval. 11. A device as defined in claim 10 further comprising means for generating a reactor power output level rate of change indication having value associated with each measuring interval, said means for generating comprising: first calculating means for deriving, during each measuring interval, a power level change representation having a value which is a function of the difference between the values of the power output level indications associated with the present measuring interval and with a preceding measuring interval; and second calculating means connected to said first calculating means for producing a reactor power output level rate of change indication having a value associated with each measuring interval by modifying the rate of change indication value associated with a preceding measuring interval according to a function of the power level change representation value associated wit the present measuring interval. 12. A device as defined in claim 11 wherein the function of the difference between the values is the difference between the logs of the values of the power output level indications, divided by the duration of one measuring interval. 13. A device as defined in claim 8 further comprising means for generating a reactor power output level rate of change indication having value associated with each measuring interval, said means for generating comprising: first calculating means for deriving, during each measuring interval, a power level change representation having a value which is a function of the difference between the values of the power output level indications associated with the present measuring interval and with a preceding measuring interval; and second calculating means connected to said first calculating means for producing a reactor power output level rate of change indication having a value associated with each measuring interval by modifying the rate of change indication value associated with a preceding measuring interval according to a function of the power level change representation value associated wit the present measuring 14. A device as defined in claim 13 wherein the function of the difference between the values is the difference between the logs of the values of the power output level indications, divided by the duration of one measuring interval. |
052951693 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Preferred embodiments of this invention will now be described with reference to the accompanying drawings. First, the first embodiment of this invention will be described with reference to FIGS. 1 to 3. This embodiment is shown as applied to a facility including a steel reactor containment vessel having a diameter of 34 m. The reason for the containment-vessel diameter of 34 m is that this dimension will allow the reactor vessel, the piping system, and the requisite equipment for operation to be housed in the containment vessel in the case of a plant whose output electrical power ranges from 600 to 1500 MW. In the other embodiments described below, containment vessels of the same diameter are adopted. In FIG. 1, the reactor containment facility of this embodiment comprises: a reactor pressure vessel 2 containing a reactor core 1; a dry well 3 providing a space in which the core 1 is arranged; a suppression chamber 4 consisting of suppression-pool water (hereinafter referred to simply as "suppression pool" or "pool water") 5 and a gaseous-phase-space wet well 6 defined above the suppression pool; a plurality of vent pipes 7 connecting the dry well 3 and the suppression pool 5 to each other; a reactor-containment-vessel wall 8 made of steel; and an outer peripheral pool 9 arranged outside the suppression pool 5. The facility further comprises: an accumulator water tank 25 and a gravity-driven water tank 26, which are situated above the reactor pressure vessel 2 and connected thereto through check valves 28; and a submerging system 27 connecting the suppression pool 5 and the reactor pressure vessel 2 to each other through a check valve. By "reactor containment vessel" is implied here the steel reactor-containment-vessel wall 8 and the structures integrally built therein, i.e., the suppression chamber 4, etc. This reactor containment vessel is formed as a vessel of natural-cooling type, with the outer peripheral pool 9 being arranged around it. The components featuring this embodiment will be described. The wet well 6 is divided by a partition 63 into first and second spaces 61 and 62, with the first space 61 being in contact with the surface of the suppression pool water. The two spaces communicate with each other through a plurality of pipes 64 extending through the partition 63. Further, provided in the partition 63 are a plurality of pipes 65, which allows the bottom section of the second space to communicate with the suppression pool water. Provided outside that portion of the steel reactor-containment-vessel wall 8 which is adjacent to the second space of the wet well, is an air passage 66, which sucks in air through an inlet in the lower section of the building and discharges it through an outlet in the upper section of the same. FIG. 2 is an enlarged view of a part of the portion around the suppression chamber. The allowable temperature for the suppression chamber 4 of the reactor containment vessel is determined as follows: The pressure in the wet well 6 (61), which is in contact with the vessel wall 8 constituting the pressure boundary of the reactor containment vessel, is the sum of the noncondensing-gas partial pressure and the vapor partial pressure, in the wet well. At the time of an accident, all the noncondensing gas that exists in the reactor containment vessel in normal operation is accumulated in the wet well, so that the maximum value of the noncondensing-gas partial pressure in the wet well in this condition is determined from the ratio of the total gaseous-phase volume of the reactor containment vessel to the wet-well volume. Further, the vapor partial pressure in the wet well is determined as the saturation vapor pressure corresponding to the surface temperature of the suppression pool 5. The temperature of the suppression pool must be limited such that, at the time of an accident, the pressure in the wet well, which is the sum of the above two categories of pressure, is not higher than the withstanding pressure of the vessel, i.e., such that the vapor partial pressure is not larger than the difference between the withstanding pressure of the vessel and the noncondensing-gas partial pressure. The temperature limit thus obtained constitutes the allowable temperature for the suppression pool. In this embodiment, the allowable temperature for the suppression chamber is raised by the following principle to increase the difference in temperature between the suppression chamber and the outer periphery thereof, and, due to the large temperature difference thus attained, it is possible to dissipate a large quantity of heat to the exterior. At the time of a loss-of-coolant accident, the noncodensing gas in the dry well 3 is forced out by the steam discharged from the reactor pressure vessel and flows through the vent pipes 7 to the suppression pool 5, accompanied by the steam. At this time, the noncondensing gas is first accumulated in the first space 61, which is in contact with the surface of the suppression-pool water, and, after raising the pressure of that region, flows into the second space 62 due to the pressure difference. Afterwards, as a result of the steam flowing through the vent pipes 7 into the suppression pool to condense therein, the water temperature of the suppression pool rises, and the vapor partial pressure in the first space 61 is raised, with the total pressure also rising. Since gaseous-phase circulation/mixing is not restricted in the first space 61, the noncondensing gas and the steam are evenly mixed with each other. Due to the pressure difference between the first and second spaces 61 and 62, this gaseous phase flows from the first space 61 into the second space 62. Since the second space 62 is cooled and the steam flowing into it accompanied by noncondensing gas due to the above action is partly or entirely condensed therein, the pressure in the second space 62 becomes lower than that in the first space 61. As a result, the steam accompanied by noncondensing gas again flows from the first space 61 to the second space 62. As this operation is repeated, the noncondensing gas in the wet well is entirely accumulated in the second space 62. Since the returning of the gaseous phase from the second space 62 to the first space 61 is restricted, the first space 61 is filled with steam only, so that when considering the pressure in this space, it is only necessary to take into account the vapor pressure. The cooling amount required at this time in the second space 62 is that required for making the temperature of the steam flowing into the second space through the passages 64 connecting the two spaces lower than the temperature when it is in the first space 61, so that it need not be a large one. Further, since the wet well is divided into upper and lower sections, the size of the reactor containment vessel is not influenced. As a result of the above operation, the saturation steam temperature corresponding to the withstanding pressure of the vessel is obtained as the allowable temperature for the suppression pool. Due to this arrangement, it is possible to raise the allowable water temperature for the suppression pool under the condition of the same withstanding pressure of the pressure vessel, without changing the thickness of the reactor-containment-vessel wall 8; furthermore, the difference in temperature between the suppression pool and the outer peripheral pool increases, thus attaining an improvement in heat dissipation characteristic. Accordingly, this reactor containment facility can be applied to a plant of a higher output power with the same containment-vessel configuration. Next, the operation of this embodiment will be explained, partly repeating what has been described above. At the time of a loss-of-coolant accident, which is taken into account from the viewpoint of safety when designing a nuclear reactor, the coolant in the reactor pressure vessel 2 flows out into the dry well 3 as steam at high temperature and pressure. Control rods (not shown) are inserted into the reactor core 1 to stop the nuclear fission; in the reactor core, however, the generation of decay heat continues for a long period after that. As the pressure in the reactor pressure vessel decreases, cooling water is supplied thereto from the accumulator water tank 25, the gravity-driven water tank 26 and the core submerging system 27, due to the difference in pressure and gravitation, thereby maintaining the submergence of the reactor core 1. The decay heat in the reactor core 1 is removed by the evaporation of this cooling water and steam is discharged through the rupture section to the dry well 3, whereby the pressure in the dry well 3 is raised to force the water level in the vent pipes 7 downwards, with the steam flowing into the suppression pool 5 to be condensed in the pool water. In this process, the noncondensing gas which has been in the dry well is forced out by the discharged steam and flows, accompanied by it, into the suppression pool, where it ascends to be accumulated in the first space 61. As a result of this accumulation, the pressure in the first space 61 is raised, so that the noncondensing gas flows through the pipes 64 into the second space 62. The transfer of the noncondensing gas from the dry well to the wet well is completed in several minutes after the occurrence of an assumed accident; afterwards, only the steam discharged from the reactor pressure vessel 2 flows into the suppression pool. Because of the decay heat generated during the steam condensation in the suppression pool 5, that portion of the pool water around the vent-pipe outlets 13 is heated, and, due to convection, the temperature of that portion of the pool water which is above the vent-pipe outlets 13 is raised in a substantially uniform fashion. With this rise in temperature, evaporation takes place at the surface of the pool water, and the vapor partial pressure in the first space 61 also rises to cause the pressure in the space to be raised. And the generated steam flows into the second space 62 along with the noncondensing gas remaining in the first space 61. The steam which has entered the second space 62 releases heat to the outer passage 66 through the steel containment-vessel wall 8 and condenses on the wall surface; afterwards, it returns to the suppression pool 5 through the pipes 65. Though its cooling capacity per unit area is small, the natural air-cooling action for cooling the second space 62 is effective since the area of the containment-vessel wall 8 serving as the heat transfer surface is large and the amount of steam flowing in through the pipes 64 is small; thus it is capable of condensing the steam entering the second space 62 and keeping the space at a temperature lower than that of the first space 61. Here, in the pipes 64, the noncondensing gas which has once entered the second space 62 is prevented from flowing back to the first space 61 by the gas flow from the first space 61, and, in the pipes 65, it is prevented by the suppression-pool water in which the pipes are immersed. By repeating the operations described above, substantially the total noncondensing gas is accumulated in the second space 62, with the first space 61 being filled with steam at a temperature equal to the surface temperature of the suppression-pool water. On the other hand, the heat dissipation from the containment vessel at the time of an assumed accident is basically effected by the heat release from the suppression pool 5, which has attained high temperature, to the outer peripheral pool 9 through the steel containment-vessel wall. The improvement in heat dissipation characteristic attained in this embodiment will be explained with reference to FIG. 3, which is a diagram comparing a case where this embodiment is applied with a case where it is not, in terms of the changes in the pressure in the containment vessel with respect to the time elapsing after the occurrence of an assumed accident. In the case where this embodiment is not applied, which is represented by the broken line, the temperature of the suppression-pool water rises in process of time; as the temperature of the suppression-pool water rises, however, the quantity of heat released to the outer peripheral pool is also augmented, until it exceeds the quantity of decay heat generated in the reactor core, with the result that the pressure in the containment vessel starts to decrease, thus keeping the pressure in the containment vessel below the withstanding pressure of the vessel. This established plant output power will be defined as a standardized output power 1.0, which is represented by the solid line A in the drawing. In the case of this standardized output power of 1.0, where this embodiment is applied, it is not the sum of the noncondensing-gas partial pressure and the vapor partial pressure but the vapor partial pressure only that is to be considered to be the pressure in the containment vessel, as stated in the description of the principle. Accordingly, the maximum pressure at the time of an assumed loss-of-coolant accident is relatively low. That is, in this embodiment, the temperature of the suppression-pool water is allowed to rise in correspondence with this lowered pressure, and the difference in temperature between the suppression pool and the outer peripheral pool, which is open to the atmospheric air and, consequently, whose temperature cannot be higher than 100.degree. C., can be augmented, whereby an improvement is attained in terms of heat dissipation characteristic, making it possible for a containment vessel of the same size to be applied to a plant of a larger output power. In the case of this embodiment, the allowable temperature for the suppression pool can be raised from approx. 122.degree. C. to approx. 144.degree. C. By applying this embodiment, the pressure suppression in the containment vessel is established at a standardized plant output power of 1.6, as indicated by the solid line B of the drawing, thus making it possible to make the applicable plant output power 1.6 times larger. When the pressure in the containment vessel begins to exhibit an inclination to decrease, the pressure in the second space 62 becomes higher than that in the first space 61, and part of the noncondensing gas accumulated in the second space 62 returns to the first space 61 through the pipes 64; at this time, however, the quantity of heat dissipated is in excess of that of decay heat, so that no serious problem is involved in terms of heat dissipation characteristic. If it is desired to prevent this returning of noncondensing gas, it is only necessary to provide check valves in the pipes 64. Further, while in this embodiment return pipes 65 for the condensed water in the second space 62 are provided, such pipes are not absolutely necessary; if they are not provided, water will accumulate in the second space, which, however, will entail no problem in terms of operation. A modification of this embodiment will be described with reference to FIG. 4. This modification differs from the above-described embodiment in that the partition 63 exhibits a stepped section at its end on the side of the containment-vessel wall 8, such that the second space 62 is extended downwards, with the extended region being in thermal contact with the outer-peripheral-pool water situated above the water level in the suppression pool 5; and, further, the pipes 65 for returning condensed water are provided in the bottom section of that region of the second space 62 extended downwards. Cooling which is effected by pool water, as in this embodiment, provides a better heat dissipation as compared with the natural air cooling described above, so that the requisite heat transfer area for cooling the second space 62 can be reduced. Accordingly, an air-cooling means which requires a large heat transfer area and which has to be installed and maintained at a relatively high position above the containment vessel, can be dispensed with. The heat dissipation characteristic which is obtained by this modification as a whole is the same as that described with reference to FIG. 3. A still another modification of the above embodiment will be described with reference to FIG. 5. This modification differs from the above embodiment in that it includes a wet-well-cooling-water pool 41, which is in contact with the outer periphery of the reactor-containment-vessel wall 8 so as to cool the wet well 6 and which is separate from the outer peripheral pool 9, as a means for cooling the second space 62 of the wet well 6, with a circumferential ring-like structure 42 being provided on that portion of the wet well 6 which constitutes the inner periphery of the reactor-containment-vessel wall 8. With this modification, the second space is cooled by pool water, which provides a better heat dissipation as compared with natural air cooling, so that the requisite heat transfer area is reduced, thereby eliminating the need to install a cooling means at a position outside and relatively higher than the reactor containment vessel. This helps to attain an improvement in terms of the ease with which the reactor containment facility is constructed and maintained. Further, by providing, as in this modification, a pool which is separate from the outer peripheral pool 9 situated below and which is intended for that portion of the cooling water which is excessively higher than the water level in the suppression pool 5, the suppression pool 5 can be protected from an excessive external pressure (water head) due to water-level difference during normal operation. Further, as to the external pressure applied to the reactor-containment-vessel wall due to the water level of the wet-well-cooling-water pool 41 that is separately provided, it can be coped with by means of a ring-like structure 42 provided on that portion of the wet well 6 which constitutes the inner periphery of the reactor-containment vessel 8, without changing the thickness of the reactor containment vessel. It is only necessary for this ring-like structure to be installed in the region of the wet well 6, which is a gaseous-phase space; if arranged in water, such a structure would hinder the water convection. As it is, the structure does not hinder the condensation heat transfer at the wall surface, thus avoiding deterioration in heat dissipation characteristic. On the contrary, this ring-like structure helps to prevent the development of a condensate film on the wall surface, so that an improvement can be expected in terms of the heat transfer at the wall surface due to augmentation in liquid film thickness. A second embodiment of this invention will be described with reference to FIG. 6. This embodiment differs from the one shown in FIG. 1 in that, instead of providing a partition structure dividing the wet well 6, there is provided on the water surface of the suppression pool 5 a layer 51 of a hydrophobic material, such as silicone oil or spindle oil, which exhibits a low saturation vapor pressure even at a temperature higher than 100.degree. C. and whose density is smaller than that of water. In this embodiment, the temperature in the suppression chamber is raised by the following principle to increase the difference in temperature between it and the outer periphery thereof, thereby making it possible to dissipate a greater quantity of heat to the outer periphery. By forming on the water surface of the suppression pool 5 a layer 51 of a hydrophobic material, such as silicone oil or spindle oil, which exhibits a low saturation vapor pressure and whose density is smaller than that of water, the pool water 5 is isolated from the wet well 6 by the hydrophobic-material layer 51. The temperature of the hydrophobic material is equal to the temperature of the water surface of the suppression pool; since, however, its saturation vapor pressure is low, the increase in the pressure in the wet well 6 is small. Since the temperature of the suppression pool 5 is raised and the pressure in the wet well 6 is kept at a low level, the water of the suppression pool will presently start boiling. The steam generated as a result of the boiling of the suppression pool passes through the hydrophobic-material layer 51 on the water surface to enter the wet well, thereby raising the pressure therein. Since, however, it is isolated from the pool surface, the wet well has a low humidity and is, consequently, in a superheated-steam condition (in which the temperature is higher than the saturation temperature corresponding to the vapor partial pressure). And when the total pressure in the wet well 6 (the sum of the noncondensing-gas partial pressure and the vapor partial pressure) attains the saturation pressure corresponding to the water temperature of the suppression pool, the suppression pool ceases to boil. This phenomenon repeats itself; that is, due to the formation of the hydrophobic-material layer 51 on the pool surface, the evaporation of the pool water, which, in the prior art, would start at the saturation temperature corresponding to the vapor partial pressure in the wet well, can be made to start at the saturation temperature corresponding to the total pressure in the wet well. In other words, the water temperature of the suppression pool can be kept at a higher level under the same wet-well pressure. In the case where the wet well is cooled, the hydrophobic material may be regarded as a substitute for the partition section dividing the wet well in the first embodiment. For the above reason, this embodiment makes it possible to raise the allowable suppression-pool-water temperature under the same allowable steam-partial pressure in the wet well (the same withstanding pressure of the vessel), thus helping to increase the difference in temperature between the suppression pool and the outer peripheral pool and attaining an improvement in terms of heat dissipation. Thus, the embodiment can be applied to a plant of a larger output power with the same reactor-containment-vessel configuration. Next, the operation of this embodiment will be explained, partly repeating what has been described above. The noncondensing gas transferred from the dry well, along with the steam, during the initial period after the occurrence of an accident, is accumulated in the wet well 6 after having passed through the suppression pool 5 and the hydrophobic-material layer 51. On the other hand, the steam coming through the vent pipes 7 is condensed in the pool water, thereby heating the pool water and the hydrophobic-material layer 51. Since, however, the pool surface which is in contact with the wet well 6 consists of a hydrophobic material having a low saturation vapor pressure, the rise in the vapor partial pressure in the wet well 6 is small (practically zero) even if the pool-water temperature is raised, so that it is not necessary to take into account the vapor partial pressure of the hydrophobic material. And, by forming, as in this embodiment, a layer 51 of a hydrophobic material which exhibits a low saturation vapor pressure and whose density is smaller than water, the evaporation of the pool water at the surface of the suppression pool 5 can be restrained, as stated above. Accordingly, it is possible to realize a condition in which the water temperature of the suppression pool 5 is high, with the pressure in the wet well 6 being low. That is, the allowable temperature for the suppression pool can be raised without changing the size of the containment vessel. Further, in the case where the wet well is cooled by natural air cooling as in this embodiment, the hydrophobic material may, as stated above, be regarded as a substitute for the partition 63 and the pipes 64 and 65 of the embodiment shown in FIG. 1, so that the improvement in heat dissipation characteristic attained in this embodiment is the same as that shown in FIG. 3. That is, in this embodiment, the allowable temperature for the suppression pool can be raised from approx. 122.degree. C. to approx. 144.degree. C., and the applicable plant output power can be made 1.6 times larger. A third embodiment of this invention will be described with reference to FIG. 7. This embodiment is featured by upper and lower openings 10 and 11, which are provided in that portion of the steel containment-vessel wall 8 which is in the suppression-pool water, such as to be positioned with the vent-pipe outlets 13 therebetween, the upper and lower openings 10 and 11 being connected with each other through a plurality of convection promoting pipes 12 provided in the outer peripheral pool 9. Apart from this, the main components of this embodiment are the same as those of the embodiment shown in FIG. 1. In this embodiment, the vent-pipe outlets 13 and the upper and lower openings 10 and 11 are positioned in such a manner that the difference in height between the upper openings 10 and the vent-pipe outlets 13 is larger than the difference in height between the vent-pipe outlets 13 and the lower openings 11. In this embodiment, it is possible to release a large quantity of heat from the reactor containment vessel by the following principle: The steam discharged from the pressure vessel 2 into the dry well 3 at the time of a loss-of-coolant accident is introduced into the suppression pool 5 through the vent pipes 7 and condensed in the pool water. As a result, the temperature of that portion of the suppression-pool water which is above the vent-pipe outlets 13 is raised, and, by virtue of the convection in the pool formed by the heating due to the steam condensation at the vent-pipe outlets, that region of the pool water attains a uniform high-temperature condition. On the other hand, the temperature of that portion of the suppression-pool water which is below the vent-pipe outlets 13 is not raised at this time, with the result that a temperature stratification occurs at the height of the vent-pipe outlets. Here, the density condition in that portion of the suppression pool which is between the upper and lower ends of the convection promoting pipes 12 and the density condition in the convection promoting pipes 12, will be considered. That portion of the suppression pool which is above the vent-pipe outlets is at a relatively high temperature and has a small density, whereas that portion thereof which is on the side of the convection promoting pipes is at a relatively low temperature and has a large density. As a result, the water head (production of density and height: .rho.g h) at the lower end of the section being considered is larger on the side of the convection promoting pipes, so that a circulation is formed which flows downwards through the convection promoting pipes 12 to enter the suppression pool. This causes the high-temperature water of the upper portion of the suppression pool to enter the upper section of the convection promoting pipes; since, however, the convection promoting pipes are immersed in the outer peripheral pool 9, which is at low temperature, heat dissipation takes place through the pipe walls, and the temperature of the water is gradually lowered as it flows downwards through the convection promoting pipes. Due to this action, a condition is constantly maintained in the section above the vent-pipe outlets 13 in which the density and, consequently, the water head, are larger on the convection-promoting-pipe side, thereby forming a drive power for the circulation flowing downwards through the convection promoting pipes. In the section below the vent-pipe outlets 13, on the other hand, the portion on the side of the convection promoting pipes first attains high temperature as a result of the circulation, so that the density (water head) condition is such as to cancel the above-mentioned drive power for circulation; however, by appropriately setting the position of the lower ends of the convection promoting pipes (i.e., by setting the position of the lower ends of the convection promoting pipes 12 at such a position as will not completely cancel the downward drive power formed in the upper section), it is possible to cause water at a relatively high temperature to flow into the region below the vent-pipe outlets while maintaining the circulation flowing downwards through the convection promoting pipes. By virtue of this action, the hot water which has flowed into the region below the vent-pipe outlets causes the water temperature of this region to be raised, which helps to eliminate the density (water head) condition which would cancel the drive force for circulation formed in the region below the vent-pipe outlets, thus promoting the circulation. By repeating the above action, the water temperature of the region below the vent-pipe outlets can be continuously raised. At this time, the water temperature of the region below the vent-pipe outlets is substantially the same as the temperature of the water flowing in at the lower ends of the convection promoting pipes, so that it does not become lower than the water temperature of the outer peripheral pool. As a result, pool water at a relatively high temperature is constantly circulated in that region of the suppression pool 5 which is below the vent-pipe outlets 13 and which is at low temperature, thereby increasing the region for absorbing heat from the reactor core 1, and, at the same time, making it possible to utilize not only the reactor-containment-vessel wall 8 corresponding to the suppression-pool region which is at high temperature, but also the walls of the convection promoting pipes 12, as the heat transfer surface through which heat is dissipated to the outer peripheral pool 9. Since it is normal for the convection promoting pipes 12 to have a diameter smaller than that of the reactor containment vessel, the convection promoting pipes can be arbitrarily provided without influencing the withstanding pressure of the reactor containment vessel, so that the heat transfer area can be augmented without changing the size or the withstanding pressure of the reactor containment vessel, thus increasing the quantity of heat that can be dissipated. Accordingly, this embodiment can be applied to a plant of a larger output power with a reactor containment vessel of the same principal dimensions, etc. Next, the operation of this embodiment will be explained, partly repeating what has been described above. The steam discharged from the reactor pressure vessel 2 at the time of an assumed loss-of-coolant accident flows through the vent-pipe outlets 13 into the suppression pool 5 to be condensed in the pool water. The pool water portion which is around the steam openings 13 is heated by the latent heat generated during the steam condensation in the suppression pool 5, and the temperature of the pool water portion which is above the vent-pipe outlets 13 is raised in a substantially uniform fashion. As a result, the density in the section where the convection promoting pipes 12 are arranged is low in the suppression-pool water, which has attained high temperature, and high in the convection promoting pipes 12, which are at low temperature. Due to this difference in density, a flow descending in the convection promoting pipes 12 is generated, and suppression-pool water which is at high temperature passes through the upper openings 10 and enters the upper sections of the convection promoting pipes 12. The high-temperature water which has thus flowed in is cooled inside the convection promoting pipes 12, which are immersed in the outer-peripheral-pool water, and becomes gradually cooler as it descends therein. As a result, a flow descending in the convection promoting pipes 12 is constantly formed, without changing the density condition in the section where the convection promoting pipes 12 are arranged. And, as stated above, water at a temperature higher than that of the water temperature of the outer peripheral pool passes through the lower openings 11 and enters that region of the suppression pool 5 which is below the vent-pipe outlets 13, thereby warming that region. FIGS. 8A and 8B show the temperature distribution and density distribution in the height direction of the suppression pool 5 and the convection promoting pipes 12 at a time after the occurrence of an assumed loss-of-coolant accident. Regarding the temperature, that section of the suppression pool 5 which is above the vent-pipe outlets 13 is at a uniformly high temperature due to the steam from the vent pipes; since the water is cooled in the convection promoting pipes 12, its temperature exhibits a linear reduction toward the lower end; and, as since water at the temperature of this lower end flows in, that region of the suppression pool 5 which is below the vent-pipe outlets 13 attains a temperature substantially equal to that. Here, the reason for the linear reduction in temperature in the convection promoting pipes 12 is that the diameter of the convection promoting pipes is uniform and that, consequently, the cooling in the outer peripheral pool is also uniform in the height direction. Further, the reason for making the water temperature of that region of the suppression pool 5 which is below the vent-pipe outlets 13 substantially equal to the temperature of the water which has flowed in is that the phenomenon in question is a gentle one which covers a long period and, consequently, can be treated as one of a quasi-constant nature. As is known, the density of water is in inverse proportion to the temperature thereof, so that, as shown in FIG. 8B, the density distribution is reverse to the temperature distribution shown in FIG. 8A. As can be seen from the drawing, in the section above the vent-pipe outlets 13, the density on the side of the convection promoting pipes 12 is larger, so that the drive force (water head) in each of the convection promoting pipes 12 works such as to cause a downward flow therein. The sum total of these downward drive forces corresponds to the area of the triangle a formed by the density lines of the two regions in the drawing. In the section below the vent-pipe outlets 13, in contrast, the density on the suppression-pool side is larger, so that the drive forces in this section work such as to cancel the downward drive force generated in the upper section. If the sum total of the drive forces generated in the section below the vent-pipe outlets (i.e., the area of the triangle b formed by the density lines in the drawing) is smaller than the sum total of the drive forces generated in the upper section, the flow in the convection promoting pipes 12 is generally a downward one. That is, if, in the drawing, the area of the triangle a is larger than that of the triangle b, a general flow descending in the convection promoting pipes 12 is constantly formed, thus enabling the present means to operate effectively. It may be concluded from this that, in the case where, as in this embodiment, the configuration of the convection promoting pipes 12 and the cooling condition are both uniform, the condition for the present means to work effectively is that the difference in height between the upper openings 10 and the vent-pipe outlets 13 (L1 in the drawing) be larger than the difference in height between the vent-pipe outlets 13 and the lower openings 11 (L2 in the drawing). Thus, the area of the heat transfer surface through which heat is transferred from the suppression pool 5 to the outer peripheral pool 9 is enlarged, thereby attaining an improvement in heat dissipation characteristic. The improvement in heat dissipation in this embodiment will be explained with reference to FIG. 9. In the drawing, changes in the containment-vessel pressure in process of time after the occurrence of an accident assumed in a case where this embodiment is applied, are, as in the case of FIG. 3, compared with those in a case where it is not. In the case where this embodiment is applied, which is represented by the solid lines in the drawing, it is assumed that approximately 500 convection promoting pipes 12 having a diameter of 50 mm are provided. Since the convection promoting pipes 12 can be arranged over the entire periphery of the containment vessel, no particular problem is involved in terms of practical application. In the case where a standardized plant output power of 1.0 is applied to this embodiment, the heat dissipation area is enlarged, as indicated by the solid line A in the drawing, so that the maximum pressure at the time of an assumed accident is lower than in the case where this embodiment is not applied (indicated by the broken line in the drawing). In correspondence with this reduction in pressure, this embodiment is more suitable for application to a plant of a larger output power. As indicated by the solid line B in the drawing, the applicable standard plant output power is 1.5; which means, this embodiment makes it possible to make the applicable plant output power 1.5 times larger. A modification of the third embodiment will be described with reference to FIG. 10. This modification differs from the embodiment shown in FIG. 7 in that isolation valves 21 are provided on the outer and inner sides of the upper and outer openings 10 and 11 in the reactor-containment-vessel wall 8. FIG. 10 shows a part of this modification where the suppression pool and the outer peripheral pool are arranged. During normal operation, the isolation valves 21 are open; when performing periodical inspection, these isolation valves 21 are closed to isolate the reactor containment vessel from the convection promoting pipes 12, thereby facilitating maintenance operations, such as the replacement of the convection promoting pipes 12. Further, if, for some reason, a leak from the convection promoting pipes 12 should occur, the reactor containment vessel can be isolated more reliably by closing the isolation valves 21. A still another modification of this embodiment will be described with reference to FIGS. 11 and 12. This modification differs from the above embodiments in that the convection promoting pipes are composed of upper header pipes 31 connected with the upper openings 10, lower header pipes 32 connected with the lower openings 11, and heat transfer pipes 33 connecting the upper and lower header pipes 31 and 32 with each other. FIG. 11 is a longitudinal sectional view showing a part of the suppression pool and the outer peripheral pool, and FIG. 12 is a cross-sectional view taken along the line A--A of FIG. 12. The water temperature in the upper header pipes 31 and that in the lower header pipes 32 are respectively made uniform by the water flowing in, so that the operation of this modification is the same as that of the above embodiment. This modification is advantageous in that the number of upper and lower openings 10 and 11 can be reduced, thereby attaining an improvement in terms of machinability and producibility. At the same time, by adjusting the diameter of the header pipes, it is possible to provide as many heat transfer pipes 33 as required, irrespective of the diameter of the reactor containment vessel. Further, by imparting a curved configuration to the heat transfer pipes 33, any expansion or contraction of the heat transfer pipes 33 due to temperature changes can be absorbed. A further modification of this embodiment will be described with reference to FIG. 13. In this modification, the convection promoting pipes are divided into two sections: In the upper section, they are composed of upper header pipes 31 connected with the upper openings 10, lower header pipes 32, and a plurality of heat transfer pipes 33 connecting the upper and lower header pipes with each other; in the lower section, they are composed of the lower header pipes 32 and pipes 34 of a relatively large diameter which are less in number than the heat transfer pipes 33 and which connect the lower header pipes 32 with the lower openings 11. FIGS. 14A and 14B show the temperature distribution and density distribution in the height direction in that section of the suppression pool and the convection promoting pipes which is between the upper and lower openings 10 and 11. In the case of this modification, the area of the contact surface with the outer peripheral pool is different between the upper and lower sections on the side of the convection promoting pipes, so that the cooling is effected to a larger degree in the upper section, which has a larger surface area. As a result, the temperature distribution in the convection promoting pipes is such that the temperature decreasing rate is high in the upper section and low in the lower section, with the density distribution being in correspondence therewith, as shown in the drawing. In this case, the generation of a drive force which would cancel the flow descending in the convection promoting pipes only occurs to a small degree, so that the lower openings 11 can be provided at a lower position as compared with the case of the embodiment shown in FIG. 7. Due to this arrangement, the region which can be effectively utilized can be enlarged in that section of the suppression pool 5 which is below the vent-pipe outlets 13. An embodiment which consists of a combination of the first and third embodiments will be described with reference to FIG. 15. Basically, this embodiment is a combination of the embodiments described with reference to FIGS. 5 and 7. In this embodiment, the allowable temperature for the suppression pool is raised by dividing the wet well into first and second spaces 61 and 62; and the heat dissipation area is enlarged by providing convection promoting pipes 12 in the region corresponding to the suppression pool. Further, for the purpose of cooling the second space 62 of the wet well, there are provided a wet well cooling pool 41 which is separate from the outer peripheral pool 9 and a ring-like structure 42, thereby attaining an improvement in terms of the withstanding pressure of the vessel during normal operation and the ease with which it is built and maintained. As for the improvement in heat dissipation characteristic, the rise in the allowable temperature for the suppression pool and the enlargement of the heat transfer area are combined with each other to make it possible to make the applicable plant output power 2.3 times larger (i.e., to realize an established standard plant output power of 2.3), as shown in FIG. 16. It is known from a presentation in the "Fall Meeting of the Atomic Energy Society of Japan in the Year 1989", mentioned in connection with the prior art, that an enlargement of the high-temperature region of the suppression pool can be realized by providing a convection promoting plate in the suppression pool. An embodiment which consists of a combination of this technique and the first embodiment will be described with reference to FIG. 17. In this embodiment, a convection promoting plate 70 is arranged in the suppression pool 5 of the embodiment shown in FIG. 1, along the reactor-containment-vessel wall 8. As stated in the above-mentioned presentation, this convection promoting plate 70 is arranged such that the vent-pipe outlets 13 are situated between the upper and lower ends of this plate, with the difference in height between the upper end and the vent-pipe outlets 13 being larger than the difference in height between the vent-pipe outlets 13 and the lower end, whereby the high-temperature region of the suppression pool is enlarged. According to this embodiment, the high-temperature region of the suppression pool is enlarged by the action of the convection promoting plate 70, and, though restricted to the section of the reactor-containment-vessel wall 8, the heat dissipation area can be enlarged. At the same time, the noncondensing gas in the wet well 6 can be concentrically collected in the second space 62, so that the allowable temperature for the suppression pool can be raised. Thus, with the enlargement of the heat dissipation area and the increase in temperature difference between the pools, it is possible to attain an enhancement in heat dissipation characteristic. The above-described basic embodiments can also be applied to a plant whose reactor-containment-vessel wall is mainly formed of concrete by forming that region thereof which corresponds to the suppression pool and the wet well as a steel wall 8, which is a good conductor of heat. In this embodiment, the wall outside the outer peripheral pool 9 constitutes the principal structure wall of the building, so that, if the reactor-containment-vessel wall, which has conventionally been formed of concrete, is formed as a steel wall 8, no serious problem is involved in terms of the strength of the building. In this embodiment, the wet well in the suppression chamber, which is partly formed by the steel reactor-containment-vessel wall 8, is divided into first and second spaces 61 and 62, and convection promoting pipes 12 are provided on that portion of the steel reactor-containment-vessel wall 8 which is in the suppression pool. Further, a wet-well-cooling-water pool 41, which is in contact with the outer periphery of the reactor-containment-vessel wall 8, is provided as a cooling means for the second space 62, with a circumferential ring-like structure 42 being provided on the inner periphery of the reactor-containment-vessel wall 8 of the wet well 6. By virtue of this arrangement, the allowable temperature for the suppression pool is raised, and the area of the heat dissipation surface through which heat is dissipated to the outer peripheral pool 9 is enlarged. While typical combinations of the different means have been presented, it is also possible realize other various combinations such as combinations of the embodiment shown in FIG. 6 similar to the examples shown in FIGS. 15, 17 and 18; such combinations do not involve any problems due to mutual interference in terms of practical application or effect. A fourth embodiment of the present invention will be described with reference to FIG. 19. Referring to FIG. 19, a reactor containment vessel of natural cooling type contains a nuclear reactor pressure vessel 102 which accommodates a nuclear reactor core 101 and which is disposed in a dry well 103. The dry well 103 communicates with a CRD chamber 104 which is below the reactor pressure vessel 102 via the internal space of a gamma shield 105. A suppression chamber 107 having a suppression pool 106 is and wet well 107A above the suppression pool 106 disposed outside the dry well 103. The dry well 103 and the suppression pool 106 communicate with each other through a plurality of vent pipes 108. The level of the openings 108A of the vent pipes 108 opening to the dry well 103 is determined in conformity with the core submerging level which is set above the reactor core 101. The discharge opening 108B of the suppression pool 106 is set to a level which is at a suitable depth in water which is determined on the basis of the results of a steam condensation test. The term "core submerging level" is used to mean the level which is reached by the water rushing from the nuclear reactor pressure vessel 102 into the dry well 103 in the event of a loss-of-coolant accident and which is high enough to ensure submerging cooling of the reactor core 101. Therefore, when the water level in the dry well has reached the core submerging level, the water starts to flood from the dry well into the suppression chamber 107 through the vent pipes 108, with the result that the water level rises in the suppression pool 106. The submerging level should be determined in consideration of factors such as the construction of the reactor pressure vessel, power of the nuclear reactor, and so forth. Practically, however, the submerging level is higher than the upper end of the reactor core 101 in the reactor pressure vessel by s suitable safety margin which is determined by possible fluctuation of the water level. For instance, the submerging level is set to be at least 50 cm higher than the upper end of the reactor core 101. The containment vessel wall 109 made of steel serves as the outer wall which defines the radially outer end of the suppression pool 106 and the suppression chamber 107. The wall 109 is surrounded by a outer peripheral pool 110 The outer peripheral pool contains water to a level which high enough to measure itself with the depth of water in the suppression pool 106 after an accident, in order to ensure efficient transfer of heat from the suppression pool 106. A relief pipe 11 with a valve 112 leads from a space above the water surface in the outer peripheral pool 110, in order to relieve steam which is generated due to a temperature rise of the water in the pool 110 in the event of an accident. The steam which is generated in the reactor pressure vessel 102 is transmitted through a main steam pipe 113 into a turbine (not shown) and is liquefied by condensation so as to be finally returned to the reactor pressure vessel 102 via a feedwater pipe 114. In case of an emergency, a main steam isolation valve (MSIV) 115 is closed and a check valve 116 in the feedwater pipe 114 prevents coolant in liquid or liquid/vapor mixed phase from flowing backward out of the reactor pressure vessel 102. However, in the event of a rupture taking place in a portion of the main steam line upstream of the MSIV 115 or in a portion of the feedwater line downstream of the check valve 116, the coolant would undesirably be allowed to flow from the reactor pressure vessel 102 into the dry well 106 so as to expose the reactor core 101, resulting in a serious accident, i.e., a loss-of-coolant accident. In order to obviate this problem, an injection line 119 having a valve 118 is provided for the purpose of injecting water from a pressure accumulator water tank 117 into the reactor pressure vessel 102. Several types of water injection system using such a pressure accumulator water tank 117 are usable. For instance, the water in the pressure accumulator water tank 117 may always be pressurized. In another type, the pressure accumulator water tank 117 is brought into communication with the vapor space in the reactor pressure vessel 102 through a specific line only when the injection is required, so that water can flow into the reactor pressure vessel as a result of the difference in the water head between the pressure accumulator water tank 117 and the reactor pressure vessel 102. In still another type of the injection system, the internal pressure of the reactor pressure vessel 102 is relieved and reduced through a relief safety valve (not shown) and then introduced into the pressure accumulator water tank 117 so as to drive the water therefrom into the reactor pressure vessel 102. All these injection systems do not require any specific power source such as a pump and are operable just through a simple valve actuation. In the drawings, therefore, only the pressure accumulator water tank 117, the valve 118 and the injection line 119 are shown. Needless to say, an injection system employing a pump may be used equally, provided that the required injection rate is ensured. The quantity W of water stored in the pressure accumulator water tank 117 is determined to be substantially equal to the sum of the water quantity W1 in the dry well 103 necessary for building up the water column up to the aforementioned submerging level and the water quantity W2 necessary for accumulating the water in the suppression pool 106 up to the submerging level, i.e., the level of the openings 108A of the vent pipes 108 opening in the dry well 103. That is, the condition of W=W1+W2 is substantially met. Under this condition, the water level in the suppression pool 106 can rise up to the same level as the submerging level in the dry well, thus maximizing the area of heat transfer between the suppression pool 106 and the outer peripheral pool 110. The suppression chamber 107 and the dry well 103 communicate with each other through a communication pipe 121 having a check valve 120 which permits fluid to flow only in a predetermined direction, i.e., from the suppression chamber 107 into the dry well 103. A communication line 125 with valves 123, 124 is provided in order that the submerging water level is maintained both in the reactor pressure vessel 102 and the dry well 103 for a long time after the occurrence of the accident. In this embodiment, a structure 125A of, for example, concrete is placed to fill a vacant space around the outer wall of the dry well 103 up to a level below the level of the openings 108A of the vent pipes 108 in the dry well 103, in such a manner as not to cause impediment to works such as installation and periodical survey. This arrangement enables the water level in the dry well 103 to rise up to the submerging level in a shorter time and also to reduce the internal volume of the pressure accumulator water tank 117, because the dead space which may otherwise be filled with water is filled by the concrete structure. In the event of a rupture of the feedwater line 114, the coolant in liquid/vapor mixed phase is discharged from the interior of the reactor pressure vessel 102 into the dry well 103 through the fracture 122. The liquid phase portion of the hot coolant thus discharged flows into the CRD chamber 104 which is below the dry well 103 to raise the water level in this chamber 104. Meanwhile, the vapor phase portion of the coolant raises the pressure inside the dry well 103, which lowers the water level in the vent pipes 108 to allow the atmosphere in the dry well 103 to be transferred to the suppression chamber 107 through the water in the suppression pool 106. Since the vapor phase portion of the discharged coolant is condensed into liquid phase through the contact with the water in the suppression pool 106, only the air from the dry well 103 reaches the suppression chamber 107, with the result that the pressure rises in the suppression chamber 107. The nuclear reactor automatically stops in response to the rise in the internal pressure of the dry well 103 or the lowering of the water level in the reactor pressure vessel 102. At the same time, the MSIV 115 is closed to terminate the supply of steam. Supply of the condensate feedwater through the feedwater line also is stopped. Then, the valve 118 is opened to allow injection of water into the reactor pressure vessel 102 from the pressure accumulator water tank 117 so as to recover the water level in the reactor pressure vessel, thereby preventing damage to the reactor core which may otherwise be caused by overheat. The water injected into the reactor pressure vessel 102 is heated to a high temperature by the decay heat derived from the reactor core 1 and rushes into the dry well 103 through the fracture 122, so that the water level in the CRD chamber 104 which is below the dry well 103 rises. A further rise of the water level in the CRD chamber 104 causes the water to flood into the dry well 103 to raise the water level therein nearly to the level of the openings 108A of the vent pipes 108. The discharge of hot water from the rupture continues further, so that hot water is introduced into the water in the suppression pool 106 from the dry well 103 via the vent pipes 108, causing both the level and temperature of the water in the suppression pool 106 to be raised. When the water level has been raised to a level near the level of the openings 108A of the vent pipes 108 in the dry well 102, the pressure accumulator water tank 117 becomes almost empty and the discharge of water through the fracture 22 is ceased. As a result, the rise in the water level inside the suppression pool 106 also ceases. The rise of the water level in the suppression pool 106 tends to cause a rise of the pressure inside the suppression chamber 107. This, however, does not hinder the rise of the water level in the suppression pool 106, because the check valve 120 is opened by the pressure differential between the suppression chamber 107 and the dry well 103, so that air returns from the suppression chamber 107 into the dry well 103 to attain a pressure equilibrium therebetween. As a result of the rise of the water level in the suppression pool 106, heat is transferred from the suppression pool 106 to the outer peripheral pool 110 through the steel wall 109 of the containment vessel. As a consequence, the water in the outer peripheral pool 110 is boiled to generate vapor. The vapor is then relieved to the exterior through the relief pipe 111 by forcibly opening the valve 112. When the water injection from the pressure accumulator water tank 117 is ceased, water levels which are almost the same are attained in the reactor pressure vessel 102, dry well 103 and in the suppression pool 106. Generation of decay heat inside the reactor core 101 progressively decreases but is still effective in heating and evaporating water inside the reactor pressure vessel 102. Consequently, the pressure inside the reactor pressure vessel 102 continues to rise as a result of generation of heat. A further rise of the internal pressure of the reactor pressure vessel 102 due to generation of vapor causes the vapor to be relieved through a relief safety valve (not shown) or to be discharged together with hot water into the dry well 103 through the fracture. Consequently, the pressure rises in the dry well 103, which serves to lower the level of water in the vent pipes 108, so that the vapor component is discharged into the suppression pool 106 to be condensed into liquid phase through contact with the water in the pool 106. The heat given to the water as a result of the condensation is transferred to the water in the outer peripheral pool 110 so as to be dissipated therefrom. Meanwhile, the level of the water inside the reactor pressure vessel 102 tends to progressively come down as a result of the evaporation. However, the valve 123 is opened when the reduction of the water level is sensed so that the water flows from the suppression pool 106 into the reactor pressure vessel 102 due to the difference in the water head, whereby the water level is recovered in the reactor pressure vessel 102. By opening the valve 124 simultaneously with the opening of the valve 123, it is possible to maintain the same submerging level in the reactor pressure vessel 102, dry well 103 and the suppression pool 106. As will be understood from the foregoing description, the embodiment shown in FIG. 19 appreciably shortens the time required for the dry well 103 to be filled with water up to the submerging level in the event of an accident. In addition, submerging cooling of the reactor core can be started quickly because the water level in the suppression vessel 106 can be raised to the highest level without causing the openings 108A of the vent pipes 108 to be flooded by the water. Furthermore, the transfer of heat from the dry well 103 to the suppression pool 106 can be promoted by virtue of condensation of vapor component in contact with the water inside the suppression pool 106, partly because the upper hot portion of the water column in the dry well 103 first moves into the suppression pool 106 and partly because the water in the vent pipes 108 can easily be displaced as the time elapses to allow vapor component to flow from the dry well 103 into the suppression pool 106. In addition, transfer of heat from the suppression pool 106 to the outer peripheral pool 110 is promoted, because a large heat transfer area, as well as a large temperature differential, is provided between the water inside the suppression pool 106 and the water inside the outer peripheral pool 110. Thus, the fourth embodiment described in connection with FIG. 19 performs quick and efficient cooling of the reactor core, as well as efficient transfer of heat from the dry well 103 to the suppression pool 106 and further to the outer peripheral pool 110 therefrom, thus offering a remarkable improvement in the safety of the natural cooling type reactor core containment vessel. It is also to be noted that this embodiment makes it possible to set the initial water level in the dry well 103 to a level which is lower than that in existing equipment, which, in combination with the use of the structure 125A filling dead space inside the dry well 103, enables the capacity of the pressure accumulator water tank 117 to be reduced. A fifth embodiment of the present invention will be described with reference to FIGS. 20 and 21. Referring to these Figures, there is shown a natural cooling type reactor containment vessel having upper and lower suppression chambers. More specifically, a reactor pressure vessel 202 accommodating a reactor core 201 is placed in a dry well 203 which is surrounded by a lower suppression chamber 207 having a lower suppression pool 205 and a wet well 207A above the suppression pool 205. The dry well 203 and the lower suppression pool 205 are communicated with each other through a plurality of vent pipes 211. An upper suppression chamber 206 provided on the upper side of the lower suppression chamber 207 has an upper suppression pool 204 which communicates with the dry well 203 through a plurality of vent pipes 208 and a wet well 206A above the suppression pool 204. A water injection line 217 having a valve 216 leads from the upper suppression pool 204 to the reactor pressure vessel 202. The steel wall 214 of the containment vessel serves also as an outer wall which defines the radially outer ends of the lower suppression pool 205 and the lower suppression chamber 207. This outer wall is surrounded by a containment vessel outer peripheral pool 215 which is filled with water up to a level high enough to cover the core submerging level of water inside the dry well. The term "submerged level" is used in the same sense as that explained in the description of the preceding embodiment. Namely, the core submerging level is a level which is higher than the top of the reactor pressure vessel 201 by a height, e.g., 50 cm, which provides a margin for fluctuation of the water level. The quantity W of water in the upper suppression pool 205 is determined to be substantially equal to the sum of the water quantity W1 in the dry well 203 necessary for raising the water level therein to the aforementioned submerging level and the water quantity W2 which is required for raising the water level inside the lower suppression pool 205 to a level substantially equal to the aforementioned submerging level. That is, the condition of W=W1+W2 is substantially met. As a consequence, the water level in the lower suppression pool 205 rises almost to the same level as the openings 212 of the vent pipes 211 opening to the dry well 212, so that the area for the transfer of heat from the lower suppression pool 205 to the outer peripheral pool 215 can be maximized. Referring to FIG. 20, in the event of a rupture in a pipe which is directly connected to the reactor pressure vessel 202, a coolant in the form of a liquid/vapor mixed phase is discharged from the reactor pressure vessel 202 into the dry well 203. The liquid phase, i.e, water component, of the discharged hot coolant is accumulated on the bottom of the dry well 203, while the steam component of the same raises the pressure inside the dry well 203 so as to lower the water levels in the vent pipes 208, 211 Consequently, the atmosphere inside the dry well 203, composed of air and steam, is moved into the suppression chambers 206, 207 via the water in the respective suppression pools 204, 205. As the atmosphere passes through the water, the steam component of the same is condensed into liquid phase through contact with the water in the suppression pools 204, 205, so that only the air component of the atmosphere reaches the suppression chambers 206, 207 so as to contribute to the rise of the pressure in the suppression chambers 206, 207 and the dry well 203. The nuclear reactor is automatically stopped and isolated when the rise of the pressure inside the dry well 203 or the lowering of water level inside the reactor pressure vessel 202 is sensed in the event of a rupture. Only vapor phase is discharged from the fracture 218 after the water level inside the reactor pressure vessel has comedown below the level of the fracture 218. The pressure inside the reactor pressure vessel decreases accordingly. If the pressure does not decrease, a safety valve (not shown) operates to relief the pressure. When the pressure in the reactor pressure vessel has been lowered sufficiently, the valve 216 in the water injection line 217 leading from the upper suppression pool 204 is opened to allow injection of water into the reactor pressure vessel 202 thereby to cool the reactor core 201 by submerging. The water injected into the reactor pressure vessel 202 is heated to a high temperature by the decay heat derived from the reactor core 201 so that hot water is discharged from the fracture 218 into the dry well 202. As a consequence, the water level inside the dry well 202 is raised up to the level of the openings 212 of the vent pipes. As the discharge of hot water from the fracture 218 continues, hot water flows from the dry well 202 into the water in the lower suppression pool 205 via the vent pipes 211, causing rises of the water level and water temperature inside the lower suppression pool 205. The upper pressure suppression pool 204 becomes almost empty so that the discharge of hot water from the fracture 218 is materially ceased when the water level in the lower suppression chamber has been raised to the level of the openings 212 of the vent pipes. The rise of the water level in the lower suppression pool also is ceased accordingly. The rise of the water level inside the lower suppression pool 205 tends to cause a rise in the pressure inside the lower suppression chamber 207. This tendency, however, is canceled as the check valve 219 opens when the internal pressure of the lower suppression chamber 207 has become higher than that of the dry well 203, so as to attain a pressure equilibrium between the lower suppression chamber 207 and the dry well 203. The water level in the lower suppression chamber 205, therefore, can be raised without impediment. As a result of the rise of the water temperature in the lower suppression pool 205, heat is transferred from the water in the lower suppression pool 205 to the water in the outer peripheral pool 215 through the steel wall 214 of the containment vessel, so that the water in the outer peripheral pool 215 is heated and evaporated to generate steam which is relieved to the exterior, so that the steel wall 214 of the containment vessel is cooled to and maintained at a certain temperature. The water inside the reactor pressure vessel 202 progressively decreases as the time elapses. It is, however, possible to maintain a predetermined water level in the reactor pressure vessel, by providing a line (not shown) which interconnects the lower suppression pool 205 and the interior of the reactor pressure vessel 202 or a line (not shown) which interconnects a power portion of the dry well 203 and the interior of the reactor pressure vessel 202. In this regard, a reference be made to the communication line 125 with valves 123, 124 (see FIG. 19) or to a pressure equalizing system 313 which will be described later in connection with FIG. 23. As will be seen from the foregoing description, the fifth embodiment offers the following advantages. In the event of an accident such as a rupture causing discharge of coolant into the dry well, the water component of the discharged coolant is accumulated on the bottom of the dry well but the steam component of the same is introduced via the upper and lower vent pipes 208, 211 into the upper and lower suppression pools 204, 205 so as to be condensed into liquid phase through contact wit water contained in these pools. This prevents overshoot of pressure rise in the dry well which may otherwise be caused due to flow resistance along the vent pipes immediately after the occurrence of the accident. Furthermore, the time required for the water level inside the dry well 203 to reach the submerging level is shortened and the water level in the suppression pool 205 can be raised to the highest level without causing flooding of the openings 212 of the vent pipes 211 in the dry well 203. For these reasons, the submerge cooling of the reactor core can be commenced quickly. Furthermore, the transfer of heat from the dry well 203 to the suppression pool 205 can be promoted by virtue of condensation of vapor component in contact with the water inside the suppression pool 205, partly because the upper hot portion of the water column in the dry well 203 first moves into the suppression pool 205 and partly because for a long period of time the water in the vent pipes 211 can easily be displaced to allow vapor component to flow from the dry well 203 into the suppression pool 205. In addition, transfer of heat from the suppression pool 205 to the outer peripheral pool 215 is promoted, because a large heat transfer area, as well as a large temperature differential, is provided between the water inside the suppression pool 205 and the water inside the outer peripheral pool 215. Thus, the fifth embodiment performs quick and efficient cooling of the reactor core, as well as efficient transfer of heat from the dry well 203 to the suppression pool 205 and further to the outer peripheral pool 215 therefrom, thus offering a remarkable improvement in the safety of the natural cooling type reactor core containment vessel. The fifth embodiment offers an additional advantage in that, since the water required for suppression of pressure rise is stored partly in the upper suppression pool and partly in the lower suppression pool, the size of the suppression pools in terms of area or diameter can be reduced as compared with arrangements having only one suppression pool, provided that the depth of water in the pool is the same, which makes it possible to reduce the diameter of the containment vessel. The described fifth embodiment may be modified such that water is initially charged only in the upper suppression pool 204 while the lower suppression pool 205 is kept empty. In such a modification, the water in the upper suppression pool 204 condenses and liquefied the vapor portion of the coolant discharged as a result of a rupture, while the lower suppression pool 207 directly receives the atmosphere of the dry well. Then, water is injected from the upper suppression pool 204 into the reactor pressure vessel 202, so that hot water is discharged from the fracture into the dry well 203 and then into the lower suppression chamber 207 so as to be accumulated in the lower suppression pool 205. Once the water level in the lower suppression pool 205 is raised above the level of the outlets 213 of the vent pipes 213, vapor component is steadily condensed by the water in the suppression pool 205 for a long time. The merit of the fifth embodiment in regard to the reduction in the size of the containment vessel cannot be fully enjoyed with this modification because the area of the suppression pools cannot be reduced sufficiently. Reduction in the diameter of the containment vessel, however, is possible to a certain extent because the upper suppression pool 204 can be used as a water injection source which does not necessitate any additional equipment such as a pump and because the lower suppression pool 205 may be designed to have a smaller area than the upper suppression pool 204. In addition, the effect to cool the containment vessel is enhanced after the water level is raised in the lower suppression pool 205. A sixth embodiment will be described with reference to FIGS. 22 to 27. Referring to FIG. 22, a boiling water reactor has a pressure vessel 302 which accommodates a reactor core 301 and a containment vessel 303 which contains the pressure vessel 302. The containment vessel 303 has a dry well 302 and a suppression chamber 306. The dry well 302 receives the pressure vessel 302 and a primary line through which a fluid of high pressure and high temperature is circulated. The bottom portion of the suppression chamber 306 defines a suppression pool 305 with a wet well 306A being provided above the suppression pool 305. The dry well 304 and the suppressing pool 305 communicate with each other through a plurality of vent tubes 307 the lower ends of which are immersed in water. The containment vessel 303 is made of a steel. In the event of a loss-of-coolant accident, decay heat is transferred to water in the suppression pool 305 and further to an outer peripheral pool 308 through the steel wall of the containment vessel 303. Referring to FIG. 23, the pressure vessel 302 is provided with a reducing valve 309. A gravity pool 310, which serves as the water supply source of an emergency core cooling system (ECCS) 311, is disposed in an upper portion of the space inside the containment vessel 303. In a short period immediately after occurrence of a loss-of-coolant accident, the reducing valve 309 is opened to relief vapor from the pressure vessel 302 so as to quickly reduce the pressure inside the pressure vessel 302, and the water in the gravity pool 310 is supplied by the force of gravity, i.e., by the difference in the water head, into the pressure vessel 302, thereby cooling the reactor core 301. As an alternative, the ECCS 311 may employ, in place of the gravity pool 310, a pressure accumulator tank 312 installed outside the containment vessel 303 as shown in FIG. 24. In case of a emergency, water is driven by the pressure accumulated in the pressure accumulator tank 312 so as to be supplied into the pressure vessel 302 to cool the reactor core 301. Thus, water is supplied into the pressure vessel 302 from the gravity pool 311 or from the pressure accumulator tank 312 for a short period immediately after the occurrence of the accident. The water thus supplied effectively cools the reactor core 301 and is discharged into the dry well 304 through the fracture (not shown) so as to fill the lower portion of the space inside the dry well 304 as denoted by 317 in FIG. 25. As a consequence, lower half part of the pressure vessel 302 is immersed. The pressure vessel 302 also is equipped with an equalizing system 313 which utilizes, as water sources, the water pooled in the suppression pool 305 and the draw-down water accumulated on the bottom portion of the dry well 304. The equalizing system 313 serves to supply water from the suppressing pool 305 and/or the draw-down water 317 from the dry well 304 into the pressure vessel for a long period after the occurrence of the loss-of-coolant accident. More specifically, the equalizing system 313 includes a first equalizer line 314a having one end connected to the pressure vessel 302 and the other end opening in the suppression pool 305, a second equalizer line 314b shunting from the first equalizing line 314a and opening in a lower portion of the dry well 304, a blasting valve 315 disposed in the first equalizing line 314a at a portion between the pressure vessel 302 and the point where the second equalizer line 314b shunts from the first equalizing line 314a, and a check valve 316 provided in each of the equalizer lines 314a and 314b. The height of the connection between the first equalizer line 314a and the pressure vessel 302 (referred to as "height of equalizer line connection") is so determined as to provide a sufficient margin for enabling submerging cooling of the reactor core for a long time after occurrence of a loss-of-coolant accident. Practically, the height of the equalizer line connection is determined to be 50 to 150 cm, preferably 100 cm or so, above the top of the reactor core. As will be described later, the force with which the water is driven from the suppression pool 305 is derived from the head difference between the outlets of the vent pipes 307 and the height of equalizer line connection. In order to obtain a sufficient force for driving the water, the outlets of the vent pipes 307 are set at a level which is 50 to 150 cm, preferably 70 cm or more, above the height of equalizer line connection and which is 100 to 200 cm above the top of the reactor core 301. The term "level of outlets of the vent pipes" means the level of the lower end openings of the vent pipes opening into the suppression pool. When vent pipes 307 have outlets at different levels, the highest one of these levels is determined as the level of outlets of the vent pipes. The force for driving the draw-down water 317 from the dry well 304 is derived from the head difference between the level of the draw-down water 317 and the height of equalizer line connection at which the first equalizer line 314a is connected to the pressure vessel The opening of the second equalizer line 314b opening to the lower portion of the dry well is determined to be substantially the same as the level of the connection to the pressure vessel 302. In addition, the second equalizer line 314b is designed and laid to have minimum length including horizontal portion, and the check valve 316 is provided in the horizontal portion, in order to reduce loss of the water driving power due to stagnation of air in the second equalizer line 314b. In FIG. 23, the portion of the equalizer system inside the one-dot-and-dash line is shown in plan, for the purpose of easier understanding. This applies also to other Figures of the drawings. The quantity of water charged in the gravity pool 310 or the pressure accumulator tank 311 of the ECCS is so determined that the level of the draw-down water in the dry well is maintained above the level of the height of equalizer connection to the pressure vessel, i.e., above the level of the second equalizer line 314b, for a long time after occurrence of the loss-of-coolant accident. In this embodiment, the wall of the containment vessel 303 functions as a heat-transmission wall through which heat is transferred from the water in the suppression pool 305 to the water in the outer peripheral pool 308. In order to obtain a large area of heat transfer through the containment vessel wall, the water level in the suppression pool 35 is set comparatively high. The operation of the equalizer system 313 should not draw water from the suppression pool 305 at such a large rate as to cause a reduction in the water level inside the suppression pool 305. In order to meet this requirement, the opening 314c of the first equalizer line 314a opening in the suppression pool 305 is set to a level near the initial water level in the suppression pool 305, e.g., 50 cm or so below the initial level, so that the water level in the suppression pool 305 does not come down below this level even when water is drawn through the equalizing system 313. To explain in more detail, in this embodiment of the present invention, the equalizer system 303 starts to operate when the pressure inside the pressure vessel 302, which progressively decreases in a long time after occurrence of the accident, is lowered to a level substantially equal to that in the dry well 304, so that water in the suppression pool 305 and the draw-down water 317 in the bottom portion of the dry well 304 are supplied into the pressure vessel 302, as shown in FIG. 25. The power with which water is driven from the suppression pool 305 and the power with which draw-down water 317 is driven from the dry well 304 are respectively given by the following formulae: (1) Power .DELTA.P.sub.1 for driving water from suppression pool 305 ##EQU1## since, EQU P.sub.WW +H.sub.V .multidot..gamma.=P.sub.R (2) (2) Power .DELTA.P.sub.2 for driving draw-down water from dry well ##EQU2## where, P.sub.R : pressure in pressure vessel P.sub.DW : pressure in dry well PA1 P.sub.WW : pressure in suppression chamber PA1 H.sub.V : depth of immersion of vent pipe PA1 H.sub.NV : height difference between vent pipe outlet and equalizer line connection to pressure vessel PA1 H.sub.D : height difference between drawn-down water level and equalizer line connection to pressure vessel PA1 .gamma.: density of water As will be seen from formula (1) above, the driving power .DELTA.P.sub.1 for driving water from the suppressing pool 305 is derived from the water head difference between the level of the outlet of the vent pipe 307 and the level at which the equalizer line 314a is connected to the pressure vessel 302. It will also be seen from formula (3) that the driving power .DELTA.P.sub.2 for driving the draw-down water 317 in the bottom of the dry well 304 is derived from the difference in the water head between the level of the draw-down water and the level at which the equalizer line 314b is connected to the pressure vessel 302. A control system for controlling the above-described equalizer system will be described with reference to FIGS. 26 and 27. Referring first to FIG. 26, the control system includes a water level sensor 319 for sensing the water level LR in the pressure vessel 302, a pressure sensor 320 for sensing the internal pressure PD in the dry well, and a pressure sensor 321 for sensing the pressure PR inside the pressure vessel 302. Signals from the water level sensor 319 and pressure sensors 320, 321 are sent to a controller 322 which operates to activate the reducing valve 309 and the equalizer system 303 in accordance with an operation logic as shown in FIG. 27. When a loss-of-coolant has taken place, the water level L.sub.R in the pressure vessel 302 is lowered while the pressure P.sub.D in the dry well is increased, whereby low-L.sub.R signal and high P.sub.D signal are transmitted from the respective sensors 319 and 320. Upon receipt of these signals, the controller 322 operates to open the reducing valve 309 so that the internal pressure P.sub.R of the pressure vessel 302 is drastically lowered and progressively approaches the level of the pressure inside the dry well 304. The controller 322, upon receipt of a signal indicative of the reduction in the internal pressure of the pressure vessel, operates to open the blast valve 315 in the equalizing system 313. After the blast valve 315 is opened, the coolant is automatically injected when the conditions of the formulae (1) and (3) are met. When either one of these conditions is not met, the check valve 316 in the equalizer system 313 prevents water from being discharged through the equalizer system 313. As will be understood from the foregoing description, in the described embodiment, the blasting valve 315 is opened in the course of the reduction in the internal pressure of the pressure vessel after occurrence of the loss-of-coolant accident, so that water in the suppression pool 305, as well as the draw-down water 317 in the dry well 304, is injected into the pressure vessel 302, thereby cooling the reactor core 301 for a long period after occurrence of a loss-of-coolant accident. Furthermore, in this embodiment which utilizes both the water in the suppression pool 305 and the draw-down water 317 on the bottom of the dry well 304, injection of water into the pressure vessel 302 is performed when the dry well is filled to a level which is slightly higher, e.g., 100 cm or less above, the level at which the equalizer lines 314a and 314b are connected to the pressure vessel 302, without requiring that the lower portion of the dry well 304 is completely filled. It is therefore possible to effectively cool the reactor core 301 by submersion for a long time after occurrence of a loss-of-coolant accident. This makes it possible to reduce the quantity of water initially held in the gravity pool 310 or the pressure accumulator tank 312 of the ECCS. Consequently, the mass which is to be held by upper portion of the nuclear reactor building is reduced to facilitate anti-earthquake designing of the nuclear reactor system. The wall of the containment vessel 303 provides surfaces through which heat is transmitted and transferred from the water in the suppression pool 305 to the water in the suppression pool 308. In the illustrated embodiment, the opening of the equalizer line 314a opening in the suppression pool 305 is set to a level which is only slightly below the initial water level so that the water level in the suppression pool 305 does not come down substantially below the initial level. As a consequence, high water level is maintained in the suppression pool 305 even during operation of the equalizer system 313 so that a large area is preserved for the heat transfer from the water in the suppression pool 305 and the water in the outer peripheral pool 308, thus attaining an improvement in the heat dissipation characteristic. In the event that the quantity of water in the suppression pool 305 has come down to such a level that the opening 314c of the first equalizer line 314a is exposed, the inlet of the second equalizer line 314b is flooded with draw-down water which is necessarily charged into the dry well 304 due to the head balance of water in the containment vessel 303, so that cooling of the reactor core 301 is continued without fail. Consequently, a remarkable improvement in the safety is achieved by a combination of the static cooling of the containment vessel offered by the above-mentioned heat dissipation characteristic and the static cooling of the reactor core performed by the equalizer system 313. A modification of the sixth embodiment will be described with reference to FIG. 28. This modification employs, in place of the blasting valve 315 used as an isolation valve in the equalizer line 314a of the fifth embodiment shown in FIG. 23, a normally-closed electrically driven valve 323 disposed in an equalizer system 313A. The electrically driven valve 323 also is driven and controlled in accordance with a logic similar to that shown in FIG. 27. This modification offers an advantage in that reliability of the isolation valve, which is an electrically driven valve, can easily be confirmed through a periodical check. In addition, operation and administration of the safety system are facilitated by virtue of the use of the electrically driven valve. FIG. 29 shows another modification which is different from the sixth embodiment shown in FIG. 23 in that a parallel connection of a pair of blasting valves 315 and a parallel connection of a pair of check valve 316 are used in each line. In general, it is necessary to assume that a normally-closed valve which is expected to open in the event of a loss-of-coolant accident may fail to open due to an unexpected reason. With the dual valve arrangement employed in this modification, it is possible to supply emergency cooling water to the pressure vessel even when one of the dynamic components, e.g., valve, has failed to operate. |
abstract | A control room for a nuclear power plant including two or more nuclear reactor units includes a central workstation providing monitoring capability for both nuclear reactor units, a first operator at the controls (OATC) workstation in front of and to one side of the central workstation providing monitoring and control capabilities for the first nuclear reactor unit, a second OATC workstation in front of and to the other side of the central workstation providing monitoring and control capabilities for the second nuclear reactor unit, and a common control workstation directly in front of the central workstation providing monitoring and control capabilities for systems serving both the first nuclear reactor unit and the second nuclear reactor unit. The central and common control workstations do not provide control capabilities for either nuclear reactor unit. The common control workstation does not include any control capabilities that must be performed by a licensed operator. |
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summary | ||
claims | 1. A body armor system comprising:an armored element having a front section and a rearwardly spaced rear section connected to the front section, portions of the armored element comprising ballistic armor; anda shirt having portions disposed beneath the armored element, the shirt having two sleeves which are connected to and which extend from a torso element, the torso element being substantially overlain by portions of the armored element, and portions of the sleeves extending beyond the armored element so as not to be overlain thereby, and the shirt having a collar which is connected to the torso element and to the two sleeves, the collar extending upwardly from the armored element, wherein the torso element is composed of a wicking material of a first level of stretch, and the two sleeves and the collar being composed of a durable material of a level of stretch which is less than the first level of stretch. 2. The shirt of claim 1 wherein the collar and the sleeves are formed of a woven material, and the torso element is formed of a knit material. 3. The shirt of claim 1 wherein the collar and the sleeves have a higher clo value than the torso element. 4. The shirt of claim 1 further comprising a closure which extends upwardly from the torso element across the collar. 5. A body armor system comprising:an armored element having a front section and a rearwardly spaced rear section connected to the front section, portions of the armored element comprising ballistic armor; anda shirt having portions disposed beneath the armored element, the shirt having two sleeves which are connected to and which extend from a torso element, the torso element being substantially overlain by portions of the armored element, and portions of the sleeves extending beyond the armored element so as not to be overlain thereby, and the shirt having a collar which is connected to the torso element and to the two sleeves, the collar extending upwardly from the armored element, wherein the torso element is composed of a knit wicking material and the two sleeves and the collar being composed of a woven material, the collar being connected between the two sleeves to restrain excessive downward distortion of the sleeves. 6. The shirt of claim 5 further comprising a closure which extends upwardly from the torso element across the collar. 7. The shirt of claim 5 wherein the collar is connected to the sleeves by connecting strips composed of a woven material. 8. The shirt of claim 7 wherein the sleeves, the collar and the connecting strips are non-conforming and the torso element is conforming. 9. The shirt of claim 5 wherein each sleeve has a portion which extends to the collar. 10. The shirt of claim 7 wherein the sleeves and the collar are non-conforming and the torso element is conforming. |
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053125975 | claims | 1. An apparatus for separating a hydrogen isotope from a gaseous mixture, said apparatus comprising: a housing having an inlet and an outlet; a conduit disposed within said housing, said conduit having an interior and an exterior, and an inlet and an outlet; particles of a hydride carried in said interior of said conduit, said hydride particles having a first temperature; a fluid in said housing and adjacent said exterior of said conduit, said fluid having a second temperature, said second temperature different from said first temperature; and means in spaced relation to said exterior of said conduit for creating turbulent flow of said fluid over said conduit so that said fluid can exchange heat with said hydride particles for adsorbing and desorbing said hydrogen isotope. a substrate; and palladium carried by said substrate. a housing; a conduit carried within said housing having an interior and an exterior, said conduit dimensioned for holding a quantity of hydride particles, said conduit formed into a coil; means for changing a temperature of a fluid when said fluid flows over said exterior of said coiled conduit; and means in spaced relation with said housing for causing said fluid to flow over said exterior of said coiled conduit so that heat energy can be exchanged between said fluid when said fluid is flowed over said exterior of said coiled conduit and said hydride particles are in said interior of said coiled conduit, said changing means comprising a baffle in spaced relation to said coiled conduit, said coiled conduit wound about said baffle, said baffle and said coiled conduit combining to confine said fluid and accelerate said fluid flow. 2. The apparatus as recited in claim 1, wherein said creating means is a cylindrical baffle in said housing. 3. The apparatus as recited in claim 1, wherein said creating means is a cylindrical baffle in said housing and said conduit is coiled around said baffle. 4. The apparatus as recited in claim 1, further comprising a pump for pumping said fluid through said housing from said inlet to said outlet of said housing. 5. The apparatus as recited in claim 1, further comprising a first pump for pumping a gaseous mixture through said conduit and a second pump for pumping said fluid through said housing over said exterior of said conduit. 6. The apparatus as recited in claim 1, further comprising means for raising and lowering said first temperature with respect to said second temperature. 7. The apparatus as recited in claim 1, wherein each hydride particle of said hydride particles comprises: 8. An apparatus for separating a hydrogen isotope from a gaseous mixture, said apparatus for use with hydride particles and a fluid, said apparatus comprising: |
053348478 | abstract | A composition for use as a radiation shield. The shield has a depleted urum core for absorbing gamma rays and a bismuth coating for preventing chemical corrosion and absorbing gamma rays. Alternatively, a sheet of gadolinium may be positioned between the uranium core and the bismuth coating for absorbing neutrons. The composition is preferably in the form of a container for storing materials that emit radiation such as gamma rays and neutrons. The container is preferably formed by casting bismuth around a pre-formed uranium container having a gadolinium sheeting, and allowing the bismuth to cool. The resulting container is a structurally sound, corrosion-resistant, radiation-absorbing container. |
060552964 | abstract | A radiographic grid with reduced line density artifacts. The radiographic grid includes a grid housing sized to receive a plurality of x-ray radiation absorbing lamellae. Each of the plurality of lamellae has a foil strip applied to its lower end portion. The foil eliminates the lamella line artifacts otherwise emanating from the lamellae. |
051376830 | claims | 1. A process for producing a chromium oxide insulating layer between pellets and a cladding (5) of a fuel element for a nuclear having a zirconium alloy cladding (5) and a stack of pellets of sintered fuel material which are introduced into the cladding (5) during manufacture of the fuel element, comprising the steps of bringing an organometallic chromium compound into contact, in gaseous form, with at least one substrate consisting of the inner surface of the cladding (5) and/or the outer surface of the fuel pellets, in the presence of an oxidizing gas, and maintaining said substrate at a temperature between 300.degree. and 600.degree. C. in order to produce chemical vapor deposition of chromium oxide. 2. The process as claimed in claim 1, wherein the organometallic chromium compound is chromium acetylacetonate (C.sub.5 H.sub.7 O.sub.2).sub.3 -Cr, and wherein the temperature of the substrate is between 350.degree. and 600.degree. C. 3. The process as claimed in claim 2, wherein the chromium acetylacetonate is heated to a temperature between 180.degree. and 200.degree. C. before being brought into contact with the substrate. 4. The process as claimed in claim 1, wherein the vapor of the organometallic chromium compound is entrained by a vector gas ensuring its contact with the substrate. 5. The process as claimed in claim 4, wherein the vector gas is argon. 6. The process as claimed in claim 4, wherein the oxidizing gas consists of oxygen mixed in a given proportion with the vector gas and with the organometallic compound vapor. 7. A fuel element for a nuclear reactor having a tubular zirconium alloy cladding (5) and a stack of pellets of sintered fuel material arranged inside the cladding (5), comprising at least one chromium oxide insulating layer less than 10 .mu.m thick inserted between the inner surface of the cladding (5) and that outer surface of the pellets situated facing the inner surface of the cladding (5). 8. The fuel element as claimed in claim 7, wherein the insulating layer is deposited on the inner surface of the cladding (5). 9. Fuel element as claimed in claim 7, wherein the insulating layer is deposited on the outer surface of the pellets of fuel material. 10. The fuel element as claimed in claim 7, wherein the insulating layer consists of both a coating deposited on the inner surface of the cladding (5) and of a coating deposited on the outer surface of the pellets. 11. The fuel element as claimed in claim 7, wherein the cladding (5) is coated on its outer surface with a chromium oxide layer. |
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