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description | The invention relates to controlling ion implantation during vacuum fluctuation. In particular, the invention relates to controlling an ion beam implantation process to compensate for vacuum fluctuation based on a measured beam current and not a measured pressure. Ion implantation is a standard technique for introducing conductivity-altering impurities into semiconductor wafers. A typical ion implantation process uses an energetic ion beam to introduce impurities into the semiconductor wafers. As is well-known, introducing the impurities at a uniform depth and density into the wafers is important to ensure that the semiconductor devices being formed operate within specification. One factor in the ion implantation process that can affect the uniformity of the impurity dose in the wafer is vacuum fluctuations during the implantation process. The vacuum fluctuations can be caused by photoresist or other materials coated on a semiconductor wafer that outgas, volatilize or sputter when the ion beam impacts the semiconductor wafer. The outgassing, volatilization or sputtering releases gas particles, which cause a pressure rise in the normally high vacuum condition along the beam line and can result in collisions between ions in the beam and released particles. These collisions can cause ions in the beam to experience a charge change. For example, singly-charged positive ions in an ion beam may collide with residual gas atoms produced by photoresist outgassing during implantation, and experience a charge exchange without a significant change in kinetic energy. The singly-charged positive ions may be neutralized by the collisions and impact the semiconductor wafer in the neutral charge state. In contrast, when outgassing, volatilization or sputtering does not occur from the semiconductor wafer surface, the vacuum level can remain relatively high and constant along the beam line, thus resulting in fewer ion charge exchanging collisions. The charge exchanging collisions that result when the vacuum level along the beam line drops can cause problems because the detectors used to determine and control the ion beam current (and also the total dose of the wafer) during implantation typically only detect charged particles, but not neutral particles. The neutral particles that are implanted in the wafer are the desired implantation species and have the desired energies for implantation and thus, should be counted in the total implant dose. Since the typical ion beam current detector, such as a Faraday cup, is not capable of detecting the neutral particles, neutral particles that should be counted as contributing to the wafer dose are not detected. As a result, a beam current that is less than the actual beam current is detected, thereby prompting an increase in the beam current and overdosing of the wafer. Previous methods for controlling implantation uniformity during vacuum fluctuations include detecting both the ion beam current and the vacuum level in the implantation chamber and controlling the ion beam accordingly, as disclosed in U.S. Pat. No. 4,587,433 to Farley and U.S. Pat. No. 5,760,409 to Chen. Such systems have drawbacks, including improper control caused by a delay between an actual change in vacuum along the beam line and the time when the vacuum change, i.e., a pressure change, is detected. This delay between actual vacuum change and detection can cause a delay in ion beam control and result in improper wafer dosing. This type of method also has the disadvantage of requiring the empirical correlation of a plurality of detected gas pressure and beam current values with a corresponding correction value for a plurality of sets of ion implantation parameters, such as gas composition, beam energy, implant species, amount of dose, photoresist type, etc. Accordingly, a method for controlling implantation during vacuum fluctuation that is independent of such implantation parameters and that can rapidly respond to vacuum fluctuation is needed. The invention provides methods and apparatus for controlling an ion beam implantation process in the presence of vacuum fluctuation along the beam line. In one aspect of the invention, vacuum fluctuations can be detected based on a detected ion beam current, and not a detected pressure. Thus, an ion implantation apparatus and method may correct for vacuum fluctuations based on beam current without detecting pressure within an implantation chamber. In one illustrative embodiment, a method for controlling an ion implantation process includes generating an ion beam, and directing the ion beam along a beamline. A beam current is detected along the beamline, and a material is implanted with ions in the ion beam. Vacuum fluctuations during implantation are compensated for based on detected beam current and not based on a detected pressure. In one illustrative embodiment, an ion beam including energetic particles for implantation into a semiconductor wafer is generated. A reference value for the ion beam current is then determined. The reference value for the ion beam current can be determined by actually measuring the ion beam current while a vacuum level along the beam line is at a desired level, e.g., is at a relatively high and stable level before implantation of the semiconductor wafer begins. The reference value for the ion beam current can also be a stored value that is retrieved from memory or input by a human operator, for example. Once the reference value for the ion beam current is determined, implantation of the semiconductor wafer is performed. The ion beam current is measured during implantation, and a difference between the reference value for the ion beam current and the measured ion beam current is determined. The ion beam current, a wafer scan rate or other implantation process parameters can be adjusted based on the difference between the reference value and the measured ion beam current. Since the measured ion beam current includes information regarding vacuum fluctuation, e.g., a decrease in measured beam current may be assumed to be caused mainly by vacuum fluctuation, the ion implantation process can be adjusted accordingly, e.g., a wafer scan rate may be decreased in response to a decrease in detected beam current. In another illustrative embodiment of the invention, a difference between the reference value and the measured beam current can be scaled and the scaled difference value used to control ion implantation process parameters. For example, the ion implantation system can include an angle corrector magnet that bends and collimates the ion beam before the beam is incident on a semiconductor wafer. In this type of arrangement, vacuum fluctuations along the beam line can cause an increase in ion charge exchanging collisions that occur in two regions along the beam line, i.e., a line-of-sight region and a non-line-of-sight region. Charge exchanging collisions in the non-line-of-sight region that neutralize a particle cause the particle to not be deflected by the angle corrector magnet along the path to the wafer. Instead, the neutral particles follow a path that does not impact the wafer. Thus, although the non-line-of-sight collisions cause a decrease in detected ion beam current at the wafer, the collisions also cause a decrease in the dose rate at the wafer. In contrast, charge exchanging collisions that occur in the line-of-sight region neutralize energetic particles, but since the particles are traveling along a path toward the wafer, the neutral particles still are implanted in the wafer and contribute to the total dose. However, since the particles are neutralized, they are not detected as part of the total ion beam current. Thus, collisions that neutralize particles in the line-of-sight region cause a decrease in detected ion beam current, but not necessarily a change in the wafer dose rate. Scaling of the difference value between the reference value for the ion beam current and the measured ion beam current may be used to adjust the implantation process parameters (e.g., the wafer scan rate) to compensate for both line-of-sight and non-line-of-sight collisions. For example, a detected decrease in ion beam current may be assumed to be largely due to vacuum fluctuations along the beam line. A portion of the detected decrease in ion beam current results from non-line-of-sight collisions, while the remaining portion of the detected decrease in ion beam current results from line-of-sight collisions. Since the line-of-sight collisions contribute to the decrease in ion beam current, but do not necessarily affect the total dose delivered to the wafer, this portion of the detected decrease in ion beam current need not necessarily be compensated for. Thus, the difference value between the reference level for the ion beam current and the detected ion beam current may be scaled to compensate mainly for non-line-of-sight collisions and to adjust the wafer dose rate to an appropriate level. The invention is described below in connection with an ion implantation system. However, the invention can be used with other systems or processes that use beams of energetic, charged particles, such as electron beam imaging systems. Thus, the invention is not limited to the specific embodiments described below. FIG. 1 is a schematic block diagram of an ion implantation system 100 in accordance with the invention. The ion implantation system 100 includes a beam generator 1 that generates and directs a beam 2 toward a semiconductor wafer 3. The beam generator 1 can include various different types of components and systems to generate a beam 2 having desired characteristics. The beam 2 can be any type of charged particle beam, such as an energetic ion beam used to implant the semiconductor wafer 3. The semiconductor wafer 3 can take various physical shapes, such as the common disk shape. The semiconductor wafer 3 can include any type of semiconductor material or any other material that is to be implanted using the beam 2. A beam current, i.e., an amount of charge carried by particles in the beam 2 to the wafer 3, is measured by a detector 4. The detector 4 can be any type of device that detects a level of the beam 2 current. For example, the detector 4 can be a Faraday cup or other device, as are well known in the art. The detector 4 can be fixed in place or movable and can be positioned in a variety of different ways, such as along the beam 2 path to the wafer 3, adjacent the wafer 3 as shown in FIG. 1, behind the wafer 3, etc. Other types of devices to measure the beam 2 current, such as devices that use calorimetry or beam-induced magnetic field measurement can be used, if desired, as the detector 4. The detector 4 outputs a signal representing the detected beam 2 current to a controller 5. The controller 5 can be or include a general purpose computer or network of general purpose computers that are programmed to perform desired input/output and other functions. The controller 5 can also include other electronic circuitry or components, such as application specific integrated circuits (e.g., ASICs), other hardwired or programmable electronic devices, discrete element circuits, FPGAs, etc. The controller 5 can also include devices, such as user input/output devices (keyboards, touch screens, user pointing devices, displays, printers, etc.), communication devices, data storage devices, mechanical drive systems, etc., to perform desired functions. The controller 5 also communicates with a wafer drive 6 that is capable of moving the wafer 3 relative to the beam 2, e.g., the wafer drive 6 can scan the wafer 3 across the beam 2 to implant the wafer 3. The wafer drive 6 can include various different devices or systems to physically move the wafer 3 in a desired way. For example, the wafer drive 6 can include servo drive motors, solenoids, screw drive mechanisms, one or more air bearings, position encoding devices, mechanical linkages, robotic arms or any other components to move the wafer 3 as are well-known in the art. The beam 2 is transported from the beam generator 1 to the wafer 3 in a relatively high vacuum environment created in a housing 8 by a vacuum system 7. By high vacuum, it is meant that low pressure exists in the housing 8. Conversely, low vacuum refers to a relatively higher pressure in the housing 8. The vacuum in the housing 8 is maintained using well-known systems, such as vacuum pumps, vacuum isolation valves, pressure sensors, etc. The vacuum system 7 may communicate with the controller 5, e.g., to provide information to the controller 5 regarding the current vacuum level in one or more portions of the housing 8. The beam 2 is shown in FIG. 1 to follow a straight path from the beam generator 1 to the wafer 3. However, the beam 2 may follow a curved path with one or more deflections within the generator 1 and/or between the beam generator 1 and the wafer 3. The beam 2 can be deflected, for example, by one or more magnets, lenses or other ion optical devices. Prior to implantation, the wafer drive 6 can move the wafer 3 away from the beam 2 so that the beam 2 is not incident on the wafer 3. The beam generator 1 then generates a beam 2 and the detector 4 detects a reference level for the beam current while a vacuum level within the housing 8 is at a desired level and/or is stable. As one example, the vacuum level at which the reference level for the beam current is determined may be a highest vacuum level generated by the vacuum system 7 within the housing 8. Of course, the reference level for the beam current may be determined for other vacuum-levels within the housing 8. The detector 4 outputs a signal to the controller 5 that can be used by the controller 5 as the reference level for the beam current, or the controller 5 can process the signal to generate a reference level for the beam current. For example, the detector 4 may output an analog signal that represents a number of detected ions, and the controller 5 may convert the analog signal to a digital number that is stored within the controller 5. The stored digital number may be used as a reference level for the beam current. During implantation, the beam 2 is incident on at least a portion of the wafer 3. The beam 2 can be scanned across the wafer 3 and/or the wafer 3 can be scanned across the beam 2 by the wafer drive 6. For example, the beam 2 may be scanned by the beam generator 1 in a plane parallel to the paper in FIG. 1, while the wafer 3 is moved in a direction perpendicular to the paper by the wafer drive 6. Materials in or on the wafer 3, such as photoresist on the surface of the wafer 3, may outgas or otherwise produce materials when impacted by particles in the beam 2. This causes a vacuum fluctuation within the housing 8 that can cause the vacuum level to decrease near the wafer 3 and along the beamline. This decrease in vacuum level can cause an increase in the number of charge extending collisions that occur for particles in the beam 2 traveling to the wafer 3. As discussed above, the charge exchanging collisions, i.e., collisions between energetic particles in the beam 2 and mateirals released by outgassing or volatilization at the wafer 3, cause the charge of individual particles in the beam 2 to be changed. For example, singly positively charged ions in the beam 2 can be neutralized by collisions along the beamline, or the positively charged ions may be doubly positively charged. Although the charge of the ions can be altered, the energy of the particles is not substantially changed. Therefore, although the charge of some particles may be altered so that the detector 4 does not detect the presence of the particles, the particles may impact the wafer 3 and contribute to the overall impurity dosing of the wafer 3. Thus, the detector 4 may output a signal during implantation that indicates a decrease in beam current even though the total dosing of the wafer 3 is not affected. The controller 5 can recognize, i.e., operate based on an assumption, that the detected decrease in beam current, or a portion of a detected decrease in beam current, has been caused by vacuum fluctuations during implantation, but that the total dose implanted in the wafer 3 is not being affected. Thus, the controller 5 can detect a vacuum fluctuation based on a detected decrease in beam current. It should be understood that the beam current may vary during implantation due to other factors, such as ion source variations, and that the controller 5 may determine that some portion of a detected beam current decrease has been caused by vacuum fluctuations, while another portion of the decrease has been caused by other factors, e.g., variations at the ion source. The controller 5 may adjust certain implantation parameters to correct for variations in beam current that are not due to vacuum fluctuations, as is known in the art and not described here. In addition, outgassing may vary with time, and the controller 5 may determine that the contribution of vacuum fluctuation to detected beam current decrease as compared to other factors may vary over time during implantation. In such cases the controller 5 may use an adjusted measured beam current that reflects only the contribution of vacuum fluctuation, and not the contribution of other factors, for purposes of controlling implantation. The controller 5 may sense a decrease in beam current, but not necessarily adjust specific implantation parameters, such as a beam 2 scan rate, wafer 3 scan rate, etc. Instead, the controller 5 may output a signal to the vacuum system 7 indicating that a rise in vacuum pressure has been detected and that the vacuum level within the housing 8 should be adjusted accordingly. This signal to the vacuum system 7 may be provided in addition to measured vacuum level signals provided by pressure sensors to the vacuum system 7. Thus, based on the signal from the controller 5, the vacuum system 7 may begin adjusting the vacuum level within the housing 8 before a decrease in vacuum level is detected by pressure sensors associated with the vacuum system 7. Alternately, the controller 5 may compare a detected beam current level provided by the detector 4 during implantation with the stored reference level for the beam current and use the difference between the two values to control either the beam 2 or the wafer drive 6. For example, the controller 5 may determine (based on stored information) that the decrease in beam current detected by the detector 4 during implantation is largely due to vacuum fluctuations along the beam line. Further, the controller 5 can determine that a portion of the detected decrease in beam current due to charge exchanging collisions does not affect the total dose delivered to the wafer 3, while another portion of the detected decrease in beam current does contribute to a decrease in the total dose delivered to the wafer 3. For example, some charge exchanging collisions may neutralize beam particles without affecting the particles' kinetic energy. The neutralized particles will not be detected by the detector 4, but still contribute to the total dose implanted in the wafer 3. Other collisions caused by the vacuum fluctuation may cause the charge and kinetic energy of a particle to be altered, or cause the particle to follow a trajectory that prevents the particle from being implanted in the wafer 3. These latter collisions cause a decrease in detected beam current, and also a decrease in the total dose implanted in the wafer 3. The controller 5 can scale the difference value between the detected beam current and the reference value for the beam current, so that a total dose delivered to the wafer 3 is adjusted to a desired level. The difference value can also be normalized, e.g., by dividing the difference value by the reference value. For example, the controller 5 may control the wafer drive 6 to move the waver 3 more slowly across the beam 2 path based on the scaled and normalized scaled reference value. The scaling factors used by the controller 5 can be determined empirically and stored in the controller 5. Thus, when a particle difference value is determined by the controller 5, a corresponding scaling factor can be retrieved and used to adjust the difference value to appropriately control the beam 2 or movement of the wafer 3. The controller 5 may also control for implantation non-uniformity in two dimensions caused by vacuum fluctuation. For example, vacuum fluctuation may cause implantation non-uniformity in two dimensions at the wafer 3, e.g., non-uniformity along the beam scan direction parallel to the paper in FIG. 1 and non-uniformity along the wafer scan direction perpendicular to the paper in FIG. 1. Thus, the controller 5 may control the beam scan rate and the wafer scan rate to control for non-uniformity in both directions. Different scale factors may also be used to control the beam scan rate and the wafer scan rate, respectively. FIG. 2 shows a more detailed schematic block diagram of an ion implantation system 100 in accordance with the invention. An ion source 11 generates ions and supplies an ion beam 2. As is known in the art, the ion source 11 may include an ion chamber and a gas box containing a gas to be ionized. The gas is supplied to the ion chamber where it is ionized. The ions thus formed are extracted from the ion chamber to form the ion beam 2. The ion beam 2 may have an elongated cross section and may be ribbon-shaped, with a long dimension of the beam 6 cross section preferably having a horizontal orientation, or the beam 2 may have a circular cross-section. An extraction power supply and extraction electrode 12 accelerate ions from the ion source 11. The extraction power supply may be adjustable, for example, from about 0.2 to 80 keV. Thus, ions from the ion source 11 may be accelerated to energies of about 0.2 to 80 keV by the extraction electrode 12. The construction and operation of ion sources are well-known to those skilled in the art. The ion beam 2 passes through a suppression electrode and a ground electrode (not shown) to a mass analyzer 13. The mass analyzer 13 includes a resolving magnet that deflects ions in the ion beam 2 such that ions of a desired ion species pass through a resolving aperture 14 and undesired ion species do not pass through the resolving aperture 14. In a preferred embodiment, the mass analyzer 13 resolving magnet deflects ions of the desired species by 90°. Ions of the desired ion species pass through the resolving aperture 14 to a scanner 15 (which is not required for systems using a ribbon-beam) positioned downstream of the mass analyzer 13. The scanner 15 may include scanning electrodes as well as other electrodes (not shown) for controlling the beam 2. Ions in the ion beam 2 are scanned and then pass through an angle corrector magnet 16. The angle corrector magnet 16 deflects ions of the desired ion species and converts the ion beam 2 from a diverging ion beam to a nearly collimated ion beam 2 having substantially-parallel ion trajectories. In a preferred embodiment, the angle corrector magnet 16 deflects ions of the desired ion species by 70°. An end station 17 supports one or more semiconductor wafers in the path of the ion beam 2 such that ions of the desired species are implanted into the semiconductor wafers (not shown). The end station 17 may include a cooled electrostatic platen and a wafer drive 6 for moving wafers perpendicular to the long dimension of the ion beam 2 cross section, so as to distribute ions over the wafer surface. The ion implantation system 100 may include additional components known to those skilled in the art. For example, the end station 17 typically includes automated wafer handling equipment for introducing wafers 3 into the ion implantation system 100 and for removing wafers after implantation. The end station 17 may also include a dose measuring system, an electron flood gun and other known components. It will be understood that the entire path traversed by the ion beam 2 is evacuated during ion implantation. Additional details of the ion implantation system 100 are not provided here since they are well-known in the art and are not necessarily important to the invention. The components of the ion implantation system 100 are controlled by a controller 5. Thus, the controller 5 can monitor the ion implantation process and take steps to adjust various aspects of the process, such as the rate at which ions are produced from the ion source 11, the scan rate of the ion beam 2, the scan rate of the wafer relative to the ion beam 2 at the end station 17, etc. FIG. 3 is a schematic diagram of a portion of the ion implantation system 100 in FIG. 2 from the resolving aperture 14 to the end station 17, and shows the beam 2 path from the resolving aperture 14 to the semiconductor wafer 3 in the end station 17. The ion beam 2 path is conceptually divided into three regions, where Region I is upstream of the angle corrector magnet 16, Region II is within the angle corrector magnet 16, and Region III is downstream of the angle corrector magnet 16. In Region I, the beam 2 is a diverging beam. In Region II, the angle corrector magnet 16 deflects the charged articles in the beam 2 by approximately 70° and collimates the beam. Thus, the beam 2 in Region III is a substantially collimated beam such that ions in the beam are all incident on the wafer 3 surface at substantially the same angle. Before implantation of the wafer 3 begins, the controller 5 controls the wafer drive 6 to move the wafer 3 out of the path of the beam 2. The beam current of the ion beam 2 is then measured by one or more of the detectors 41, 42 and 43. The detector 41 is a profiler or traveling Faraday cup that is moved transverse to the beam 2 near the plane where the beam 2 is incident on the wafer 3. Use of this type of traveling Faraday detector 41 is well-known, and a signal produced by the detector 41 is used to determine uniformity of the beam 2. The beam current can also be measured by a Faraday detector 42 that is positioned adjacent the wafer 3 during implantation. Since the detector 42 is positioned adjacent the wafer 3, the detector 42 can detect the beam current either before the wafer 3 is positioned for implantation or during implantation of the wafer 3. A third detector 43 can also be used to detect beam current. Since the detector 43 is positioned downstream of the wafer 3 during implantation, the detector 43, which is typically a Faraday-type detector, is typically used before implantation to measure a total dose expected to be implanted in the wafer 3. It should be understood that although the detectors 41-43 in this example are Faraday-type detectors, other types of detectors for sensing the beam current, such as those using calorimetry or beam induced magnetic field measurement, may be used in addition to, or in place of, the detectors 41-43. In addition, it is not necessary to use all three detectors 41-43. For example, the detectors 41 and 43 may be eliminated or not used to detect the beam current to control the implantation process with respect to vacuum fluctuation. Thus, only the detector 42 may be used to detect beam current. In the following example, only the detector 42 is used to detect the ion beam current and control the implantation process during vacuum fluctuation. When a vacuum level within the end station 17 and along the beam line is at a desired reference level, e.g., when the wafer 3 is positioned away from the beam 2 path and the vacuum along the beam line is at a relatively high and uniform level, the detector 42 detects the beam current of the ion beam 2. This detected current measured when the vacuum level is at a reference level is used by the controller 5 as a reference value for the ion beam current. The reference value for the ion beam current is not necessarily determined when the wafer 3 is out of the beam 2 path. That is, the reference value for the ion beam current may be a detected beam current when the wafer 3 is in a position to be implanted, and the vacuum along the beamline is at a desired level, e.g., at the beginning of implantation of the wafer 3. Alternately, the controller 5 may use an empirically determined reference value for the beam current that is stored within the memory of the controller 5. After implantation of the wafer 3 begins, substances in or on the wafer 3, such as photoresist, may be begin to outgas or otherwise release particles. This particle release causes a fluctuation in the vacuum level near the wafer 3 and along the beam line, e.g., along the beam path back to the resolving aperture 14. As discussed above, a decrease in the vacuum level along the beamline can cause an increase in the number of charge exchanging collisions between ions in the beam 2 and other particles, such as those released from the wafer 3. These charge exchanging collisions can cause ions in the beam 2 to become neutralized. Neutral particles are not acted on by the angle corrector magnet 16, and thus the neutral particles follow a straight line path from the point at which the particle was neutralized. Charge exchanging collisions can occur anywhere along the beamline, and the location of the charge exchanging collision typically determines whether the neutralized particle impacts the wafer 3 and contributes to the overall dose implanted in the wafer 3. In this example, charge exchanging collisions that neutralize ions in the beam 2 and that occur upstream of a line 9 cause the neutralized particles to not impact the wafer 3 or to be measured as contributing to the beam current. However, charge exchanging collisions that neutralize ions in the beam 2 and that occur downstream of the line 9 result in the neutralized particles impacting the wafer 3. The path of neutralized particles in FIG. 3 is shown by dashed line trajectories. Since the detector 42 cannot detect neutralized particles, the detector 42 detects a decrease in beam current, even though some of the neutralized particles that undergo charge exchanging collisions downstream of the line 9 contribute to the overall dose of the wafer 3. Charge exchanging collisions that occur upstream of the line 9 are termed non-line of sight collisions, while collisions that occur downstream of the line 9 are termed line of sight collisions, since the neutralized particles have a line of sight to the wafer and contribute to the dose of the wafer 3. Neutralizing of particles in the beam 2 when the vacuum level drops along the beamline causes the detector 42 to detect a decrease in the beam current. A difference ΔI between the reference value for the beam current Iref and the measured beam current Im, is contributed to by both non-line of sight collisions and line of sight collisions that neutralize particles in the beam 2. Thus, the beam current difference ΔI is a function of both the line of sight collisions and non-line of sight collisions. Since the line of sight collisions contribute to the overall dose of the wafer 3, the controller 5 cannot simply either increase the beam current of the beam 2 so that the measured current Im is equal to the reference level Iref or adjust the scan rate of the wafer 3 to compensate for the beam current difference ΔI, without taking into account the neutralized particles that are implanted in the wafer 3, but do not contribute to the detected beam current. Accordingly, the difference value ΔI is scaled using an appropriate scale factor, and the scaled value is used to control the wafer scan rate, beam current, or other implantation parameters to achieve a desired dose of the wafer 3. Preferably, the scan rate of the wafer 3 is adjusted. Scale factors may generally represent an estimation of the percentage amount of a detected decrease in measured beam current Im that is caused by non-line of sight collisions. Scale factors can be determined empirically, e.g., an appropriate scale factor may be determined based on a measured beam current Im and the actual dose implanted in the wafer 3 during an actual implantation process. Scale factors may also be determined by mathematically modeling an implantation process. For example, scale factors may be based on calculated distance*density products that are obtained from implantation system models. Neutral particle densities, e.g., from outgassing products and other sources, may be calculated based on a model of the vacuum system, and the beam path length*neutral particle density may be determined for line-of-sight and non-line-of-sight paths. Based on the relative values of these distance*density products, the scale factors for the implantation system may be derived. An empirical approach may be more accurate than a modeling approach for determining scale factors, but using an empirical approach may be more time consuming. As one example of how a scale factor may be used, if a 50% decrease in beam current is detected, e.g., a difference value ΔI is normalized by dividing ΔI by the reference value Iref giving ΔI/Iref=0.5, and half of the detected decrease in beam current is caused by non-line of sight collisions and the other half of the detected decrease in beam current is caused by line of sight collisions, the wafer scan rate may be decreased by ¼ of its original value (adjusted by a scale factor of 0.25) to achieve a desired dose rate for the wafer 3. That is, although the detected beam current Im is half the reference value for the beam current Iref, only ½ of the decrease in detected beam current need be compensated for since ½ of the detected decrease in beam current is caused by neutralized particles that still contribute to the overall dose of the wafer 3. Thus, a 25% increase in detected beam current Im or a 25% decrease in the wafer scan rate may be used to compensate for the non-line of sight collisions causing an overall decrease in the dosing of the wafer 3. The above example is an overly simplified example used to describe one aspect of the invention. It should be understood that other differences in detected beam current and a reference value for the beam current may be determined, and that other ratios of charge exchanging collisions that do and do not contribute to the overall wafer 3 dose may be encountered. In addition, non-uniformity effects in two dimensions may be compensated for by the controller 5. For example, empirically-derived scale factors may be used to adjust both a beam scan rate and a wafer scan rate to adjust for non-uniform dosing in both the beam scan and wafer scan directions that is caused by vacuum fluctuations. Of course, the scale factors may be derived mathematically, e.g., by modeling beam paths and neutral particle densities in the corrector magnet 16 to estimate the beam scan direction and wafer scan direction non-uniformity and scan rates to adjust for the non-uniformity in two directions. The controller 5 can use the determined beam current difference ΔI for other purposes, such as controlling the vacuum system in the ion implantation system 100 to compensate for the vacuum fluctuations during implantation. The controller 5 may also use the beam current difference information to control other parameters of the implantation process, such as to control scanning of the beam 2, e.g., by adjusting a scanning waveform applied to scan plates in the scanner 15, etc. Control of scanning of the beam 2 may be used to adjust for horizontal non-uniformity, i.e., implantation non-uniformity in the wafer 3 in a direction perpendicular to the wafer scan direction. The controller 5 can also use the determined beam current difference information to make adjustments for beam non-uniformity effects of the vacuum fluctuations along the beam line. As can be seen in FIG. 3, ions on the outside of the beam envelope travel a longer distance from the resolving aperture 14 before reaching the line 9 than ions traveling on the inside of the beam envelope. Thus, ions traveling on the outside of the beam envelope may have a higher probability of experiencing a charge exchanging collision before reaching the line 9. This alone may result in underdosing one side of the wafer 3 (e.g., the right side of the wafer 3 in FIG. 3) compared to the other side. However, ions traveling along the inside of the beam envelope travel some distance from the line 9 before reaching Region III. If an ion experiences a charge exchanging collision between the line 9 and Region III, the neutralized particle is no longer steered by the angle corrector magnet 16 and follows a straight line path toward a portion of the wafer 3 further to the right than the ion would have originally impacted. This tends to cause extra dosing of the right side of the wafer 3, and may nearly counteract charge exchanging collisions that more frequently occur along outer paths of the beam 2. The inventor has found that the effects of vacuum fluctuations on beam uniformity in the arrangement shown in FIG. 3 typically have a relatively small effect on implant uniformity. However, in other configurations, vacuum fluctuations may have a greater effect on implant uniformity. In those cases, the controller 5 may adjust the beam parameters to counteract the effects of vacuum fluctuation. FIG. 4 shows a flowchart of steps of a method for adjusting ion implantation parameters based on an ion beam current reference value and a measured beam current. In step S10 an ion beam is generated. The ion beam can be generated in any one of several well-known ways in the art and can include any type of desired ion species at any desired energy. The ion beam typically includes substantially only one ion species, but the beam may include different ion species, if desired. In step S10, a reference value for the ion beam current is determined. The ion beam current is a measure of an amount of charge carried by particles in the beam per unit area or for a total cross-sectional area of the beam over a period of time. The reference value can be a measured beam current when the vacuum level within an ion implantation system is at a desired level. For example, the ion beam current may be measured before implantation of a wafer has begun, when the vacuum level within the ion implantation system and along an ion beam path is relatively high and stable. In step S30, a material, such as a semiconductor wafer, is implanted using the ion beam generated in step S10. Implantation of the material can be performed by directing the ion beam at a desired angle toward the semiconductor wafer such that energetic particles in the beam are implanted in the material. In step S40, the ion beam current is measured during implantation. The beam current can be measured using any desired beam current measuring device, such as a Faraday cup, a detector that uses calorimetry or beam-induced magnetic field measurements, or other detector. The ion beam current can be measured at a position adjacent the material being implanted or along the beam path to the material being implanted. In step S50, ion implantation parameters are adjusted based on the reference value for the ion beam current and the measured beam current during implantation. For example, the difference between the reference value and the measured current can be determined and ion implantation parameters can be adjusted based on the difference value. Various different ion implantation parameters can be adjusted based on the difference value. For example, the wafer scan rate can be adjusted, e.g., decreased, to accommodate for a decrease in dosing level resulting from vacuum fluctuations along the beam path during implantation. Other ion implantation parameters can be adjusted, such as the beam current, the beam scan rate or frequency, beam uniformity, the evacuation rate of a vacuum system used to control the vacuum level along the beamline, etc. The difference value between the reference value for the beam current and the measured beam current during implantation can be scaled to account for non-line of sight collisions that contribute to a decrease in detected beam current and a decrease in wafer dosing, and line of sight collisions that contribute to a decrease in detected beam current, but do not affect wafer dosing. Thus, the difference value can be scaled and implant parameters can be adjusted so that a desired dose is delivered to the semiconductor material. While the invention has been described in conjunction with specific embodiments thereof, it is evident that many alternatives, modifications and variations will be apparent to those skilled in the art. Accordingly, preferred embodiments of the invention as set forth herein are intended to be illustrative, not limiting. Various changes may be made without departing from the invention. |
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052727432 | claims | 1. A key member for use in insertion and removal of a fuel rod into and from a nuclear fuel assembly grid which includes a plurality of elongated straps intersecting generally perpendicular with each other to define a plurality of grid cells therein, and a plurality of pairs of dimples and springs formed on the straps for supporting a plurality of fuel rods such that each pair of dimple and spring are disposed in facing relation to each other, and which includes key member accommodating openings formed at the intersections of the straps so as to define peripheral ends therearound, said key member comprising an elongated key body having a pair of opposite surfaces, said key body including a plurality of first projections adapted to be held in engagement with said peripheral end of said opening of said strap and a plurality of second projections adapted to be held in engagement with said spring to deflect the spring away from the dimple facing thereto, said first projections and said second projections being formed on said opposite surfaces alternately in a staggered relation to one another. |
041773861 | summary | BACKGROUND OF THE INVENTION The present invention relates to long term or semipermanent storage of spent nuclear fuel assemblies, and more particularly to a storage rack for spent fuel assemblies and a method of storage of such fuel assemblies to maximize the capacity of the rack while preventing criticality. Nuclear reactors consist of an array of fuel rods containing the nuclear fuel. The fuel rods are metal tubes, typically from 8 to 15 feet in length and about 1/2 inch in diameter, and are supported in groups in fuel assemblies which may comprise a considerable number of rods. The large reactors utilized for power generation contain a large number of these fuel assemblies arranged in a suitable configuration. After an extended period of operation, the irradiated or spent fuel assemblies must be removed from the reactor and replaced. The spent fuel rods contain residual amounts of the original fuel material, and varying amounts of numerous fission products resulting from fission of the nuclei of the original fuel and other nuclear reactions, as well as from radioactive decay of initially-formed fission products. Certain of these materials are themselves fissionable. Many of the fission products are highly radioactive, at least initially, and thus produce considerable heat, while the entire fuel assembly is dangerously radioactive. The fuel rods can be reprocessed by chemically separating the fissionable material for reuse as fuel and recovering various other fission products, such as certain rare earth elements, for example, which have substantial commercial value. Suitable facilities must be provided for the storage of these highly radioactive fuel assemblies after removal from the reactor until they can be reprocessed or otherwise disposed of. Such storage presents serious problems since the fuel assemblies are initially highly radioactive and generate a great deal of heat. They must, therefore, be kept submerged in water which serves as a coolant to prevent overheating as well as a radiation shield and moderator for the fast neutrons which are still being emitted. It is also necessary to be sure that the assemblies are stored in a manner that will prevent criticality of the collection of fuel assemblies while keeping the space required to a minimum. After some period of time, the heat generated and the radioactivity of the fuel assemblies decline, since many of the fission products have relatively short halflives, and the nature of the storage problem changes as both the heat to be dissipated and the radiation hazard decrease. In our prior U.S. Pat. No. 4,010,375, there is disclosed a storage rack for spent nuclear fuel assemblies which is primarily intended for temporary storage of fuel assemblies in a water-filled pit. The rack consists of a checkerboard array of storage cells with the spent fuel assemblies placed in alternate cells. The intervening cells are filled with water which functions as a moderator and as a coolant, and also include a poison or neutron-absorbing material. This arrangement prevents criticality of the collection of fuel assemblies while maximizing the capacity of the rack for relatively short term storage. We have also proposed, in our copending application Ser. No. 851,038, filed Nov. 14, 1977, to use a similar rack structure for very long term or permanent storage, and to maximize the capacity by completely filling all the cells in the array after sufficient irradiation of the fuel and subsequent radioactive decay has occurred to make this safely possible. Concrete shielding may be provided for permanent storage. There is also a need, however, for relatively long term or semipermanent storage of spent nuclear fuel assemblies in a manner which will permit the storage of a maximum number of such assemblies in a given space, with complete safety, for a relatively long period until they can be disposed of by reprocessing, placing them in permanent storage, or otherwise. In the conventional design of spent fuel storage facilities, criticality is prevented by means of the spacing, or pitch, between adjacent fuel assemblies in the storage rack. There is a possibility that it may at some time be necessary to completely unload the reactor to make repairs or inspections inside the reactor pressure vessel, and it is, therefore, assumed that it may be necessary to place nearly fresh or unirradiated fuel assemblies in the storage rack during such repairs or inspection. In conventional designs, therefore, the pitch between adjacent fuel storage locations has been determined on the basis of the reactivity of fresh or unirradiated fuel assemblies. This, of course, requires a much larger pitch than is needed to prevent criticality with spent fuel assemblies after discharge from the reactor, and results in reduced storage capacity of a given space. Furthermore, once the pitch has been determined in a conventional design, it cannot be changed without rebuilding the storage rack. The arrangement and dimensions of such a rack are such that additional spent fuel assemblies cannot be placed between the initial storage positions both because the usual designs do not provide for accommodating such additional fuel assemblies and because the dimensions are generally too small. SUMMARY OF THE INVENTION In accordance with the present invention, a maximum density rack structure and method of storage are provided for long term or semipermanent storage of spent nuclear fuel assemblies which provides maximum capacity and complete safety. The problems discussed above of conventional storage rack designs are overcome by providing the ability to change the pitch between adjacent fuel assemblies when the actual reactivity level of the spent fuel assemblies is known. Packing or storage densities as high as 90% can thus be attained. These results are obtained by means of a storage rack consisting of a plurality of identical storage cells disposed in a regular array. The cells are of proper size to contain one or more of the fuel assemblies to be stored and are preferably made of stainless steel, which is capable of absorbing neutrons. The rack is intended to be immersed in water which serves as both a coolant and a moderator for the fuel assemblies to be stored in the rack. Cap members are placed on alternate cells in each row with the caps in adjoining rows staggered with respect to each other, the intervening cells being left open, and certain of the caps being removable. The cap members determine which cells are available for storage by covering the others, and also serve as lead-in guides to guide the fuel assemblies into the open cells. Initially, the rack is filled by placing fuel assemblies in the cells that are left open, so that about 50% of the available storage area is filled in this way. This may take a substantial period of time, such as several years for a rack at a large power generating site. The rack may then be surveyed and the reactivity level of the fuel assemblies stored in the rack assessed. This survey may be based on the known operating history of the fuel assemblies or on actual measurements. On the basis of such a survey, or on the basis of the known and calculated characteristics of the fuel, rate of burn-up, time of exposure in the reactor and other known factors, a final storage configuration is determined which will prevent criticality. The removable caps are removed from the cells which they previously covered and rearranged to open additional storage cells in the pattern thus determined. These additional storage cells are then filled with fuel assemblies, the remaining cells still being filled with water. The final result is a rack which contains fuel assemblies in a carefully determined pattern such that unfilled cells containing water are distributed in the proper positions to act as moderators and neutron-absorbing elements to prevent criticality. In this way, the final storage pattern, or pitch between fuel assemblies, is determined by the reduced reactivity of the spent fuel assemblies as they are actually discharged from the reactor. The capacity or packing density of the rack is thus maximized, and in some cases as much as 90% of the available storage area can be filled with complete safety. |
claims | 1. A method for interlocking a lens assembly power supply in ion implantation equipment comprising:relaying a control signal to a positive and a negative power supply; applying a positive voltage generated by the positive power supply and a negative voltage generated by the negative power supply to a lens assembly;detecting a positive voltage and a negative voltage from the positive and negative power supplies respectively;comparing a first relative value of the positive voltage and a second relative value of the negative voltage;changing a mode of the positive and negative power supplies; andgenerating an interlocking signal which interlocks the positive and negative power supplies when a sum of the first relative value and the second relative value does not equal zero. 2. The method according to claim 1, wherein the ion implantation equipment operates in a normal manner when the sum of the first and second relative values equals zero. 3. The method according to claim 1, wherein when the sum of the first and second relative values is not zero, the control signal is interrupted and the positive and the negative power supplies are operated manually. 4. The method according to claim 1, wherein changing a mode of the positive and negative power supplies and generating the interlocking signal occur at substantially the same time. 5. The method according to claim 1, wherein the positive and negative power supplies operate in a remote mode when the sum of the first and second relative values equals zero. 6. A method of generating an interlocking signal comprising:detecting a positive voltage from a positive power supply and a negative voltage from a negative power supply;comparing a first relative value of the positive voltage and a second relative value of the negative voltage; andgenerating an interlocking signal when a sum of the first relative value and the second relative value does not equal zero. 7. The method of claim 6, wherein the interlocking signal interlocks a power supply in an ion implantation device. 8. A method for interrupting an ion implantation process comprising:detecting a positive voltage from a positive power supply and a negative voltage from a negative power supply;comparing a first relative value of the positive voltage and a second relative value of the negative voltage;changing a mode of the positive and negative power supplies; andgenerating an interlocking signal which interrupts the ion implantation process when a sum of the first relative value and the second relative value does not equal zero. 9. The method of claim 8, wherein changing a mode of the positive and negative power supplies and generating the interlocking signal occur at substantially the same time. |
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059149981 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS To better understand the present invention, brief reference will be made to a conventional device for condensing X-ray beam, shown in FIG. 1. As shown, the device, generally 10, includes an X-ray generator or X-ray source for emitting X-rays 14. The X-rays 14 issuing from the X-ray generator 12 are condensed by an X-ray Fresnel zone plate 16 to a focus or virtual light source 18. The X-ray Fresnel zone plate 16 is a Fresnel zone plate originally established for visible rays and applied to X-rays. FIG. 2 shows another conventional X-ray beam condensing device. As shown, the device, generally 10A, includes a mirror 20 for totally reflecting X-rays in place of the Fresnel zone plate 16. This device is based on the fact that because the X-rays 14 have a refractive index smaller than 1, they are totally reflected when incident to the surface of the mirror 20 at an angle less than a critical angle. FIG. 3 shows still another conventional X-ray beam condensing device. As shown, the device, generally 10B, spatially reduces the sectional area of the X-ray beam 14 by using a pin hole or a slit 22. The conventional device shown in FIGS. 1-3 have some problems left unsolved, as discussed earlier. Basically, in accordance with the present invention, X-rays are condensed to form a microbeam. Then, a part of the microbeam which can be considered to be a plane wave is separated. Specifically, a plane wave X-ray microbeam generating device in accordance with the present invention includes an X-ray generator or X-ray source and a condensing element. A simultaneous reflection Borrmann element is located at the focus of the condensing element. X-rays issuing from the X-ray generator have their divergence angle restricted by the Borrmann element. The X-ray generator may be implemented by synchrotron radiation or an X-ray tube. In a diffraction condition wherein divergence planes defined by the direction of incident X-rays and that of diffracted X-rays are perpendicular to each other, angular divergence in the direction contained in the divergence planes can be restricted to the order of seconds. When the divergence angle is restricted by such dynamical diffraction, not only a wave diffracted in the direction of reflection but also a wave diffracted in the direction of transmission can be restricted in divergence angle. FIG. 4 shows Laue-case diffraction. As shown, assume that a single crystal of silicon 24 has a sufficient thickness. Then, Laue-case diffraction increases the X-ray beam transmittance in the transmission direction, compared to a case without diffraction, and further restricts the angular divergence. Such an anomalous transmission phenomenon is referred to as the Borrmann effect. When a plurality of lattice planes joining in diffraction are present, there appear a wave in the transmission direction and the same number of waves as the lattice planes in the reflection direction (simultaneous reflection or multiple-beam diffraction). The simultaneous reflection refers to a condition wherein when diffraction satisfying the Bragg condition occurs for a certain lattice plane (h, k, l), it also satisfies the Bragg condition for another lattice plane (m, n, o) at the same time. FIGS. 5A and 5B demonstrate simultaneous reflection to which a plurality of lattice planes are related. FIGS. 5A and 5B are sections perpendicular to each other; FIG. 5B is a section as seen in the direction of an arrow B shown in FIG. 5A. While lattice planes and the direction of diffracted X-rays indicated by broken lines are representative of diffraction incidentally allowable due to the symmetry of a single crystal of silicon 26, they are not relevant to the present invention. Because the two diffraction planes are perpendicular to each other, the X-ray beam in the transmission direction has its divergence angle restricted in the direction contained in the individual scattering plane by diffraction. As a result, an X-ray beam restricted in the two different directions is achievable. A slit is positioned after the Borrmann diffraction element. A part of the X-rays transmitted and diffracted by an optical element, i.e., satisfied the diffraction conditions is selectively produced at the outlet side of the above slit. This successfully generates a plane wave X-ray microbeam. The prerequisite with the above arrangement is that the X-ray generator, condensing element, simultaneous reflection Borrmann element and slit be sequentially arranged in this order. Should the condensing element be positioned after the Borrmann element, the divergence angle would increase and would prevent an X-ray beam having a small beam size and a small divergence angle from being achieved. Referring to FIG. 6, an X-ray microbeam generating device embodying the present invention will be described. As shown, the microbeam generating device, generally 30, includes an X-ray generator 32 capable of emitting X-rays having a size of 3 mm square, a divergence angle of 4 mrad, and a number of photons of 10.sup.-9 /sec. A condensing element is implemented by a Fresnel zone plate 34. A simultaneous reflection Borrmann element 46 has a single crystal of silicon which is 2 mm thick (1.4 mm or above) and has a (001) plane. A 1 mm to 5 mm tantalum plate 38 is spaced from the diffraction element 36 by about 5 cm and formed with an aperture having a diameter of 5 mm. If desired, the Fresnel zone plate 34 may be replaced with a mirror totally reflecting X-rays or a Bragg Fresnel lens which is a reflection type Fresnel lens. In the above device 30, X-rays issuing from the X-ray generator 32 is spatially restricted by the Fresnel zone plate 34 to turn out an X-ray beam. The X-ray beam has its divergence angle restricted by the Borrmann element 36 located at the focus of the Fresnel zone plate 34 (focal distance of 1 m). As a result, a plane wave X-ray microbeam is generated. Subsequently, the diffraction element 36 causes 333, 333, 333 and 333 reflections to occur at the same time for the X-ray with the wavelength of 0.12 nm. Waves diffracted by 70 degrees with respect to the incidence direction are excluded by a slit 38a formed in the tantalum plate 38, so that only a wave diffracted in the transmission direction is separated. Experiments showed that the transmitted wave had a divergence angle of 1 second to 2 seconds and a beam diameter of up to about 10 .mu.m. The illustrative embodiment is not limited to the above parameters, but allows any suitable lattice planes matching with a wavelength to be selected. For example, when X-rays having a wavelength of 0.36 nm may be incident perpendicularly to a silicon (001) plane in order to cause 111, 111, 111 and 111 reflections to occur at the same time. Likewise, for 0.0-72 nm or 0.052 nm X-rays, use may be made of 555, 555, 555 and 555 reflections or 777, 777, 777 and 777 reflections. Further, silicon playing the role of a diffracting element may be replaced with, e.g., germanium or crystal so as to change the distance between lattice planes. Such an alternative crystal is adaptive to another wavelength. Assume that the slit 38a of the tantalum plate 38 is replaced with a pin hole. Then, the pin hole is located at a position where the X-rays are incident to the diffraction element, because the size of the X-ray beam is minimum at the pin hole. For this purpose, metal or the like is deposited on the incidence surface of the silicone crystal of the diffraction element 36, FIG. 6, and a pin hole (up to 1 .mu.m) is formed at the incidence point by a laser. With this configuration, it is also possible to generate a plane wave X-ray microbeam. So long as the silicon crystal has a sufficient thickness, the planeness of the wave is not effected due to the Borrmann effect although the intensity of the output beam is reduced. As stated above, the illustrative embodiment is capable of generating an X-ray microbeam having a restricted divergence angle and desirable planeness in regions other than the focus. This realizes the use of a plane wave X-ray microbeam having a sufficiently small spatial spread. Consequently, limitations heretofore posed on the work region due to the focus and on the work distance are obviated, so that the fine structure of a substance can be easily analyzed by, e.g., X-ray analysis. Reference will be made to FIG. 7 for describing an alternative embodiment of the present invention. As shown, in this embodiment, the size of the X-ray beam is reduced by asymmetrical reflection using a reflection plane not parallel to a crystal surface 42, i.e., a lattice plane 44. A crystal 40 is rotated about an axis 46 perpendicular to the lattice plane 44 in order to vary the incident angle and exit angle from the crystal surface 42. This allows the asymmetric factor, i.e., the degree of asymmetry ascriable to a change in the energy of X-rays to remain constant and thereby implements X-ray energy scanning without effecting the condensing efficiency. Assuming that the asymmetry factor is b, then b is expressed in terms of an angle .theta..sub.o between the crystal surface 42 and the input X-rays and an angle .theta..sub.G between the surface 42 and the output X-rays, as follows: EQU b=sin .theta..sub.o /sin .theta..sub.G Eq.(1) By diffraction with the above degree of asymmetry, the spatial spread of the input X-rays in the scattering plane is increased by 1/b times in terms of output X-rays, while the angular divergence is increased by b times. Assuming a Bragg angle .theta..sub.B and an angle .alpha. between the lattice plane 44 relating to the diffraction and the crystal surface 42, then the degree of asymmetry b is produced by: EQU b=sin (.theta..sub.B +.alpha.)/sin (.theta..sub.B -.alpha.)Eq.(2) where .alpha. may range from -.theta..sub.B to .theta..sub.B. If a plurality of crystals are used to effect sequential reflection, then the beam size can be further reduced. In the asymmetric reflection, by rotating the crystal 40 about the axis 46 perpendicular to the lattice plane 44, it is possible to vary the angles of the input X-rays and output X-rays to the crystal surface. Consequently, in the range of rotation of from 0 degree to 180 degrees, the asymmetry factor can be varied from b up to 1/b, including b=1 holding when the angle of rotation is 90 degrees (.alpha.=0). The rotation of the crystal 40 therefore compensates for a change in the wavelength (or energy) of the input X-rays and therefor a change in the degree of asymmetry, i.e., Bragg angle, thereby maintaining the degree of asymmetry constant. Further, any desired condensing conditions or values are selectable on the basis of the degree of asymmetry b, so that the beam size can be varied. FIGS. 8A and 8B show another alternative embodiment of the present invention. Briefly, this embodiment sequentially uses perpendicular scattering planes for reflection in order to reduce the beam size. In addition, the embodiment reduces the angular width relating to the diffraction of incident X-rays to the order of seconds, thereby generating an X-ray beam having a restricted angular width. As shown in FIGS. 8A and 8B, an X-ray beam 52 issuing from an X-ray generator 50 has its beam size restricted by a single crystal of silicon 54 effecting asymmetrical Bragg reflection. The X-ray generator 50 is implemented by a rotary anode type X-ray generator; the beam size is 1 mm.times.1 mm. For 0.05 nm X-rays, the Bragg angle for 422 reflection is 13.0 degrees. When the crystal 54 is cut such that the angel between the (422) plane and the crystal surface is 12.0 degrees, the degree of asymmetry b is 24.3 The X-rays diffracted by the crystal 54 are further diffracted by a similar crystal 56, so that the beam size can be further reduced to about 10 .mu.m, as determined by experiments. Crystals 58 and 60 are arranged to define a scattering plane perpendicular to the scattering plane of the crystals 54 and 56. As a result, the beam size is reduced to about 10 .mu.m in both the horizontal direction and the vertical direction, as also determined by experiments. The angular divergence of the diffracted X-rays was found to be about 10 seconds. Then, the crystals 54-60 are so rotated as to output X-rays whose wavelength is 0.15 nm. In this case, the Bragg angle and the asymmetry factor are 42.6 degrees and 57.0, respectively. Experiments showed that under the above conditions the condensing conditions noticeable changed and implemented a beam size of about 5 .mu.m. It was found by experiments that when the axis 46 of the individual crystal was rotated to implement an angle of 2.3 degrees between the output X-rays and the crystal surface and a degree of asymmetry of about 2.4, the beam size remained to be about 10 .mu.m despite a change in wavelength. Further, by varying the angle between the output X-rays and the crystal surface, it was possible to vary the beam size steplessly from 10 .mu.m to several centimeters. As stated above, in the embodiments shown in FIGS. 7, 8A and 8B, the energy of an X-ray beam having a small diameter can be scanned over a broad range without effecting condensing conditions. This allows EXAFS (Extended X-ray Absorption Fine Structure) or similar experiment to be easily executed with a small beam size. Moreover, the beam size is freely variable via the condensing conditions in order to execute the local strain analysis of a sample or the analysis of a fine structure. Specifically, it is possible to compensate for a change in the degree of asymmetry ascriable to a change in the wavelength of X-rays selected, and therefore to maintain the degree of asymmetry constant. In addition, the condensing conditions including the energy of X-rays and beam size each can be set independently of the others. Various modifications will become possible for those skilled in the art after receiving the teachings of the present disclosure without departing from the scope thereof. |
summary | ||
054266777 | abstract | A pressurizer vessel of a nuclear power plant contains a liquid and steam both of which function to maintain pressure in a reactor coolant system. A heater support assembly is disposed in an interior portion of the pressurizer vessel and receive a plurality of heaters which are matingly fitted with the heater support assembly for heating the liquid. A temperature detector is operatively connected to the heater support assembly in a structural arrangement which allows for measuring the temperature of the liquid at a plurality of preselected elevations. The temperature detector further includes temperature measuring means for measuring a plurality of temperature readings of the liquid at preselected elevations of the liquid. |
description | This application is a national phase of PCT Application No. PCT/EP2018/055374 Filed Mar. 5, 2018, which in turn claims the benefit of German Patent Application No. 10 2017 107 584.4 filed Apr. 7, 2017. The present invention relates to a decontamination solution that contains zinc for decontaminating light-water reactors, and to a method for decontaminating radioactive metal surfaces using the decontamination solution. In the field of nuclear reactor technology, metal components are radioactively contaminated. Such contamination routinely occurs during normal operation of reactors and relates in particular to metal components located in the primary circuit, for example of a pressurized water reactor. In this case, radioactive substances are deposited in the oxide layers formed on the surface of the components, causing them to become radioactively contaminated. In the event of an inspection of the nuclear power station, it is routinely necessary to free the contaminated components from the radioactivity, i.e. from the deposits on the metal surface, in order to protect the inspection personnel from radiation. The components can then continue to be processed in the nuclear power station. The same applies if the nuclear power station is intended to be demolished. In principle, mechanical means can be used to remove such deposits, wherein the oxide layers and therefore the contaminated regions are ground, for example. This is in particular disadvantageous for components that are difficult for the grinding tool to access due to their dimensions or their position. Furthermore, it is known to decontaminate the components using a decontamination solution that comprises a complexing agent, which includes various carboxylic acids such as oxalic acid. In this case, the portions of the oxide layers of low solubility are first oxidized or reduced in a preceding step, with permanganates (potassium permanganate, permanganic acid) being used to oxidize Cr-III to Cr-VI, for example. The oxide layer, which mainly consists of iron and nickel ions, is then dissolved with the aid of the complexing agent and the released cations, which also include 60Co2+ or 58Co2+, are removed from the decontamination solution by means of ion-exchange. This decontamination process is usually carried out in several rounds, the oxide layer being broken down bit by bit. In addition to these radioactive isotopes, inactive ions are also always released into the decontamination solution, which are likewise removed from the decontamination solution by means of the ion-exchange resins. Furthermore, recontamination of the components takes place as early as during the decontamination process as a result of the radioactive ions present in the decontamination solution. As a result, the efficiency of the decontamination process is reduced, leading to a larger number of decontamination cycles being required that are time-consuming and expensive, and also leading to a greater amount of contaminated ion-exchange resins that need to be disposed of, which requires an enormous amount of effort. Of course, the above-mentioned problems do not only occur in nuclear power stations, but in principle in situations in which metal components come into contact with radioactivity and in which decontamination is required. Accordingly, there is the need for an improved method for decontaminating radioactively contaminated metal surfaces. In particular, there is the need for a more efficient decontamination method, in which decontamination can be carried out by means of a smaller number of decontamination cycles and/or a smaller amount of contaminated ion-exchange resins. This object is achieved according to the invention by a method having the features specified in claim 1. Embodiments are specified in the dependent claims. More precisely, the method according to the invention is a method for decontaminating a radioactively contaminated metal surface, which comprises the step of bringing at least a portion of the radioactively contaminated metal surface into contact with a decontamination solution comprising a complexing agent and a transition metal. As could surprisingly be shown, when a transition metal is added to the decontamination solution, recontamination of the metal surface, which occurs during the decontamination process, is effectively reduced. Without being limited hereto, it is assumed that the transition metal added to the decontamination solution competes with the radioactive isotopes released for (re-)incorporation in the metal surface (or the oxide layer thereon). As a result, a larger amount of radioactive isotopes can advantageously be removed from the decontamination solution by means of the ion-exchange process, which in turn reduces the number of rounds of the decontamination steps required and/or reduces the amount of ion-exchange resins to be disposed of. The decontamination solution is preferably an aqueous solution. The transition metal is preferably an ion of the transition metal, more preferably a cation of the transition metal, even more preferably a bivalent or trivalent cation of the transition metal. Most preferably, the transition metal is a bivalent cation of the transition metal. The transition metal is more preferably a depleted transition metal, i.e. a transition metal having a reduced proportion of isotopes compared with the proportion of isotopes that naturally occurs, which isotopes can be easily activated by neutrons. The use of a depleted transition metal is particularly advantageous when the metal to be decontaminated, for example the component of a reactor, is not disposed of after decontamination, but is recycled and intended to be subjected to neutron flux. The transition metal is likewise preferably selected from the group consisting of zinc, nickel, cobalt or mixtures thereof. More preferably, the transition metal is selected from the group consisting of zinc and nickel. Most preferably, the transition metal is zinc. The use of zinc in the decontamination solution surprisingly showed the greatest effect when reducing the extent to which the metal surface is recontaminated, as per the invention. The transition metal is preferably present in the decontamination solution in a concentration in the range of from ≥0.5 mg/kg to ≤15 mg/kg, more preferably from ≥0.5 mg/kg to ≤10 mg/kg, more preferably from ≥1.5 mg/kg to ≤5 mg/kg or from ≥2 mg/kg to ≤5 mg/kg, and most preferably from approximately 3 mg/kg to ≤4 mg/kg. Instead of mg/kg, mmol/L can also be stated, the stated mg/kg value having to be divided by the atomic weight of the particular transition metal. The transition metal is preferably present in the decontamination solution in a concentration in the range of from ≥7 μmol/L to ≤230 μmol/L, more preferably from ≥7 μmol/L to ≤155 μmol/L, more preferably from ≥23 μmol/L to ≤70 μmol/L or from ≥30 μmol/L to ≤80 μmol/L, and most preferably from approximately ≥46 μmol/L to ≤62 μmol/L. The concentration ranges specified preferably hold true for the concentration of the transition metals when the metal surface is brought into contact with the decontamination solution according to the invention. The concentrations specified are likewise preferably the average concentrations. In the following, reference will be made to the element zinc purely by way of example instead of to “transition metals”. If applicable, the comments made also apply in a similar way to transition metals in general, and preferably also to nickel and/or cobalt. The term “zinc” is preferably intended to be understood to mean the zinc ions present in the decontamination solution, more preferably Zn2+. More preferably, this can be depleted zinc, in particular zinc depleted in 64Zn. The zinc is more preferably introduced into the decontamination solution by means of a soluble zinc compound. Preferred soluble zinc compounds are selected from the groups of the acids used and/or the complexing agents used with zinc, comprising zinc methanesulfonate (Zn(CH3SO3)2), zinc nitrate (Zn(NO3)2), zinc permanganate (Zn(MnO4)2), zinc sulfate (ZnSO4) and/or a soluble zinc complex. The zinc complex is more preferably a complex of zinc and the complexing agent used. The term “decontamination” is known to a person skilled in the art. This is intended to be understood in particular to mean the reduction and/or removal of radioactivity present on the metal surface. In particular, this is intended to be understood to mean the removal of a layer of metal oxide deposits on a metal component, the deposited layer comprising radioactive isotopes, preferably cobalt. In other words, by means of the method according to the invention, radioactive isotopes are removed from the metal surface to be decontaminated. These radioactive isotopes are preferably selected from the group consisting of 55Fe ions, 63Ni ions, 54Mn ions, 65Zn ions, 125Sb ions, 137Cs ions, 58Co ions and 60Co ions. The radioactive isotopes are more preferably selected from the group consisting of 54Mn ions, 125Sb ions, 137Cs ions, 58Co ions and 60Co ions. These radioactive isotopes are most preferably 58Co ions and/or 60Co ions, even more preferably 60Co ions. The decontamination method of the present invention can preferably also be referred to as chemical decontamination. More preferably, the decontamination method can be a method for decontaminating a nuclear reactor that is to be demolished or a nuclear reactor that shall continue to be operated. The clearance of solid and liquid substances is regulated according to the Radiological Protection Ordinance (RPO, Strahlenschutzverordnung StrlSchV) and is substantially divided into unrestricted clearance and clearance for disposal on landfills. Following decontamination of the metal surface, what is preferably left is a component that is cleared for disposal on landfills. Following decontamination of the metal surface, what is even more preferably left is a component that is suitable for unrestricted clearance. Hereinafter, the term “radioactively contaminated metal surface” is intended to preferably be understood to mean the surface of a metal component including the radioactively contaminated deposited layer located thereon, which forms during normal use of the component in a pressurized water reactor, for example. Such a deposited layer preferably consists of metal oxides of low solubility. In other words, the radioactive metal surface to be decontaminated preferably comprises at least one radioactively contaminated layer of metal oxides of low solubility, which layer is arranged on the surface and is made of basic metal material. More preferably, the deposited layer is spinels, preferably Cr—Ni spinels and/or Cr—Fe spinels. Spinels are minerals of low solubility from the mineral class of oxides and hydroxides, which are usually present in crystal form, and are preferably oxides having the amount-of-substance ratio of metal:oxygen=3:4. The metal of the metal surface to be decontaminated can in principle be any suitable metal. The metal is preferably a metal selected from the group consisting of iron, nickel, chromium, manganese, titanium, niobium, copper, cobalt and combinations of at least two of these metals. The metal is more preferably selected from the group consisting of iron, chromium, nickel, cobalt and combinations of at least two of these metals. According to the invention, at least a portion of the metal surface is also brought into contact with the decontamination solution. Preferably a plurality of portions, and more preferably the entire metal surface, is/are brought into contact with the decontamination solution. For better understanding, reference will be made to the radioactively contaminated metal surface in the following, even though a portion of said surface is also always meant thereby. The radioactively contaminated metal surface can be brought into contact with the decontamination solution, as per the invention, in any suitable manner. The metal surface to be decontaminated is preferably wetted with the decontamination solution. The decontamination solution is more preferably introduced into the primary circuit of a reactor. More preferably, the decontamination solution can be circulated. As a result, concentration gradients in the region of the metal surface can advantageously be avoided and the efficiency of the decontamination process can be increased. More preferably, circulation is continuous and is likewise preferably carried out using pumps. Likewise preferably, the metal surface to be decontaminated is the inner lateral face of a metal and cylindrical component (such as a tube of a recuperator) and the decontamination solution is introduced into the cavity of the cylindrical component. Before the method step of bringing the at least one portion of the metal surface into contact with the decontamination solution according to the invention, the method according to the invention preferably comprises an additional method step, i.e. as the first method step, for oxidizing or reducing the radioactively contaminated metal surface. In the case of oxidation, this method step can also be referred to as the pre-oxidation of the radioactively contaminated metal surface. More preferably, during pre-oxidation, Cr-III is oxidized to Cr-VI. Pre-oxidation is preferably carried out by bringing the radioactively contaminated metal surface into contact with nitric acid and potassium permanganate, with sodium hydroxide and potassium permanganate, a vanadium compound (preferably vanadium formate) or with permanganic acid, the permanganic acid treatment being the most preferable. In the case of a preceding reduction method step, the oxidation layer is preferably reduced by means of a vanadium compound. In the method step that follows this step, the dissolved products are preferably complexed with picolinic acid. After the pre-oxidation step and before the at least one portion of the metal surface is brought into contact with the decontamination solution according to the invention, an additional method step can more preferably be carried out for reducing the excess oxidizing agent, for example the permanganate (potassium permanganate, permanganic acid). After bringing the at least one portion of the metal surface into contact with the decontamination solution according to the invention, the method according to the invention likewise preferably comprises the additional method step of removing at least some of the radioactive isotopes, or ions thereof, present in the decontamination solution. These radioactive isotopes are preferably selected from the group consisting of 55Fe, 63Ni, 54Mn, 65Zn, 125Sb, 137Cs, 58Co and 60Co. More preferably, the radioactive isotopes are selected from the group consisting of 54Mn, 125Sb, 137Cs, 58Co and 60Co. These radioactive isotopes are most preferably 58Co and/or 60Co, more preferably 60Co. The radioactive isotopes are preferably removed by means of binding to an ion-exchange resin, more preferably a cation-exchange resin and/or a synthetic ion-exchange resin. Most preferably, the ion-exchange is a strongly acidic cation-exchange, in which protons are exchanged for the bound cations. Such ion-exchange resins are well-known to a person skilled in the art. More preferably, approximately ≥50%, even more preferably approximately ≥70%, ≥80%, ≥90% or ≥99% of the radioactive isotopes present in the decontamination solution are removed. Most preferably, approximately ≥99% and <100% of the isotopes present in the decontamination solution are removed. More preferably, the method according to the invention is cyclic. In other words, at least the method steps of bringing the metal surface into contact with the decontamination solution according to the invention and subsequently removing at least some of the radioactive isotopes present in the decontamination solution are repeated at least once. Of course, individual method steps or all the additional method steps mentioned above can also additionally be repeated in this case. The method according to the invention is preferably repeated until a decontamination factor has been reached that corresponds to a reduction in the activity of the radioactively contaminated metal surface from ≥1 to ≤3 orders, more preferably approximately 2 orders. The decontamination factor is preferably determined by measuring the activity of the ion-exchange resin used to remove the radioactive isotopes present in the decontamination solution, or by comparing the activity of the ion-exchange resin before and after carrying out the method according to the invention. The method according to the invention is likewise preferably repeated in cycles from 1 to 30 times, more preferably from 10 to 25 times, even more preferably from 13 to 20 times. A range of from 13 to 17 cycles showed particularly good results when using oxalic acid. According to the invention, in addition to the transition metal, the decontamination solution comprises at least one complexing agent. The complexing agent can also be referred to as a chelating agent. Together with metal ions, complexing agents form chelate complexes. Examples of complexing agents include acids, such as nitrilotriacetic acid, ethylenediaminetetraacetic acid, fluoric acid, phosphoric acid, oxalic acid, tartaric acid, citric acid and salts thereof. The complexing agent is particularly preferably an acid. The decontamination solution also comprises water, as a result of which the aqueous components of the decontamination solution can be in their dissolved form. In other words, the decontamination solution is an aqueous solution. The acid is preferably selected from the group consisting of carboxylic acid, methane sulfonic acid, oxalic acid, picolinic acid, nitric acid and citric acid. More preferably, the acid is a mixture of methane sulfonic acid and oxalic acid. The acid is most preferably oxalic acid. More preferably, the decontamination solution also contains an oxidizing agent, preferably permanganic acid, or a reducing agent. In other preferred embodiments, the decontamination solution contains zinc methane sulfonate, zinc nitrate, zinc permanganate, zinc sulfate and/or a zinc complex of the complexing agent used. The complex consisting of the transition metal and the complexing agent used is particularly preferred. The use of the decontamination solution according to the invention to carry out the method according to the invention is likewise a component of this invention. Decontaminations of the primary circuit of a light-water reactor were carried out, whereby the average Zn and Fe concentration in the decontamination medium and the 60Co removed from the decontamination solution in this case by means of the ion-exchange resin (strongly acidic cation-exchange) was determined. The primary circuit decontaminations were carried out over 15 cycles. As can be seen in FIGS. 1 and 2 (determination of the 60Co decontamination on the basis of the Zn concentration), there is a very good correlation between this transition metal and the amount of 60Co removed. In comparison thereto, it was not possible to demonstrate a very good correlation of this type between the Fe concentration and 60Co (see FIG. 3). Example 1 was repeated, whereby the Ni concentration or the Cr concentration was observed instead of the Zn concentration. In this case, a correlation was likewise shown between the concentration of the transition metal and the activity removed by means of 60Co in each case. The correlation determined decreased tendentially, and in comparison with Zn, from Ni by means of Cr. |
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description | This application is a divisional application of U.S. patent application Ser. No. 13/744,440, filed Jan. 18, 2013, entitled “Passive System for Cooling the Core of a Nuclear Reactor, which application is related to U.S. patent application Ser. No. 13/495,083, filed Jun. 13, 2012, entitled “Small Modular Reactor Safety Systems.” 1. Field This invention pertains generally to nuclear reactor safety systems, and more particularly, to a system for passively cooling the core of a nuclear reactor and a spent fuel pool during a refueling outage in the event of a nuclear station blackout. 2. Description of Related Art A pressurized water reactor has a large number of elongated fuel assemblies mounted within an upright reactor vessel. Pressurized coolant is circulated through the fuel assemblies to absorb heat generated by nuclear reactions in fissionable material contained in the fuel assemblies. The primary side of such a nuclear reactor power generating system which is cooled with water under pressure comprises a closed circuit which is isolated from and in heat exchange relationship with a secondary circuit for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internal structure that supports the plurality of fuel assemblies containing the fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. In conventional nuclear plants of that type each of the parts of the primary side comprising the steam generator, a pump and a system of pipes which are connected to the reactor vessel form a loop of the primary side. For the purpose of illustration, FIG. 1 shows a simplified conventional nuclear reactor primary system, including a generally cylindrical pressure vessel 10 having a closure head 12 enclosing a nuclear core 14. A liquid coolant, such as water or borated water, is pumped into the vessel 10 by pump 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18, typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump 16, completing the primary loop. Typically, a plurality of the above-described loops are connected to a single reactor vessel 10 by reactor coolant piping 20. An exemplary conventional reactor design is shown in more detail in FIG. 2. In addition to the core 14 comprised of a plurality of parallel, vertically co-extending fuel assemblies 22, for the purpose of this description, the vessel internal structures can be divided into the lower internals 24 and the upper internals 26. In conventional designs, the lower internals function to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity in FIG. 2), and support and guide instrumentation and components, such as control rods 28. In the exemplary reactor shown in FIG. 2, coolant enters the reactor vessel through one or more inlet nozzles 30, flows down through an annulus between the reactor vessel and the core barrel 32, is turned 180° in a lower plenum 34, passes upwardly through a lower support plate 37 and a lower core plate 36 upon which the fuel assemblies are seated and through and about the fuel assemblies 22. In some designs, the lower support plate 37 and the lower core plate 36 are replaced by a single structure, a lower core support plate having the same elevation as 37. The coolant flow through the core and surrounding area 38 is typically large on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate 40. Coolant exiting the core 14 flows along the underside of the upper core plate and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially to one or more outlet nozzles 44. The upper internals 26 can be supported from the vessel or the vessel head and include an upper support assembly 46. Loads are transmitted between the upper support assembly 46 and the upper core plate 40 primarily by a plurality of support columns 48. Each support column is aligned above a selected fuel assembly 22 and perforations 42 in the upper core plate 40. Rectilinearly moveable control rods 28 which typically include a drive shaft or drive rod 50 and a spider assembly 52 of neutron poison rods, are guided through the upper internals 26 and into aligned fuel assemblies 22 by control rod guide tubes 54. The guide tubes are fixedly joined to the upper support assembly 46 and the top of the upper core plate 40. The support column 48 arrangement assists in retarding guide tube deformation under accident conditions which could detrimentally affect control rod insertion capability. To control the fission process, a number of control rods 28 are reciprocally moveable in guide thimbles at predetermined positions in the fuel assemblies 22. Specifically, a control rod mechanism positioned above the top nozzle of the fuel assemblies supports a plurality of control rods. The control rod mechanism (also known as a rod cluster control assembly) has an internally threaded cylindrical hub member with a plurality of radially extending flukes or arms that form the spider 52 previously noted with regard to FIG. 2. Each arm is interconnected to a control rod 28 such that the control rod assembly mechanism 72 is operable to move the control rods 28 vertically within the guide thimbles within the fuel assemblies to thereby control the fission process in the fuel assembly 22, under the motive power of the control rod drive shaft 50 which is coupled to the control rod mechanism hub, all in a well-known manner. The upper internals 26 also have a number of in-core instrumentation that extend through axial passages within the support columns 48 and into instrumentation thimbles generally, centrally located within the fuel assemblies. The in-core instrumentation typically includes a thermocouple for measuring the coolant core exit temperature and axially disposed neutron detectors for monitoring the axial and radial profile of the neutron activity within the core. Nuclear power plants, which employ light water reactors require periodic outages for refueling of the reactor. New fuel assemblies are delivered to the plant and temporarily stored in a fuel storage building in a spent fuel pool, along with used fuel assemblies which may have been previously removed from the reactor. During a refueling outage, a portion of the fuel assemblies in the reactor are removed from the reactor to the fuel storage building. A second portion of the fuel assemblies are moved from one support location in the reactor to another support location in the reactor. New fuel assemblies are moved from the fuel storage building into the reactor to replace those fuel assemblies which were removed. These movements are done in accordance with a detailed sequence plan so that each fuel assembly is placed in a specific location in accordance with an overall refueling plan prepared by the reactor core designer. In conventional reactors, the removal of the reactor internal components necessary to access the fuel and the movement of the new and old fuel between the reactor and the spent fuel pool in the fuel storage building is performed under water to shield the plant maintenance personnel. This is accomplished by raising the water level in the refueling cavity and canal that is integral to the plant building structure. The water level of more than 20 feet provides shielding for the movement of the reactor internal structures and the fuel assemblies. A typical pressurized water reactor needs to be refueled every 18 to 24 months. Commercial power plants employing the conventional designs generally illustrated in FIGS. 1 and 2 are typically on the order of 1,100 megawatts or more. More recently, Westinghouse Electric Company LLC has proposed a small modular reactor in the 200 megawatt class. The small modular reactor is an integral pressurized water reactor with all primary loop components located inside the reactor vessel. The reactor vessel is surrounded by a compact, high pressure containment. Due to both limited space within the containment and the lower cost requirement for integral pressurized light water reactors, the overall number of auxiliary systems including those associated with refueling needs to be minimized without compromising safety or functionality. For that reason, it is desirable to maintain most of the components in fluid communication with the primary loop of the reactor system within the compact, high pressure containment. Typical conventional pressurized water reactor designs make use of active safety systems that rely on emergency AC power after an accident to power pumps required to cool down the reactor and spent fuel pool. Advanced designs, like the AP1000®, offered by Westinghouse Electric Company LLC, Cranberry Township, Pa., make use of passive safety systems that only rely on natural circulation, boiling and condensation to remove the decay heat from the core and spent fuel pool. It is desirable to apply these passive safety system principals to a small modular reactor design and, preferably, simplify the design while still maintaining the safety margins of active systems as was provided for in U.S. application Ser. No. 13/495,083, filed Jun. 13, 2012, entitled “Small Modular Reactor Safety Systems.” In many of these Generation III+ pressurized water reactors and small modular reactors which feature passive cooling systems that remove decay heat from the reactor core during a postulated accident, the systems need to be taken out of service before the reactor can be refueled. For a reactor design to be truly passive, it must be able to passively cool fuel in the reactor and spent fuel pool during all modes of refueling. Accordingly, it is an object of this invention to provide a means for removing decay heat from the reactor core during a postulated accident that will function during all modes of reactor operation including, continuously, during a refueling outage. It is a further object of this invention to provide such a passive safety system that will function during a station blackout for an extended period of time. These and other objects are achieved by a nuclear power generating facility having a containment building and an elongated reactor vessel housed within the containment building. The reactor vessel has a nuclear core having fissile material in which fission reactions take place and an open end axially spaced from the nuclear core, with the open end sealed by a head at a flange. A spent fuel pool is supported outside the containment at an elevation that extends substantially above the reactor vessel, with the spent fuel pool being in fluid communication with an interior of the reactor vessel through a first valve. The nuclear power generating facility further includes an ultimate heat sink coolant reservoir whose upper coolant level under normal operation of a nuclear power generating facility is supported at an elevation substantially above the spent fuel pool. A lower portion of the ultimate heat sink reservoir is in fluid communication with the spent fuel pool through a second valve whose operation is controlled by the level of coolant in the spent fuel pool to maintain the coolant in the spent fuel pool at approximately a preselected level. Preferably, the first valve is either passively operated and/or designed to fail in an open position. Similarly, it is desirable that the second valve is either passively or manually operated and in one embodiment, the second valve is a float valve. In accordance herewith, the nuclear power generating facility may also include a passive safety system supported within the containment building approximately at or above a first elevation of the flange and structured to maintain a given level of coolant within the reactor vessel for a first selected period of time, when the coolant level in the reactor vessel unintentionally drops. However, the passive safety system is structured to be out of operation during a refueling of the nuclear core. The nuclear power generating facility may also include a refueling canal establishing a fluid communication path between an inside of the containment building at an elevation above the reactor vessel flange and the spent fuel pool, through which a fuel assembly can pass. Means are also provided for isolating the fluid communication path from the inside of the containment. A refueling cavity may also be supported above the reactor vessel flange and the reactor vessel may be fitted with a branch coolant line. Preferably a gauge is provided on the branch coolant line that has an output indicative of the coolant level above the core to control the first valve to adjust the coolant level to a preprogrammed level. In one embodiment, the gauge is a pressure gauge. The invention also contemplates a method of passively, safely maintaining the coolant level of the nuclear power generating facility described heretofore, above the nuclear core for an extended period of time during a facility outage in which the reactor vessel is substantially depressurized. The method includes the step of sensing a level of coolant above the nuclear core. The method then controls the first valve to drain coolant from the spent fuel pool into the reactor vessel to maintain the coolant within the reactor vessel at a preprogrammed level above the nuclear core. The method also controls the second valve to drain coolant from the ultimate heat sink coolant reservoir into the spent fuel pool to maintain the coolant in the spent fuel pool at approximately the preselected level. In the foregoing embodiment, in which the nuclear power generating facility has a station blackout, the method includes the steps of opening the first and second valves and flooding at least a portion of the containment vessel. This embodiment also may include a branch coolant line connected to the reactor vessel and a gauge on the branch coolant line having an output indicative of a coolant level above the nuclear core, including the steps of controlling the first valve in response to the output indicative of the coolant level above the core to maintain the coolant at the preprogrammed level. Desirably, the preprogrammed level is approximately at the reactor vessel flange. In this embodiment wherein the nuclear power generating facility includes a refueling cavity supported above the reactor vessel flange. After the reactor vessel head has been removed, the gauge control is no longer needed to the level of coolant above the nuclear core within the refueling cavity. The first valve is opened and water drains from the spent fuel pool to the refueling tank until the water levels match. Furthermore, in this embodiment wherein the nuclear power generating facility includes a refueling canal, establishing a fluid communication path between an inside of the refueling cavity at an elevation above the reactor vessel flange, and the spent fuel pool, through which a fuel assembly can pass, and means for isolating the fluid communication path from the inside of the refueling cavity, the method further includes the steps of opening the means for isolating the fluid communication path and controlling a level of the coolant within the refueling cavity through the fluid communication path. In such event, under circumstances where the facility has a station blackout, the method further includes the steps of opening the first valve by virtue of its fail safe position ensuring that a required water level above the core is maintained. In the short refueling window in which the reactor has been disassembled but the refueling tank has yet to be installed, this action results in flooding the containment vessel. Water level is maintained by the passive action of the second valve which maintains the spent fuel pool level. FIGS. 3 and 4 illustrate a small modular reactor design available from Westinghouse Electric Company LLC, Cranberry Township, Pa., to which this invention may be applied, though it should be appreciated that the invention can also be applied to a conventional pressurized water reactor such as the one illustrated in FIGS. 1 and 2. FIG. 3 shows a perspective view of the reactor containment 11, partially cut away, to show the pressure vessel 10 and its internal components. FIG. 4 is an enlarged view of the pressure vessel shown in FIG. 3. The pressurizer 58 is common to most pressurized water reactor designs, though not shown in FIG. 1, and is typically included in one loop to maintain the systems' pressure. In the small modular reactor design illustrated in FIGS. 3 and 4, the pressurizer 58 is integrated into the upper portion of the reactor vessel head 12 and eliminates the need for a separate component. It should be appreciated that the same reference characters are employed for corresponding components among the several figures. A hot leg riser 60 directs primary coolant from the core 14 to a steam generator 18 which surrounds the hot leg riser 60. A number of coolant pumps 16 are circumferentially spaced around the reactor vessel 10 at an elevation near the upper end of the upper internals 26. The reactor coolant pumps 16 are horizontally mounted axial flow canned motor pumps. The reactor core 14 and the upper internals 26, except for their size, are substantially the same as the corresponding components previously described with regard to FIGS. 1 and 2. A further understanding of the operation of the small modular reactor illustrated in FIGS. 3 and 4 can be found in U.S. patent application Ser. No. 13/495,050, filed Jun. 13, 2012, entitled “Pressurized Water Reactor Compact Steam Generator.” Generation III+ pressurized water reactors such as the AP1000® nuclear plant design and small modular reactors like the one just described often feature passive cooling systems that remove decay heat from the reactor core during a postulated accident. In many plant designs, these systems need to be taken out of service before the reactor can be refueled, which is typically every 18 to 24 months. This invention provides a means to passively cool nuclear fuel in a pressurized water reactor during refueling. This invention employs gravity and a series of valves, that can be aligned using battery reserves or fail in a safe position, to maintain water above the reactor core during reactor disassembly and refueling. The embodiment described hereafter applies these principles to a small modular reactor with passive safety systems similar to those disclosed above, however, this principle can be applied to any pressurized water reactor with a compatible plant layout. In the case of the small modular reactor, the embodiment disclosed herein maintains a large reserve of water 90 within or outside the reactor building, which is used to remove decay heat from the reactor core 14 after the reaction has been successfully stopped. Decay heat, typically about one percent of reactor power, is removed by boiling this large reserve of water, known as the ultimate heat sink. The ultimate heat sink 90 is preferably supported, or at least has an outlet that is above the elevation of the spent fuel pool 80 so that water can drain from the ultimate heat sink 90 into the spent fuel pool 80 by gravity as shown in FIGS. 5 through 13. Similarly, the spent fuel pool 80 is maintained at an elevation that is preferably above the core 14 with an outlet conduit 74 well above the core 14 so that coolant in the spent fuel pool can drain into the reactor vessel 10 through inlet 98 by gravity. The outlet 100 from the spent fuel pool through the conduit 74 to the reactor vessel is preferably high enough so that the spent fuel 82 within the pool does not become uncovered and the pool maintains an adequate depth to satisfactorily cool the spent fuel in the pool. The large volume of water in the ultimate heat sink 90 can allow the plant to maintain a safe shutdown condition without outside support for many days. The number of days is determined by the size of the pool in the ultimate heat sink 90. If decay heat is not removed effectively from the reactor core 14 or the spent fuel 82, the fuel cladding material could exceed its design temperature resulting in loss of integrity and failure of a fuel. This condition is commonly known as a meltdown. Plant safety systems figuratively represented by reference character 66 in FIGS. 5 through 13 are designed to deal with all postulated accidents. These systems shut down the nuclear reaction and begin removing decay heat from the reactor core 14 when an adverse operating event is detected. Nuclear power plants are designed to spend the vast majority of their time producing steam to ultimately make electricity. This condition is commonly referred to as normal operation. Every 18-24 months, the plant will shut down normally to replenish its fuel. During refueling, the water level in the reactor coolant system is lowered so that the reactor can be disassembled, allowing access to the fuel assemblies in the core. The water in the reactor coolant system is an integral part of the safety system that is designed to remove decay heat. During refueling, safety systems designed to remove decay heat may be taken out of service because of the reduced water level. This invention describes how the water level in the plant can be maintained at the appropriate level for the various stages of refueling using the spent fuel pool 80. It takes a significant amount of heat to change the phase of water; therefore, heat continues to be removed from the core by boiling this water. This is different from existing traditional pressurized water reactor designs that use dedicated storage tanks to manage reactor coolant system inventory as explained in application Ser. No. 13/495,083, filed Jun. 13, 2012 (RTU 2011-011). In accordance with this invention, the spent fuel pool level 84 is maintained passively from the ultimate heat sink 90, preferably using a passively operated valve 88. The valve 88 may be a float valve, other passively operated valve, or a fail in a safe position valve which opens the conduit 86 from the ultimate heat sink 90 into the spent fuel pool 80 when the pool level 84 is reduced below a preset limit, as illustrated in the embodiments shown in FIGS. 5-13. The following sections describe the arrangement and the function of one or more embodiments of the system claimed hereafter, throughout the refueling process as illustrated in FIGS. 5-13. FIGS. 5-9 illustrate a normal refueling sequence. FIGS. 10-13 show how the water levels are passively maintained during postulated accidents, including a station blackout. FIG. 5 shows the plant in normal operation. The water in the ultimate heat sink 90 and in the spent fuel pool 80 is maintained at the required levels. The reactor coolant system within the vessel 10 and the safety system components are also at the full levels 72 and 68. In FIG. 6, the water level 72 in the reactor coolant system has been reduced by moving water from the reactor coolant system to the spent fuel pool 80 through conduit 74 and valve 76. This raises the spent fuel pool water level 84. Since the spent fuel pool volume is very large compared to the reactor coolant system, the level 84 is raised only a few inches. A reactor coolant system vent 70 is opened to allow the level 72 in the vessel 10 to drop. A pressure gauge 78 on one of the reactor branch lines 102 is used to measure the level 72 in the reactor coolant system. In FIG. 7, the reactor vessel closure head 12 and upper internals are removed from the reactor vessel 10 so the fuel assemblies in the core 14 can be accessed. FIG. 8 shows the installation of a refueling tank 94. This tank can be integral to the refueling machine used to move the fuel, as disclosed in U.S. patent application Ser. No. 13/461,821, filed May 2, 2012, entitled “A Method of Refueling a Nuclear Reactor,” or integral to the reactor containment building 11 design. The refueling tank 94 allows for the water to be drained, from the spent fuel pool 80 to the reactor vessel 14 through the reactor vessel penetration 98, until the level in the tank is at the same level as the spent fuel pool. At this point, the gauge 78 located on the branch line 102 is not needed to maintain the water level in the refueling tank 94. FIG. 9 shows the installation (or opening) of the fuel transfer canal 96. At this point, the pools 80 and 94 are connected and the levels 84 and 72 are maintained through the transfer canal connection 96 and normal refueling can begin. FIGS. 10-13 illustrate the response of this embodiment to a number of postulated accident conditions including one in which off-site power is lost during various stages of refueling. FIG. 10 shows an event wherein the coolant level 72 above the core 14 in the reactor vessel 10 is being maintained by water from the spent fuel pool 80 and the spent fuel pool level 84 is being maintained by the water in the ultimate heat sink tank 90 (though it should be appreciated that more than one tank 90 may be used). The pressure gauge 78 on the reactor branch line 102 controls the level of coolant 72 in the reactor vessel 10. At this stage of refueling, the level in the reactor coolant system has been reduced to a level at which the other passive safety systems 66 have been taken out of service. Vent 70 in the reactor coolant system allows the steam to exit the system into the containment 11. The steam will either condense on the containment vessel walls or be filtered and released to the atmosphere. FIG. 11 again shows the reactor level 72 being maintained by the spent fuel pool 80 and the spent fuel pool level 84 being maintained passively during a loss of off-site power event by the ultimate heat sink 90 and passive valve 88 through the conduit 86. In this stage of refueling, the reactor vessel head 12 has been removed from the reactor vessel 10. The steam generated from the heated reactor coolant is vented through the flange of the open reactor vessel 10 and condenses on the containment vessel 11 or is released to the atmosphere after passing through filters that would contain radioactive contaminants. In FIG. 12, the refueling tank 94 and the refueling canal 96 have been filled to match the level 84 of the spent fuel pool 80 and the fuel transfer canal has been opened at the opening 62 (shown in FIG. 3) and flooded prior to a station blackout. At this point in the refueling process, the connection through the transfer canal 96 maintains water level above the reactor core 14. Under these circumstances, the pressure gauge 78 is no longer required to maintain this level. The ultimate heat sink 90 is still used to passively maintain the level 84 in the spent fuel pool 80. Since the spent fuel pool 80 is directly connected to the ultimate heat sink, the ultimate heat sink tank 90 maintains the water level above the core 14. FIG. 13 illustrates a condition resulting from a loss of off-site power and loss of DC backup power which has occurred during the small window of the plant refueling outage in which the reactor has been disassembled but the refueling tank has not yet been installed. In this case, the valve 76 in the conduit 74, between the spent fuel pool 80 and the reactor penetration 98 fails in the safe position. With this valve open, the level 72 is not controlled and the spent fuel pool 80 continues to drain to the containment vessel 11 until the level 72 in the containment vessel matches the level 84 in the spent fuel pool. Again, steam 104, produced by boiling the water within and above the reactor vessel 10, will condense on the walls of the containment vessel 11 or be filtered before being released to the atmosphere. The ultimate heat sink tank 90 drains into the spent fuel pool to maintain the required level 84 in the spent fuel pool 80 and subsequently the containment vessel 11. This arrangement ensures that the reactor core 14 and the spent fuel 82 remain covered with water for the period required by the design. This time period is controlled solely by the amount of water available in the ultimate heat sink tank 90, thus providing truly passive safety during refueling. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof. |
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claims | 1. A molten salt reactor, comprising:a tubular reactor core comprising graphite and defining an internal space; anda graphite fuel wedge that defines multiple fuel channels, wherein the fuel wedge is received within the internal space, wherein an outer surface of the fuel wedge comprises a first contoured shape, wherein an internal surface of the internal space comprises a second contoured shape, wherein the first contoured shape of the fuel wedge substantially corresponds in shape to and contacts the second contoured shape of the internal surface of the tubular reactor core as the reactor core operates, and wherein the fuel channels are configured to allow a fissionable fuel to flow from a first end of the fuel wedge to a second end of the fuel wedge. 2. The reactor of claim 1, wherein the fuel wedge comprises:a first section comprising a first portion of the fuel channels; anda second section comprising a second portion of the fuel channels,wherein the first section and the second section are disposed end to end within the internal space such that the first and second portions of the fuel channels are aligned such that the fissionable fuel flows from the first portion to the second portion of the fuel channels during reactor operation. 3. The reactor of claim 2, further comprising an alignment pin that extends between the first section and the second section of the fuel wedge to keep the first and second portions of the fuel channels aligned. 4. The reactor of claim 2, further comprising a seal comprising graphite, which is disposed between the first section and the second section of the fuel wedge. 5. The reactor of claim 2, further comprising a raised graphite seal portion that extends from a first end of the first section of the fuel wedge, and wherein a second end of the second section of the fuel wedge comprises a graphite recess that is configured to receive the raised graphite seal portion to form a seal between the first section and the second section. 6. The reactor of claim 1, further comprising a variable pump that is configured to force the fissionable fuel through the fuel channels at a variable rate. 7. The reactor of claim 1, wherein the reactor core further comprises a first reactor end cap that caps a first portion of the reactor core, and wherein the reactor further comprises a diffuser plate defining a plurality of holes, the diffuser plate being disposed between a portion of the first end cap and the fuel wedge. 8. The reactor of claim 1, wherein the fuel wedge further comprises a standoff spacer that is disposed at an end of the fuel wedge and that is configured to channel the fissionable fuel into the fuel channels. 9. The reactor of claim 5, wherein a second end of the first section of the fuel wedge comprises a standoff spacer that is configured to channel the fissionable fuel into the fuel channels. 10. A molten salt reactor, comprising:a tubular molten salt reactor core comprising graphite and defining a tubular internal space;a first fuel wedge comprising graphite and defining a first set of fuel channels; anda second fuel wedge comprising graphite and defining a second set of fuel channels,wherein the first and second fuel wedges each comprise an outer surface that is configured to contact and to substantially follow an internal circumference of the tubular internal space as the reactor operates,wherein the first and second fuel wedges are received within the internal space,wherein the first and second sets of fuel channels comprise a fissionable fuel comprising a molten salt, and wherein the first and second sets of fuel channels allow the fissionable fuel to flow from a first end of the internal space to a second end of the internal space through the first and second sets of fuel channels. 11. The reactor of claim 10, further comprising a fuel pin rod that is smaller than the first fuel wedge, that has a different shape than the first fuel wedge, that defines a first fuel channel, and that is disposed between the first fuel wedge and the second fuel wedge. 12. The reactor of claim 10, wherein the molten salt reactor core is longer than about 4 meters. 13. The reactor of claim 10, wherein the first fuel wedge comprises:a first section comprising a first length of the first set of fuel channels; anda second section comprising a second length of the first set of fuel channels,wherein the first section and the second section are disposed end to end within the internal space such that the first and second lengths of the first set of fuel channels are aligned and allow the fissionable fuel to flow directly from the first length to the second length of the first set of fuel channels as the reactor operates. 14. The reactor of claim 13, further comprising an alignment pin that extends between the first section and the second section of the first fuel wedge to keep the first and second lengths of the first set of fuel channels aligned with each other. 15. The reactor of claim 13, further comprising a raised seal portion that comprises graphite and extends from a first end of the first section of the first fuel wedge, and wherein a second end of the second section of the first fuel wedge comprises a recess that is configured to receive the raised seal portion to form a seal between the first section and the second section. 16. The reactor of claim 15, further comprising an alignment pin that extends between the first section and the second section of the first fuel wedge to keep the first and second lengths of the first set of fuel channels aligned with each other. 17. The reactor of claim 10, wherein a length of a longitudinal axis of the molten salt reactor core runs at angle between 5° and about 45° with respect to a horizontal plane. 18. A molten salt reactor, comprising:a tubular reactor core comprising graphite and defining an internal space; anda first graphite fuel wedge that defines multiple fuel channels,wherein the first fuel wedge is received within the internal space,wherein the fuel channels are configured to allow a fissionable fuel to flow from a first end to a second end of the first fuel wedge,wherein the first fuel wedge has a substantially sector-shaped prism configuration having an outer surface comprising a first contour that is configured to substantially match and contact a shape of a contour of an internal surface of the internal space, andwherein the first fuel wedge comprises a first fuel wedge section and a second fuel wedge section that are coupled together end to end with a seal comprising graphite, which is disposed between the first and second sections such that the fissionable fuel is able to flow between the first and second sections. 19. The reactor of claim 18, further comprising an alignment pin that extends between the first fuel wedge section and the second fuel wedge section such that the fissionable fuel flows between the first and second fuel wedges as the reactor operates. 20. The reactor of claim 18, wherein the reactor core is disposed in and in contact with a reflector comprising graphite. |
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summary | ||
050892134 | claims | 1. A nuclear fuel identification code reader comprising: an optical sensor for detecting a first nuclear fuel identification code, marked on a fuel assembly for identifying said fuel assembly; an ultrasonic wave sensor for detecting a second nuclear fuel identification code marked on said fuel assembly at a position adjacent to said first nuclear fuel identification code and which identifies said fuel assembly, said second nuclear fuel identification code being expressed in a different form from said first nuclear fuel identification code; drive means for driving said optical sensor and ultrasonic wave sensor above said fuel assembly whose codes are to be detected; first recognition means for recognizing said first nuclear fuel identification code based on information derived from said optical sensor; means for enabling the detection of said first nuclear fuel identification code by the optical sensor; means for enabling the detection of said second nuclear fuel identification code by said ultrasonic wave sensor when said first recognition means cannot recognize said first nuclear fuel identification code; and second recognition means for recognizing said second nuclear fuel identification code based on information derived from said ultrasonic wave sensor. means for comparing a current recognition result of one of said first and second nuclear fuel identification codes and a past recognition result thereof. means for determining the necessity of detection of the second nuclear fuel identification code by said ultrasonic wave sensor in accordance with the recognition result of said first recognition means; and means for comparing a current recognition result of one of said first and second recognition means and a past recognition result thereof. an optical sensor for detecting a first nuclear fuel identification code marked on a fuel assembly, for identifying said fuel assembly; an ultrasonic wave sensor for detecting a second nuclear fuel identification code marked on said fuel assembly at a position adjacent to said first nuclear fuel identification code and which identifies said fuel assembly, said second nuclear fuel identification code being expressed in a different form from said first nuclear fuel identification code; drive means for driving said optical sensor and ultrasonic wave sensor above said fuel assembly whose codes are to be detected; determination means for determining the necessity of detection of said second nuclear fuel identification code by said ultrasonic wave sensor in accordance with the recognition result of the first nuclear fuel identification code based on information derived from said optical sensor; and means for enabling the detection of said second nuclear fuel identification code by said ultrasonic wave sensor when said determination means determines that it is necessary to detect said second nuclear fuel identification code. means for comparing a current recognition result of one of said first and second nuclear fuel identification codes and a past recognition result thereof. an optical sensor for detecting a first nuclear fuel identification code marked on a fuel assembly, for identifying said fuel assembly; an ultrasonic wave sensor for detecting a second nuclear fuel identification code marked on said fuel assembly at a position adjacent to said first nuclear fuel identification code and which identifies said fuel assembly, said second nuclear fuel identification code being expressed in a different form from said first nuclear fuel identification code; drive means for driving said optical sensor and ultrasonic wave sensor above said fuel assembly whose codes are to be detected; means for enabling the detection of said first nuclear fuel identification code by said optical sensor; and means for enabling the detection of said second nuclear fuel identification code by said ultrasonic wave sensor when the recognition of said first nuclear fuel identification code based on information derived from said optical sensor is not successful. an optical sensor for detecting a first nuclear fuel identification code marked on a fuel assembly, for identifying said fuel assembly; an ultrasonic wave sensor for detecting a second nuclear fuel identification code marked on said fuel assembly at a position adjacent to said first nuclear fuel identification code and which identifies said fuel assembly, said second nuclear fuel identification code being expressed in a different form from said first nuclear fuel identification code; drive means for driving said optical sensor and ultrasonic wave sensor above said fuel assembly whose codes are to be detected; means for enabling the detection of said first nuclear fuel identification code by said optical sensor; means for selecting one of continued detection by said optical sensor and detection by said ultrasonic wave sensor in accordance with a recognition result of said first nuclear fuel identification code; and means responsive to said selecting means for enabling detection of said second nuclear fuel identification code by said ultrasonic wave sensor when detection by said ultrasonic wave sensor is selected. an optical sensor for detecting a first nuclear fuel identification code marked on a fuel assembly, for identifying said a fuel assembly; an ultrasonic wave sensor for detecting a second nuclear fuel identification code marked on said fuel assembly at a position adjacent to said first nuclear fuel identification code and which identifies said fuel assembly, said second nuclear fuel identification code being expressed in a different form from said first nuclear fuel identification code; determination means for determining the necessity of detection of said second nuclear fuel identification code by said ultrasonic wave sensor in accordance with a recognition result of the first nuclear fuel identification code based on information derived from said optical sensor; means for enabling detection of said second nuclear fuel identification code by said ultrasonic wave sensor when said determination means determines that it is necessary to detect said second nuclear fuel identification code; pick-up means for picking up a Chelencoff light generated in said fuel assembly; means for image processing a video signal of the Chelencoff light picked up by said pick-up means; and drive means for driving said optical sensor, ultrasonic wave sensor and said pick-up means above said fuel assembly whose codes are to be detected. means for creating a Chelencoff light pattern based on binary data derived from said image processing means; and means for comparing the Chelencoff light pattern with a reference Chelencoff light pattern for a corresponding nuclear fuel identification code. an optical sensor for detecting a first nuclear fuel identification code marked on a fuel assembly, for identifying said fuel assembly; nuclear fuel identification code sense means including an ultrasonic wave sensor for detecting a second nuclear fuel identification code marked on said fuel assembly at a position adjacent to said first nuclear fuel identification code and which identifies said fuel assembly, said second nuclear fuel identification code being expressed in a different form from said first nuclear fuel identification code; drive means for driving said nuclear fuel identification code sense means above said fuel assembly whose codes are to be detected; means for enabling detection of said first nuclear fuel identification code by said optical sensor; means for determining whether the sensing and recognition of the nuclear fuel identification code are to be effected by said ultrasonic wave sensor in accordance with a recognition result of said first nuclear fuel identification code based on information derived from said optical sensor; and means for enabling detection of said second nuclear fuel identification code by said ultrasonic wave sensor when the sensing and recognition are determined to be effected. first signal processing means for recognizing a first nuclear fuel identification code marked on a fuel assembly based on a video signal of the first nuclear fuel identification code detected by an optical sensor, and for enabling the detection of a second nuclear fuel identification code marked on the fuel assembly at a position adjacent to the first nuclear fuel identification code by an ultrasonic wave sensor when the first nuclear fuel identification code cannot be recognized; and second processing means for recognizing the second nuclear fuel identification code based on information derived from said ultrasonic wave sensor. 2. A nuclear fuel identification code reader according to claim 1, further comprising: 3. A nuclear fuel identification code reader according to claim 1, wherein said optical sensor detects the first nuclear fuel identification code expressed in a character form, and said ultrasonic wave sensor detects the second nuclear fuel identification code corresponding to the first nuclear fuel identification code, expressed in a form of a plurality of separated recesses. 4. A nuclear fuel identification code reader according to claim 1, further comprising: 5. A nuclear fuel identification code reader according to claim 1, wherein said drive means includes a movable truck, a laterally movable truck movably mounted on said movable truck and means mounted on said laterally movable truck for vertically driving said optical sensor and said ultrasonic wave sensor. 6. A nuclear fuel identification code reader according to claim 1, further comprising control means for applying a first control signal to said drive means to position said optical sensor above a portion of said fuel assembly on which said fuel identification codes are portion on which said fuel identification codes are marked when the recognition of first nuclear fuel identification code based on information derived from said optical sensor is not successful. 7. A nuclear fuel identification code reader according to claim 1, wherein said drive means also drives said fuel assembly, and further comprising a control unit for controlling the movement of said drive means. 8. A nuclear fuel identification code reader according to claim 1, wherein said drive means includes vertical moving means on which said optical sensor and ultrasonic wave sensor are mounted. 9. A nuclear fuel identification code reader comprising: 10. A nuclear fuel identification code reader according to claim 9, further comprising: 11. A nuclear fuel identification code reader according to claim 9, wherein said optical sensor detects said first nuclear fuel identification code expressed in a character form, and said ultrasonic wave sensor detects said second nuclear fuel identification code corresponding to said first nuclear fuel identification code, expressed in a form of a plurality of separated recesses. 12. A nuclear fuel identification code reader according to claim 9, wherein said drive means includes a movable truck, a laterally movable truck movably mounted on said movable truck, and means mounted on said laterally movable truck for vertically driving said optical sensor and said ultrasonic wave sensor. 13. A nuclear fuel identification code reader comprising: 14. A nuclear fuel identification code reader according to claim 13, wherein said optical sensor detects said first nuclear fuel identification code expressed in a character form, and said ultrasonic wave sensor detects said second nuclear fuel identification code corresponding to said first nuclear fuel identification code, expressed in a form of a plurality of separated recesses. 15. A nuclear fuel identification code reader according to claim 13, wherein said drive means is comprised of a movable truck, a laterally movable truck movably mounted on said movable truck, and means mounted on said laterally movable truck for vertically driving said optical sensor and said ultrasonic wave sensor. 16. A nuclear fuel identification code reader according to claim 15, wherein said laterally movable truck is provided with fuel assembly clamp means. 17. A nuclear fuel identification code reader comprising: 18. A nuclear fuel identification code reader according to claim 17, wherein said optical sensor detects said first nuclear fuel identification code expressed in a character form, and said ultrasonic wave sensor detects said second nuclear fuel identification code corresponding to said first nuclear fuel identification code, expressed in a form of a plurality of separated recesses. 19. A nuclear fuel identification code reader comprising: 20. A nuclear fuel identification code reader according to claim 19, further comprising: 21. A nuclear fuel identification code reader according to claim 19, further comprising means for creating a Chelencoff light pattern by combining video signals picked up by said pick-up means inclined by a predetermined angle at two positions symmetric to the handle mounted on said fuel assembly. 22. A nuclear fuel identification code reader according to claim 19, wherein said drive means includes a movable truck, a laterally movable truck movably mounted on said movable truck, and means mounted on said laterally movable truck for vertically driving said optical sensor and said ultrasonic wave sensor. 23. A nuclear fuel identification code reader comprising: 24. A signal processor for a nuclear fuel identification code sensor comprising: |
abstract | A high-temperature gas-cooled reactor steam generating system comprises a plurality of nuclear steam supply systems, a high-pressure cylinder (21), a low-pressure cylinder (22), a condenser (23), a condensate pump (24), a low-pressure heater (25), a deaerator (26), a water supply pump (27), and a high-pressure heater (28) which are sequentially connected end to end to form a close steam loop. On one hand, the inherent safety of the reactor is guaranteed and the generating system is simplified with the inherent safety. On the other hand, the scale economy of the steam engine system and other systems of a whole power station is guaranteed through batch copy, a shared auxiliary system and a scale effect. |
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048250880 | abstract | An improved, lightweight cask assembly is disclosed herein. The cask assembly generally comprises at least two, cylindrically shaped and concentrically disposed structural walls formed from a titanium alloy and a shielding wall disposed within and supported by the titanium structural walls. Both the inner and outer walls are formed from a high-strength titanium alloy. The use of such an alloy advantageously allows the thickness of the inner wall to be minimized, thereby optimizing the shielding geometry of the shielding wall and minimizing the weight of the amount of shielding material used in the cask. The use of such an alloy in the outer structural wall also minimizes the weight of the outer wall thereby contributing to the weight reduction of the cask assembly as a whole. The upper and lower edges of the inner and outer structural walls are bound together by a reinforcing ring and an end plate assembly likewise formed from a titanium alloy. In the preferred embodiment, two separate shielding walls made of depleted uranium and particles of boron suspended in a silicone matrix are disposed in the spaces between inner, intermediate and outer titanium walls, respectively. |
040654001 | description | DETAILED DESCRIPTION The drawing illustrates apparatus that may be used in practicing this invention. The calciner vessel 10 includes a lower reaction section 12 and an upper recovery of disengaging section 14 which is of greater cross-sectional area than the lower reaction portion 12 to permit disengaging of particles from the gas, as is generally known in the art. Beneath the lower reaction section 12 is a gas inlet section 16 which receives a fluidizing gas 18 such as air from a suitable source 20. The gas is uniformly distributed throughout the cross-section of the reaction section 12 by gas distribution plate 22 in order to uniformly disperse and fluidize the bed material 24 and effect the desired reaction. The fluidizing gas and elutriated particles then pass into disengaging section 14 and subsequently exit section 14 through port 29 via conduit 30 interconnecting an upper portion 28 of section 14 with a solids/gas separator means 31 such as a cyclone in combination with appropriate filters medium as sintered metal filters or the like. The elutriated solids are separated from the gas in separator means 31 and pass through conduit 32 and collected or otherwise treated in collecting means or receptacle 34. Arrow 33 indicates the removal of gases from the solids separator means. Conduit 36 interconnects bed support wall or gas distribution plate 22 with a bottom wall 38 of calciner vessel 10 and provides a passageway for removal of particulate material from the lower reaction section 12 to receptacle 34, as indicated by arrow 37. A valve 40 is likewise provided to control removal of the particulate material. Conduit 36 and valve 40 may be used, for example, when it is desired to completely remove the bed from section 12 at the end of a run, or in other like circumstances. Normally the valve is closed during continuous high level radioactive liquid waste solidification processing. Appropriate heating means, such as an electrical resistance heater, may be used to heat the fluidized bed 24 within the calciner vessel. Alternatively, a fuel may be supplied to and combusted within lower reaction section 12 to provide the desired heating, as generally known in the art. A conduit, indicated by arrow 42 may feed fuel from fuel source 44 into lower reaction portion 12 for these latter described systems. A suitable fuel mixture in such a system may be such as oxygen and propane or kerosene. Heating of the fluidized bed may be to a temperature of from about 400.degree. to about 1300.degree. C and preferably is at a temperature of from about 400.degree. to about 800.degree. C. Heating in these ranges provides desirable process and product characteristics such as high combustion rates, minimal nitrate content in product, and the like. The inert bed material 24 is granular, inert silica (SiO.sub.2) in the form of sand or the like, which inert bed material is durable and insensitive to extreme temperature changes, and is not chemically affected by the high level radioactive waste feed material or melted or otherwise affected by the temperatures encountered in processing. The inert bed material or silica particles may have an average particle diameter of from about 0.20 to about 0.40 millimeters. While silica is referred to herein, other inert bed materials may likewise be used. Fluidizing gas is passed through the gas distribution plate 22 at a velocity sufficient to effectively fluidize the material to the desired level as known in the art. High level liquid radioactive waste feed solutions from source 45 is passed into lower reaction section 12 through feed conduit 46. The radioactive liquid waste is atomized by introducing an atomizing gas from atomizing gas source 48 through suitable conduit 50 which is interconnected with conduit 46 at an appropriate location. The atomizing gas may be air or another suitable gas. One or more atomizing nozzles (not shown) may be interconnected with conduit 46 and strategically located to feed the atomized waste into the fluid bed. For example, a plurality of nozzles may be conveniently located around a circumference of lower reaction portion 12 for atomizing the liquid radioactive waste and directing it into the fluidized bed section. It may be desirable to locate the atomizing nozzles adjacent the combustion zone, if fuel combustion is used for heating, to achieve high feed rates. Atomized waste feed at a generally upper portion of the fluidized bed section may enhance attrition, and would be favored where a high attrition rate and a low inert bed loss through carry-over are desired. The feed solutions introduced into lower reaction portion 12 are converted to metal oxides (referred to herein as calcine) and nonmetal oxides (off-gas). The calcine either coats the inert silica bed particles, is spray dried, or coats and attrits from the inert particles. The result may be influenced by manipulating several variables such as feed composition, atomizing gas rates, temperatures, etc. Overflow conduit 52 provides a passageway for overflow removal of a portion of the calcined material from the fluidized bed. As the atomized liquid waste is fed into the fluidized bed, the radioactive waste becomes calcined, may dry as a spray, and may coat onto the inert bed material such as silica particles and attrit therefrom through contact with other particulate material from movement thereof in the fluidized bed such that the attrited particles or fines of calcine are elutriated and separted from the gas by solids and gas separator means 31. An upper portion of the fluidized bed particle may be removed through overflow into conduit 52 to effect rmoval of a portion of the calcine material with some inert bed material. The resultant product collected in receptacle 34 may be readily vitrified. Further, this enables the provision of fresh inert bed material from inert bed particle reservoir or source 53 through conduit 54 to continuously maintain the calcination process. At or near the same time as waste feed is introduced, addition of inert material from the inert material storage container to the reaction chamber may be initiated. The rate of addition and whether it is continous or semi-continuous is dependent on the specific calcination process being operated. As a lower limit, the rate of inert addition will be equal to the bed material attrited and elutriated to maintain the desirable fluidization quality. An upper limit would be dependent on the next processing step, i.e., the weight ratio of calcined oxides to inerts desired, calcine inventory of the bed, etc. Acceptable operation from a nil rate to ten times the calcine oxide generation have been practiced. Control of the bed level is appropriately maintained by the overflow conduit in the bed or by using a control valve in that conduit. If waste feed rates change, inert addition is correspondingly adjusted. This ability to have continuous control over the material in the reaction chamber is a further distinquishing feature of this process from previous calcination operations. For example, if it is noted that the particles in the bed are getting too big, small sized silica particles may be added, while if the bed particles are getting too small, larger sized silica particles may be added. Thus much better control of the bed and of the resultant product may be achieved. Overflow removal of inert bed material may be minimized if the desired product is to have a low silica content. In this case, it may be further desirable to provide means for increasing attrition and fine formation of the calcined waste, Such as by introducing high velocity jet streams into the fluid bed to enhance or promote turbulence and an increased attrition rate. In a high attrition rate process, the fluidized bed may be comprised of from about 10 to about 15 weight percent calcined radioactive waste, the remainder being inert bed particles such as silica. The resultant calcined waste recovered product would be about 90 weight percent calcined waste and about 10 weight percent silica. If the product is to be vitrified, calcined waste recovered product may be processed to contain about 50 weight percent calcine waste and about 50 weight percent silica. The material that is collected in the collection and/or storage receptacle 34 may be used for processing the calcined material into a furnace to accomplish melting and subsequent transfer to a melt receiver for encapsulation. In the alternative, the calcined material may be stored for an indefinite period. The collection receptacle 34 may be substituted with a suitable melter for vitrification purposes. In one embodiment of the invention, the calciner vessel was 1.7 meters long and had a 17.1 centimeters square bed section and a 24.8 centimeter square disengaging section. The vessel employed a perforated gas distributor plate. As shown in the drawing, product materials exited the calciner vessel by way of an overflow conduit 52, off-gas conduit 30, and a bed removal conduit 36. The overflow conduit 52 and the bed removal conduit 36 communicate with a gas particulate separator means 31 via conduit 32. Gas and particulates were separated by a cyclone and sintered metal filters. Conventional blowback procedures were employed to maintain satisfactory low pressure drop across the filters. In an operation of this invention, a starting bed of inert material of silicon dioxide was fluidized at about 30 centimeter per second superficial velocity while process heat was supplied by the combustion of oxygen and kerosene directly into the bed. As waste feed was introduced though an air atomized nozzle and the calcination reaction occurred, the continuous addition of inert to the bed was started. The calcine coated the particles, was spray dried, or coated and attrited from the material. Product was overflowed and/or elutriated from the bed to maintain the proper inventory. By using jet grinders, high attrition type feed nozzles or operating conditions (such as high temperature, high fluidizing velocity) conducive to the generation of fines, the amount of calcine in the bed was reduced significantly because of high attrition rates and of spray drying of the waste. The particle size of the inert silica particulate material added was generally in the size range of from about 0.2 to about 0.4 millimeters in diameter. The rate of inert solids addition was dependent on the next processing step and may be generaly equivalent to the bed attrition rate, i.e., that necessary to maintain proper fluidized bed level. When the calcined waste material is to be vitrivied, the weight of inert material added to the fluidized bed would be about equal to the weight of the calcine oxide being generated. Glass frit would be added to the melter used for vitrification. Because the reaction bed is silica, the inert mateial carried over into the collection receptacle or the melter is readily incorporated in the frit. This invention is versatile over wide operating ranges, as indicated by the summary of several runs in the Table. The process readily accommodates most waste compositions. The calciner or reaction vessel has been coupled directly to an in-can melter and indirectly to a continuous ceramic melter without any problems. Sintered metal filters have been used satisfactorily in separating particulate material from gases. TABLE ______________________________________ Feed Types 320 - 575 l/MTU.sup..alpha. 0.01 - 1M Na.sup.b 80 - 210 g oxide/l.sup.c Feed Rates 20 - 40 l/hr 80 - 120 l/hr/ft.sup.2 of bedcross sectional area Atomizing Air to Feed Volumetric Ratios 200 -700 Vessel Operating Pressure 740 mm mercury Operating Temperature 500 -800.degree. C Bed Properties 0.2 - 0.5 mm dia. 14 - 50% calcine Avg. .about. 20% Product Properties 0.1 - 0.3 mm dia. Heavy fines or no fines dependent on feed, etc. ______________________________________ .sup..alpha. MTU = Metric ton of uranium processed .sup.b M Na = Molar Sodium .sup.c g oxide/l = grams calcine per liter of high level waste As noted in the Table, the feed materials ranged from 320 to 575 liters of waste feed obtained from a metric ton of uranium processed. These feed materials contained from 80 to 210 grams of calcine per liter of liquid waste, and contained from 0.01 molar to 1 molar sodium. In the apparatus described having the aforesaid dimensions, the feed rates of high level liquid waste were from 20 to 40 liters per hour, and employed air as the atomizing gas at volumetric ratios of atomizing gas to liquid feed of from about 200 to about 700. In this particular series, the operating temperature ranged from 500.degree. C to 800.degree. C. The calcine product properties ranged from 0.1 to 0.3 millimeters diameter. These were controlled by inert addition rate, varying the feed rates, adjusting the atomizing gas rates, etc. Distinct advantages of this invention are that the fission product inventory in the bed is substantially reduced, with the concurrent effect of substantially reducing the danger of a temperature excursion due to self-heating of the bed following a loss of fluidizing air incident or agglomeration of the bed. Second, the bed is operated in an attriting or grinding fashion to discharge the calcine as an overhead powder, together with an overflow of silica particles having a thin calcine waste coating thereon, and supplementary inert bed particles are added to the fluidized bed to achieve continuous operation. Third, the use of silica particles permits high temperature operation, even as high as from about 400.degree. C to 1300.degree. C. |
063309181 | abstract | A device automatically makes up a riser string from riser spools, each having an ancillary line and dog-type connectors. A support member supports a second riser spool for connection. A torque arm above the support member hinges open to accept a first riser spool and hinges close to engage the first spool and rotate the first spool to align its ancillary line with the ancillary line of the second spool. A guide arm above the support member hinges open to accept the first spool and hinges closed to engage and radially position the first spool while allowing vertical movement of the first spool. Connector actuators are positioned around the support member to actuate the connectors when the first spool is lowered onto the second spool. |
056446152 | description | FIG. 1 shows a relevant part of an X-ray analysis apparatus in which the collimator in accordance with the invention can be used. An X-ray source 2 produces an X-ray beam 4 which is incident on a specimen 6 to be examined. In the specimen 6 the X-ray beam 4 excites X-rays which are analysed according to wavelength by an analyser crystal 8. As this analyser crystal operates according to the well-known Bragg law 2d.sin.delta.=n.lambda. (d=distance between the reflecting lattice planes in the analyser crystal, .delta.=the angle between the incident X-ray beam and the lattice planes, n=the order of the reflection, and .lambda.=the X-ray wavelength), the X-rays incident on the analyser crystal must be parallel, i.e. have only one value of .delta.. To this end, the specimen to be examined is succeeded by a first collimator 10 which selects only the radiation extending in parallel within the (narrow) divergence range of the collimator from the X-ray beam emanating from the specimen 6. The collimator 10 is preceded by a first beam limiter 12 for a first coarse directional selection of the X-rays emanating from the specimen. Depending on the angular position .delta. of the analyser crystal 8 relative to the X-ray beam incident on the crystal, a given wavelength .lambda. in conformity with said Bragg law is selected. This beam is reflected, in the form of a reflected beam 16, in the direction of the X-ray detector, via a second beam limiter 20 and a second collimator 22. The second beam limiter 20 intercepts X-rays scattered upstream of the beam limiter in a variety of locations within the analysis apparatus. The second collimator 22 parallelizes the analysed beam again in order to remove non-desirable directions from the X-rays emanating from the analyser crystal. Finally, the detector 18 measures the intensity of the wavelength thus selected so that after all desired wavelengths have been covered by rotation of the analyser crystal, the intensity has been determined in dependence on the wavelength. FIG. 2 shows a collimator plate for use in a collimator in accordance with the invention. The collimator plate 30 (having a height of, for example 29 mm and a width of, for example 36 mm) is made of tungsten and has a thickness of, for example 0.1 mm. The plate is subdivided into three areas 32a, 32b and 32c with rectangular holes 34, each of which has a width of 9.8 mm and a height of 0.1 mm. These holes can be formed by way of a customary precision manufacturing method, for example by photochemical etching as is customary in the manufacture of integrated circuits. Even though in reality all three areas are fully subdivided into holes, for the sake clarity the Figure does not show the three areas completely filled with holes. As appears more clearly from the part 36 which is shown at an enlarged scale, between the rows of holes 32a, 32b and 32c there is situated a non-interrupted part 38 which has a width of 0.2 mm and serves to strengthen the collimator plate 30. The holes are provided in rows of three adjacent columns, each of which is subdivided into a large number of rows which are situated one over the other. Within a column a vertical period p.sub.1 of 0.2 mm exists, which period equals the distance between two corresponding points of two rows situated one above the other in a column, for example the distance between the upper sides of the rectangular hole 40 and the rectangular hole 42. The period p.sub.1 has a fraction t.sub.1 (of, for example 50%) which is taken up by the opening, for example 40 or 42, so that the vertical dimension of this opening equals t.sub.1 p.sub.1, being 0.1 mm in this numerical example. Similarly, the collimator plate 30 has a period p.sub.2 of 10 mm with an opening fraction t.sub.2 of 98% in the horizontal direction, so that the absolute value of the opening in this direction equals t.sub.2 p.sub.2, being 9.8 mm in this numerical example. FIG. 3 shows a geometrical diagram illustrating the operation of the collimator in accordance with the invention. The Figure is a diagrammatic cross-sectional view of two collimator plates 30a and 30b as shown in FIG. 2. Each of the plates 30a and 30b is subdivided into openings 52a, 52b etc. and 56a, 56b etc. which correspond to the openings 40 or 42 in FIG. 2. Between the openings 52 and 56 there are provided areas 50a, 50b, 50c and 54a, 54b, 54c, respectively, having X-ray absorbing properties. The distance between the openings is determined by the period p which may represent the vertical period p.sub.1 as well as the horizontal period p.sub.2. The period p is subdivided into transmissive areas 52 and 56 amounting to a fraction t, so that the open part is dimensioned t.p, and non-transmissive areas 50 and 54 amounting to a fraction 1-t, so that the non-transmissive part is dimensioned (1-t).p. The collimator is bounded by two outer, identical plates 30a and 30b wherebetween further identical collimator plates are arranged. The outer plates are arranged at a distance d.sub.c from one another, d.sub.c being determined from the maximum desirable angular divergence (defined as half the angle between two extreme rays) of the transmitted X-ray beam, amounting to t.p/d.sub.c. It is assumed that the X-ray beam to be collimated originates from an X-ray source which is not shown in FIG. 3 and which has a large emissive surface area, so that X-rays extending in all directions are present in the X-ray beam incident on the collimator plate 30a. This means that at the top 51 of the opening 52a X-rays extend in all directions, notably in the directions 58, 60, 62 and 64 indicated. X-rays emanating from the point 51 may be transmitted by the corresponding opening 56a in the plate 30b, but not by the other openings 56b etc. in this plate. A boundary line of the beam aimed at the inhibited opening 56b is formed by the line 58. The beam emanating from the point 51 is tangent to the lower side of the absorbing part 54b by way of the line 58; a part of the beam emanating from the point 51 is intercepted by arranging a plate 30c between the plate 30a and the plate 30b, i.e. the part which is tangent to said lower side. This situation occurs if the distance d.sub.1 between the plate 30b and the intermediate plate 30c is: EQU d.sub.1 :d.sub.c =(1-t).p:p (1) wherefrom it follows that: EQU d.sub.1 =d.sub.c (1-t) (2) Below the part of the X-ray beam thus intercepted there is situated a further part which can be intercepted by arranging a further intermediate plate 30d at a distance d.sub.2 from the plate 30c, in which case it analogously holds that (using d.sub.1 =d.sub.c (1-t)): EQU d.sub.2 :{d.sub.c -d.sub.c (1-t)}={(1-t)p}:p (3) wherefrom it follows that: EQU d.sub.2 =d.sub.c.t.(1-t) (4) Similarly, for the distance d.sub.3 between a possibly further plate 30e and 30d it can be deduced that: EQU d.sub.3 =d.sub.c.t.sup.2.(1-t) (5) When this procedure is continued, the general expression for the distance d.sub.n is: EQU d.sub.n =d.sub.c.t.sup.n-1.(1-t) (6) Comparison of the formules (2), (4) and (5) teaches that the ratio of two successive distances between the plates equals the opening fraction t of the period p. A comparable derivation can be performed for a period and an opening fraction extending perpendicularly to the above period and opening fraction, so that transverse collimation can thus be achieved by choosing a different (or the same) value for t (i.e. t.sub.2) in a direction transversely of the direction of the first value of t (i.e. t.sub.1). FIG. 4 shows a housing for the collimator plates in accordance with the invention. The housing consists of a bottom section 70 and a lid section 72. In the bottom section there are provided slots (not shown) in which the collimator plates 30 can be arranged. The position of the collimator plates is thus defined. In the lid section there are also provided slots in which the collimator plates can be arranged. The Figure clearly shows the spacings d.sub.1, d.sub.2, d.sub.3 etc. It is equally visible that the distance between the plates 30b and 30c is comparatively large, so that further elements for influencing the X-ray beam to be collimated can be accommodated in the collimator housing. |
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description | Referring now to the drawings, and more particularly to FIG. 1, an anti-scatter X-ray raster according to the first embodiment has a plurality of parallel tubular channels 1 for an X-rays transporting, enclosed by two parallel planes 2, 3, being perpendicular to the longitudinal axes of the channels 1. The channels form a honeycomb structure (hereinafter referred to as a cellular structure). Their walls are in the form of a cylindrical surface or a lateral surface of a prism. FIG. 6 and FIG. 7 depict a form of the said cellular structure as viewed from of inputs or outputs of the channels. FIG. 6 and FIG. 7 illustrate as well embodiments, when the channels in the cross-section are round or oval, or they are in the form of a regular hexagon. In the latter embodiment there are no spaces between sidewalls of the neighboring channels. However the spaces, when they are represented, can operate as channels, and their availability has not an adverse effect on the transparency of a raster. An anti-scatter raster is produced according to the known technology of producing monolithic X-rays lenses (see, for instance: V. M. Andreevsky, M. V. Gubarev, P. I. Zhidkin, M. A. Kumakhov, A. V. Noshkin, I. Yu. Ponomarev, Kh. Z. Ustok. X-ray waveguide system with a variable cross-section of the sections. The IV-th All Union Conference on Interaction of radiation with Solids (May 15-19, 1990, Elbrus settlement, Kabardino-Balkarian ASSR, USSR). Book of abstracts. Moscow, 1990, pp. 177-178; U.S. Pat. No. 5,570,408 to Gibson). This technology consists of assembling the tubular stocks of starting diameter, heating up to the temperature of their material softening and drawing with compression in order to obtain a required form of the cross-section of the product. The walls of neighboring channels, which cross size is significantly smaller than a starting cross size of the stocks as a result of stretching and can reach a submicron level, become spliced (i.e., fused). The mentioned similarity of technologies of producing an anti-scatter raster and a monolithic X-ray lens does not mean that their technical principle is closely related. In an X-ray lens transporting of a radiation is based on the usage of an effect of multiple total external reflection from the interior side of the walls of the channels. Therefore they are designed to provide a capability of such a reflection. In the suggested anti-scatter raster, vice versa, a situation, when an effective component of a radiation passes immediately from the input to the output of a channel and a reflection of this radiation from the walls of the channels has interference affect on the raster coefficients, is ideal. A total absorption of a radiation by the walls of the channels without reflection from them is desirable for a secondary radiation. Following the condition 2d/h greater than xcex8c, (at this condition no more than one reflection is possible) provides the lack of multiple reflection of a radiation at its transporting along the channel. In the given inequality a critical angle of total external reflection is the following: xcex8c=hxcfx89p/E, where h is a Planck constant, xcfx89p is a plasma frequency for the material of the walls of the channels, E is energy of photon of radiation. Namely, for glass xcex8c [radian]=30/E[eV]. At E=17 keV xcex8c is on the order of 1.8xc3x9710xe2x88x923 radian. When an anti-scatter raster is used, it is placed between the object under study and the detecting device. The anti-scatter raster of the first embodiment is not focused and does not provide the identical conditions for passing an effective component of an X-rays for all channels. These conditions are the best for the channels of the central zone of the raster. Photons of a primary radiation, deflected from the direction to the geometrical center of a raster aperture at the angle exceeding the quantity inverse to the aspect relationship, cannot pass through the raster. Therefore the raster should he distanced well away from the source of a primary radiation, where a peripheral zone of the raster is still transparent for a primary radiation, passed through the object under study. Taking into account the above, if an anti-scatter raster is used according to the first embodiment there is no point in obtaining high values of an aspect relationship. Nevertheless if an anti-scatter raster is made as a cellular structure (and not a slot one), it provides good selection of a secondary radiation. The above is illustrated by FIG. 2 and FIG. 3. FIG. 2 depicts possible trajectories of photons of a secondary scattered radiation from one of the point of the object under study, when they can pass through a slot channel 1 to the detecting device. FIG. 3 depicts a raster, where a slot channel is substituted with channels-cells 1, having the same total size in a FIG. 1. The photons of a secondary radiation, reached a detecting device, have trajectories with smaller angle then shown on the FIG. 2, therefor a possibility of photons reaching the detecting device is reduced. An amount of suppressing of a secondary radiation has same proportional value to the relation of a length of a slot channel to its width, if the channels have a cross-section size equal to a width of a slot channel. It is possible to decrease an aspect relationship without the deterioration of a selection of a secondary radiation in comparison with the known slot anti-scatter grids in the limits of value, defined by this achievable gain. Owing to this fact a distance between a raster and a source of primary X-rays can be reduced, and if this distance remains the same a transparency of peripheral channels for a primary radiation in comparison with the known anti-scatter grids of the same sizes can be increased. The described anti-scatter raster according to the first embodiment is the simplest to produce. Producing a raster according to the second embodiment (FIG. 4) requires additional operations of shaping to make its channels 1 narrow simultaneously with a raster narrowing as from inputs to outputs of the channels. These operations can be carried out by using forming devices with spherical planes, a xe2x80x9cplanexe2x80x9d raster according to the first variant is fastened between, and the raster is simultaneously heated up to the temperature of the material softening. The input and output planes 4 and 5 are placed on the forming device. The walls of the channels have a form of side surfaces of truncated cones or truncated pyramids with a common top, coinciding with the center of the concentric spherical planes 4 and 5. A focused anti-scatter raster according to the second embodiment provides good selection of a secondary radiation in combination with a steady transparency along the whole aperture for a primary radiation. Therefore a choice of an aspect relationship in it is not limited by the factors, being taken into account when a raster according to the first embodiment is designed, and it can be realized with full usage of possibilities. Owing to this fact a degree of suppressing a secondary radiation can be very high with corresponding increasing of an image contrast. As the advantages, provided by high aspect relationship, can be realized fully in the raster according to the second embodiment, it is possible to assemble it from polycapillaries, and the sizes of a cross-section of a single channel of a capillary are already very small. In this case before assembling the polycapillaries can be shaped lengthwise to a required narrowing form, thus it is possible to form the raster in pieces rather than as a unit. The input and output apertures of an anti-scatter raster according to the second embodiment are non-planar, what makes it inconvenient in use (namely in storing and delivering). This inconvenience is removed in the design of a raster according to the third embodiment (FIG. 5), the channels 1 have the same form as in the raster according to the second variant, but the planes 6 and 7, their inputs and outputs are planar. Such surfaces can be obtained by cutting a raster according to the second variant from two sides (this raster has a xe2x80x9creservexe2x80x9d of a longitudinal size of a channel). Different longitudinal sizes of the channels, increasing to the periphery, are characteristic for the raster according to the third variant. It can cause unequal losses of radiation intensity in the channels, being at different distances from the central zone of the raster, particularly if the channels are not hollow. However the differences at the sizes of the raster and the focal distance (i.e. a distance from the input of the channels of the central zone of the raster to the source of a primary radiation), being characteristic for practical applications, are small. A form of a cross-section of the channels according to the second and the third embodiments can be the same, as according to the first embodiment (FIG. 6 and FIG. 7). In both embodiments, as well as in the first one, a condition 2d/h greater than xcex8c is followed, and owing to this fact when a radiation is transported along the channels its multiple reflection is lacking. An anti-scatter raster, intended for the use in a scanning system of X-ray diagnostics or flaw detection, can be made as a narrow brick (parallelepiped shaped) (FIG. 8), and the quantity of channels 1 along its larger side is significantly more than in a perpendicular direction (i.e. in direction of a raster moving when scanning is realized). Realizing a high aspect relationship may be not to the purpose if an anti-scatter raster is made as a narrow parallelepiped according to the first embodiment, and taking into account a straight path of its moving and the above said reasons. Therefore such a raster can be made with rather large cross size of the channels, placed in one row along the parallelepiped""s length. If an anti-scatter raster is made as a narrow parallelepiped according to the second and third embodiments, i.e. when they are focused, to realize advantages of this method it is reasonable to move the parallelepiped in an arc of a circle with a center in a focus (i.e. in a point where a source of a primary X-rays is placed) in order to keep an orientation of the longitudinal axes of the channels to the source. In this case it is reasonable to realize a high aspect relationship, and a parallelepiped can be made of several rows of channels as cross sizes of the channels are small. The indexes of a raster in a form of a narrow parallelepiped according to the first embodiment can be improved when a brick of channels is moved in an arc of a circle, as the conditions of a primary radiation passing through a raster will be equal in all points of a trajectory, it is moved along. The channels for radiation transporting, according to all three embodiments, can be made of glass; lead glass is preferable. It is possible to make the channels of lead or other heavy metals as well. An advantage of these materials is that in this case the walls of the channels can be very thin. For instance, if a raster is used for mammography, when an X-ray tube with molybdenum anode is used as a source (quanta photon energy is E=1.74 keV), the walls may be as thick as 10-20 xcexcm. It increases the transparency of a raster and makes possible to keep a raster immovable as a survey is realized, as the shadows of such thin walls are not practically visible on the film. Sometimes, in dependence of the energies to be used, it can be reasonable to make the walls of the channels of light metals (for instance aluminum) or dielectrics. It can be realized when a radiation fall on the wall made of heavy metal causes hard secondary radiation, which can reaches an X-ray film or a detector. When light metals and dielectrics are used an emerging secondary radiation is soft and it is absorbed in a air layer of some centimeters thickness. It is desirable to make the channels hollow in most cases, for instance a raster can be made of glass mono or polycapillaries. In this case a transparency of a raster (a relationship of an open area of the cross-section of a raster to a total area) can reach 80%. Such a high transparency makes possible to decrease an irradiation dose of a patient. A technology of such raster producing is complicated. Therefore sometimes in order to simplify producing it is possible to make a raster with the channels filled with an organic material (for instance, polyvinyl chloride) or light metal, which slightly absorbs a primary radiation, carrying the information about the object under study. The filled channels are less affected by uncontrollable deformations during forming of a raster. Experimental researches of a raster, produced according to the suggested inventions, show that it is real to produce a device with cross sizes of 20xc3x9720 cm2 and more. When the raster is made of lead glass, attained transmission factor for an effective component of a radiation is on the order of 0.85. If an aspect relationship H/d is on the order of 60 to 100, attenuation of a scattered radiation on the output of a raster reaches 100 to 1000 times. An image contrast increasing in 4 to 7 times (for instance, in mammographic researches) corresponds to such indexes. Thus the usage of a raster makes possible to decrease an irradiation dose in 3 to 5 times. While the invention has been described in terms of a single preferred embodiment, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the appended claims. |
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abstract | Techniques for vaporizing and handling a vaporized metallic element or metallic element salt with a heated inert carrier gas for further processing. The vaporized metallic element or salt is carried by an inert carrier gas heated to the same temperature as the vaporizing temperature to a heated processing chamber. The metal or salt vapor may be ionized (and implanted) or deposited on substrates. Apparatus for accomplishing these techniques, which include carrier gas heating chambers and heated processing chambers are also provided. |
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063114763 | abstract | A solar thermal engine for propelling and powering a craft. The solar thermal engine includes a housing, a propellant annulus, a plurality of static power converters and an electrical energy storage device. The housing has an optical cavity for receiving a beam of concentrated sunlight and converting the beam into ambient thermal energy. The propellant annulus is coupled to the housing and is selectively operable in a heating mode wherein the propellant annulus transmits at least a first portion of the ambient thermal energy to heat a flow of propellant. The plurality of static power converters are coupled to the housing and receive the first portion of the ambient thermal energy when the propellant annulus is not operated in the heating mode. The plurality of static power converters employ the first portion of the ambient thermal energy to generate electrical energy. The electrical energy storage device is coupled to the plurality of static power converters and receives and stores the electrical energy generated by the plurality of static power converters. A method for propelling and powering a craft is also provided. |
054886422 | claims | 1. An auxiliary system for cooling water from a spent fuel pool of a nuclear power generating plant comprising a heat exchanger for transferring heat from said water to be cooled while said water is flowing on a water side of a heat transfer surface of the heat exchanger to air flowing on an air side of said heat transfer surface of the heat exchanger and including means for spraying fine droplets of water into the air flowing on the air side of the heat transfer surface of the heat exchanger to enhance the effectiveness of the heat exchanger. 2. The auxiliary cooling system of claim 1 wherein the means for spraying fine droplets of water comprises a plurality of spray nozzles. 3. The auxiliary cooling system of claim 1 wherein the heat transfer surface is a compact plate-fin surface. 4. The auxiliary cooling system of claim 3 wherein the plate-fin surface has strip fins. 5. The auxiliary cooling system of claim 1 wherein the heat exchanger is a tube-fin type heat exchanger. 6. In a nuclear power generating plant having a pool of water for cooling spent fuel assemblies and a water to water heat exchanger for maintaining the water in said pool of water at a desired cooling temperature, an auxiliary heat exchanger which employs a stream of air carrying a mist of fine water droplets as a coolant medium for indirect heat exchange with water circulated through said auxiliary heat exchanger from said pool of cooling water. 7. The apparatus of claim 6 wherein the auxiliary heat exchanger is a plate-fin heat exchanger. 8. The apparatus of claim 7 wherein the fins of said plate-fin heat exchanger are strip fins. 9. The apparatus of claim 6 including a plurality of nozzles selectively operable in groups to spray water as a mist of fine water droplets into the stream of air. 10. The apparatus of claim 6 including a plurality of nozzles operable to spray water as a mist of droplets having a mean diameter of 250 microns or less. 11. The apparatus of claim 10 wherein the nozzles are operable to spray water as a mist of droplets having a mean diameter of 100 microns or less. 12. The apparatus of claim 10 wherein the nozzles are operable to spray water as a mist of droplets having a mean diameter of about 50 microns. |
claims | The ornamental design for an X-ray collimator, as shown and described. |
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summary | ||
claims | 1. An electron exit window foil for use with a high performance electron beam generator operating in a corrosive environment, the electron exit window foil comprising a sandwich structure havinga film of Ti,a first layer of a material having a higher thermal conductivity than Ti,a flexible second layer of a material being able to protect said film from said corrosive environment, wherein the second layer is facing the corrosive environment, andat least one adhesive barrier coating between the Ti film and the first layer and/or the second layer,wherein said at least one adhesive barrier coating is a layer of Al2O or ZrO2. 2. The electron exit window foil according to claim 1, wherein the first layer is arranged between the film and the second layer. 3. The electron exit window foil according to claim 1, wherein the Ti film is arranged between the first layer and the second layer. 4. The electron exit window foil according to claim 1, wherein the material of the first layer has a ratio between thermal conductivity and density higher than that of Ti. 5. The electron exit window foil according to claim 1, wherein the first layer is selected from the group consisting of Al, Cu, Ag, Au, and Mo. 6. The electron exit window foil according to claim 1, wherein the second layer is selected from the group consisting of Al2O3, Zr, Ta, and Nb. 7. The electron exit window foil according to claim 1, wherein said at least one adhesive barrier coating is between the first layer and the second layer, and wherein said layers are arranged on the same side of the Ti film. 8. The electron exit window foil according to claim 1, wherein said at least one adhesive barrier coating has a thickness between 1 and 150 nm. 9. An electron beam generator configured to operate in a corrosive environment, comprisinga body housing and protecting an assembly generating and shaping the electron beam, anda support carrying components relating to the output of the electron beam, said support comprising an electron exit window foil according to claim 1. 10. A method for providing an electron exit window foil for use with a high performance electron beam generator operating in a corrosive environment, said method comprising:providing a film of Ti,providing a first layer of a material having a higher thermal conductivity than Ti onto a first side of said film,providing a flexible second layer of a material being able to protect said film from said corrosive environment, wherein the second layer is facing the corrosive environment, andproviding at least one adhesive barrier coating between the Ti film and the first layer and/or the second layer,wherein said at least one adhesive barrier coating is a layer of Al2O or ZrO2. 11. The method of claim 10, wherein the providing of the flexible second layer comprises arranging said flexible second layer onto a second side of said film. 12. The method according to claim 10, wherein the providing of the flexible second layer comprises arranging said flexible second layer onto said first layer. 13. The method according to claim 10, wherein at least one of the providing of the first layer or the providing of the flexible second layer is preceded by providing an adhesive coating onto said film. 14. A method for providing a high performance electron beam device, comprising:providing a foil-frame subassembly comprising the steps of:attaching a film of Ti onto a frame, andprocessing said film by providing a first layer of a material having a higher thermal conductivity than Ti onto a first side of said film, and providing a flexible second layer of a material being able to protect said film from said corrosive environment, wherein the second layer is facing the corrosive environment andattaching said foil-frame subassembly to a tube housing of an electron beam device for sealing said electron beam device. |
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description | This application is a continuation of U.S. patent application Ser. No. 15/344,243, filed Nov. 4, 2016, now U.S. Pat. No. 9,741,460 B2, which is a continuation of U.S. patent application Ser. No. 14/942,643, filed Nov. 16, 2015, now U.S. Pat. No. 9,700,922 B2, which is a continuation of U.S. patent application Ser. No. 14/129,504 filed Dec. 26, 2013, now U.S. Pat. No. 9,190,181 B2, which was published on Aug. 7, 2014 under Publication No. US2014/0221722 A1, which is United States National Phase of PCT Patent Application No. US2012/045084 filed on Jun. 29, 2012, which was published on Jan. 3, 2013 under Publication No. WO 2013/003796 A1, which claims priority to U.S. Provisional Patent Application No. 61/502,557 filed Jun. 29, 2011, which are incorporated herein by reference. Nuclear reactors generate 19 percent of the electricity in the U.S., and this process generates high-level radioactive waste in the form of uranium oxide or mixed oxide fuels. Approximately 1000 m3 (6200 bbl) of high-level waste is produced each year from commercial reactors in the U.S., and additional material is generated by military operations. Europe is also heavily invested in nuclear power (e.g. more than three-fourths of the electricity in France is generated by nuclear reactors), and other countries worldwide have started to aggressively pursue nuclear energy to power their growing economies. As a result, the current rate of nuclear waste generation is approximately 10,000 m3/yr, and the amount of radioactive waste being generated worldwide is expected to increase significantly. Yet, there are no safe, reliable ways to dispose of nuclear waste on site, that is, at the source of the waste's generation. This waste includes but is not limited to spent nuclear fuel from nuclear reactors, high-level waste from the reprocessing of spent nuclear fuel, transuranic waste mainly from defense programs, and uranium mill tailings from the mining and milling of uranium ore. High-level nuclear waste is currently stored at the reactor where it was generated. The only serious options for disposal being considered are to place the waste in low permeability geologic formations, like tight rock or clay. The current approach for disposal of radioactive waste is not without problems. Congress has mandated a 10,000-year period of isolation, but it is difficult to guarantee that waste at the shallow depths of current repositories will remain isolated from the biosphere, or human intervention, for even a fraction of this time. Yucca Mountain, a 300-m-deep facility near Las Vegas, is the only U.S. option for high-level waste disposal. This facility has been scrutinized for 20 years, and even after a $50 B expenditure the earliest it could open is 2017. Considerable political opposition by Congress, the state of Nevada and others may delay opening even further. For example, Congress did not provide any funding for development of the site in the 2011 federal budget. Significant uncertainty exists about the feasibility of waste placed at a depth of 300 m remaining isolated from the biosphere for 10,000 years, and this uncertainty is the basis for much of the opposition to Yucca Mountain. Even if Yucca Mountain does open, all its capacity has been allocated and options for additional capacity are being considered. The politics involved in finding permanent disposal sites is, at best, difficult and, at worst, intractable. Because the waste remains radioactive for a very long time, no one wants this waste traveling through their “backyard” on its way to a permanent disposal site or in their “backyard” as the disposal site. As politicians and the public continue to debate the issue, the waste remains temporarily stored on site in ways that are arguably far less safe than any proposed permanent disposal solution. For example, nuclear reactors temporarily store the waste on site in water pools. The devastating earthquake and tsunami in northeast Japan, which knocked out power sources and cooling systems at Tokyo Electric Power Co's Fukushima Daiichi plant, demonstrates how tenuous and potentially dangerous this storage practice really is. Therefore, a need exits for a safe, reliable method of disposing nuclear waste on site and one that could achieve the 10,000 year period of isolation required by Congress and sought by other countries. A system and method according to this invention involves storing nuclear waste or hazardous waste in hydraulic fractures driven by gravity, a process referred to herein as “gravity fracturing.” For the purposes of this disclosure, nuclear or radioactive waste is considered a hazardous waste although in the environmental industry radioactive waste is often not labeled as “hazardous waste.” The method creates a dense fluid containing waste, introduces the dense fluid into a fracture, and extends the fracture downward until it becomes long enough to propagate independently. The fracture will continue to propagate downward to great depth, permanently isolating the waste. Storing solid wastes by mixing the wastes with fluids and injecting them into hydraulic fractures is a well-known technology in the petroleum industry. Nuclear waste was injected into hydraulic fractures at Oak Ridge in the 1960s. The essence of the invention differs from conventional hydraulic fracturing techniques in that it uses fracturing fluid heavier than the surrounding rock. This difference is fundamental because it allows hydraulic fractures to propagate downward (rather than horizontally) and carry wastes by gravity instead of by pumping. More specifically, the method of disposing nuclear waste and other hazardous waste includes the steps of blending the waste with water or other fluid and a weighting material to make a dense fluid or slurry of a predetermined density, temperature and viscosity; and injecting the dense fluid or slurry—at a predetermined pressure and/or rate into a well so that the fluid or slurry enters the strata at a predetermined depth and continues to travel downward through the strata until the fluid or slurry, becomes immobilized. Prior to the blending step, the waste, if in solid form, may be ground into particles of a predetermined size. The pressurized blended mixture cracks and dilates the rock structure, which is preferably a stable, low permeability rock structure such as many igneous and metamorphic rocks as well as some sedimentary rocks. (Initially, propping the fracture is avoided). Because the dense fluid has a density greater than that of the rock, the fluid or slurry has an absolute tendency to travel downward by gravity (until the density relationship changes or other mechanics arrest the downward travel) and remain far below the earth's surface. The dense fluid may include water, oil, gel or any fluid suitable for providing the required viscosity and density. The well is preferably drilled at and on the site which generates the nuclear waste or other hazardous waste, thereby eliminating the need to transport the waste off-site and to the disposal site. The well includes a work string or tubing for receiving the blended fluid, waste and weighting material; a packer; and a cemented steel casing with perforations located at or about the predetermined depth. The predetermined depth is preferably in a range of about 10,000 to 30,000 feet (about 3,000 to 9,000 meters) but it can be shallower or deeper depending upon rock properties and drilling limitations. The weighting material may be other nuclear waste (including, for example, radionuclides such as uranium), other hazardous waste or a metal such as bismuth, lead, or iron in order to add weight to the primary waste which is being disposed. Metals or alloys that are in liquid phase at the temperature and pressure encountered in the subsurface are particularly suitable as a weighting material. The work string may be pulled for routine cleaning or replacement. The blender used to blend the water, waste and weighting material is preferably shielded, as is the pumping unit (e.g., a pumping truck) used to pump the mixture at pressure into the well. Hydraulic fractures are created when the pressure in a fluid-filled crack causes the material at the crack tip to fail. The fracture advances and fluid flows forward to fill the newly created space. Hydraulic fractures are commonly created by using a pump to inject fluid into a well, but this is by no means the only occurrence. Geologic examples are well known in which hydraulic fractures grow upward through the Earth's crust because the fractures are filled with liquid lighter than their enveloping rock. A dike filled with magma that propagates upward to feed a volcanic eruption is one example of a hydraulic fracture propagating by gravity. A system and method according to this invention involves propagating hydraulic fractures downward by filling the fractures with dense fluid containing waste. Propagation occurs when the pressure in the fracture creates a stress intensity that exceeds the toughness or strength of the rock. Referring to FIGS. 1 to 3, an open borehole is created and filled with the dense fluid until the pressure at the bottom is sufficient to create a fracture (FIG. 3 at “a”). A similar fracturing process occurs during overbalanced drilling when the mud weight is too great and causes circulation to be lost by initiating a fracture and causing it to grow away from the borehole. Fluid will flow into the fracture and the level of fluid in the well will drop (FIG. 3 at “b”). However, the fracture is expected to advance faster than the rate of drop of fluid level in the well, so the overall height from the tip of the fracture to the top of the fluid column in the well lengthens. This increases the driving pressure and furthers downward propagation as the fluid in the wellbore drains by gravity into the fracture (FIG. 3 at “c”). The vertical span of the fracture continuously increases, causing the pressure at the bottom of the fracture to increase and ensuring continued downward propagation, even after all the liquid has drained from the well into the fracture (FIG. 3 at “d”). The pressure distribution causes the lower part of the fracture to bulge open and the upper part to pinch shut. A residual coating of fluid will be left behind when the fracture closes, and this will diminish the volume of fluid in the fracture. Eventually the original fluid will be spread as a thin coating on the fracture wall, extending from the bottom of the borehole to great depth. In the case of slurry, the fracture may be propped if the liquid leaks off into the rock. The process is repeated by putting additional fluid into the well. This will create a new fracture that will follow the path of the earlier one (FIG. 3 at “e”). The additional fluid reaches an even greater depth than the original batch. The maximum depth that can be reached by dense fluids is unclear, but it could exceed tens of kilometers. A method of disposing nuclear waste and other hazardous waste practiced according to this invention, therefore, effectively removes the waste from exposure to human activities at a time scale relevant to both societal actions and the half-lives of many hazardous radionuclides. The method includes the steps of blending the waste with materials suitable for creating a dense fluid or slurry which has a predetermined density and viscosity; and injecting the dense fluid at a predetermined pressure or rate into a well so that the dense fluid enters the strata at a predetermined depth and continues to travel downward through the strata until its flow stops, for example, because the solid-to-liquid ratio is too high to allow flow. Propagation may also stop when a sufficient amount of the dense fluid or fluid/slurry has been spread as a film or residue over the upper closed portion of the fracture. Oil, gel or any fluid suitable for providing the required viscosity and density may be used Weighting material adds density to the primary waste which may be other types of nuclear waste, other hazardous waste or a metal such as, but not limited to, bismuth, lead, iron, copper, or low melting point metals or alloys (e.g., mercury, woods metal, indalloy 15, gallium) that could mix with and possibly dissolve or amalgamate high-level waste material. The low-melting-point alloys are a liquid under the expected pressure and temperature conditions at the bottom of the injection well. Solid compounds such as metals used for weighting material may be mixed with a high-shear-strength liquid, including polymer gels that may be crosslinked, or inorganic gels that may formed by hydrating clay minerals, to create a dense slurry. Prior to the blending step, the waste, if in solid form, may be ground to a predetermined size. The pressurized dense fluid creates a vertical fracture or crack in the rock structure. The dense fluid enters the crack and serves to prop the rock structure. The rock structure is preferably a stable, low permeability rock formation, of the kind that nuclear reactors are typically built over and upon. Because of the weighting material, the density of the dense fluid is greater than that of the rock and this causes an absolute tendency for the fluid to travel downward until it becomes immobilized. If the density of the dense fluid is exactly equal to that of the rock, the dense fluid may be unable to overcome the rock fracture toughness. This is required for fracture propagation, hence the density should be somewhat higher to ensure the fracture growth. How much higher depends upon the fracture toughness magnitude, fluid properties, and other effects standard in industrial hydraulic fracturing. In general terms, the density of rock increases as depth increases. Therefore, once the fracture propagates, a point can be reached where the density of the dense fluid becomes the same as the density of the rock, thereby limiting any further propagation downward. Eventually, the fracture becomes sub-horizontal and the dense fluid fills the fracture horizontally. This is similar to geological sills and does not hamper the proposed technology as the horizontal part of the growing fracture also allows for safe waste storage. Fracture toughness also increases with depth because it increases with such factors as temperature, pressure and size of the fracture. However, the effect of fracture toughness can be overcome by pressurizing the fracture. For example, and just by way of example the immobilization point may occur at about 2,000 to 50,000 feet (about 600 to 15,000 meters) below the dense fluid's initial entry point into the strata. (The depth can be greater and is mostly constrained by drilling and pumping limitations.) The dense fluid can be monitored by using conventional tracer means to see whether any movement or migration has occurred upward relative to the perforations in the well casing, or it can be monitored using microseismics means to evaluate downward migration below the bottom of the region accessible to the well casing. The well is preferably drilled at and on the site which generates the nuclear waste or other hazardous waste, thereby eliminating the need to transport the waste off-site and to the disposal site. The well also eliminates the need for temporary storage means on site because the waste can be transported directly to the well for immediate permanent disposal. As shown in FIG. 2, the well includes a work string or tubing for receiving the blended water, waste and weighting material; a packer; and a cement casing with perforations located at or about the predetermined depth. The predetermined depth is preferably in a range of about 10,000 to 30,000 feet (about 3,000 to 9,000 meters). The work string may be pulled for routine cleaning or replacement. The blender used to blend the water, waste and weighting material is preferably shielded, as is the pump truck used to pump the dense fluid at pressure into the well (see FIG. 1). Preferred embodiments of a system and method for abyssal sequestration of nuclear waste and other types of hazardous waste have been described and illustrated, but not all possible embodiments. The inventive system and method itself is defined and limited by the following claims. |
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050698605 | description | MODE(S) FOR CARRYING OUT THE INVENTION Illustrated in FIG. 1 is an exemplary nuclear reactor vessel 10 having a plurality of fine motion control rod drives 12 (FMCRD), only one of which is shown. In one exemplary embodiment, there are 205 FMCRDs 12 extending into the vessel 10 through the bottom thereof. Referring also to FIG. 2, an enlarged, sectional view of one of the control rod drives 12 is illustrated. The rod drive 12 includes a tubular housing 14 extending outwardly from the vessel 10 and conventionally secured thereto. The housing 14 is conventionally connected to a manifold 16 which is disposed in flow communication with a scram line or conduit 18 which is conventionally selectively provided with high-pressure water 20 from a conventional high-pressure water accumulator 22 conventionally joined to the scram line 18. Conventionally disposed inside the housing 14 is a conventional ball screw or spindle 24, which in this exemplary embodiment includes conventional right-handed threads 26. The control rod drive 12 includes a longitudinal centerline axis 28, with the housing 12 and spindle 24 being disposed coaxially therewith. A conventional ball nut 30 is positioned over the spindle 24 and is conventionally restrained from rotating therewith so that as the spindle is rotated in a clockwise direction, the ball nut is translated in a downward direction away from the vessel 10, and when the spindle is rotated in a counterclockwise direction, the ball nut 30 is translated in an upward direction toward the vessel 10. A conventional hollow, elongate piston 32 is disposed coaxially with the spindle 24 and includes a conical base end 34 which rests on the ball nut 30, and a tip end 36 extending through a central aperture 38 in the outer end of the housing 14 into the vessel 10. The tip end 36 is conventionally coupled to a respective control rod 40 by a bayonet coupling, for example. The spindle 24 extends downwardly from the manifold 16 through a conventional electrical motor 42 which selectively rotates the spindle 24 in either the clockwise direction or counterclockwise direction. The motor 42 is electrically connected to a conventional control 44 by a conventional electrical line 46 for selectively controlling operation of the motor 42. In accordance with the preferred embodiment of the present invention, the rod drive 12 further includes a rotary lock 48 joined to the motor 42 into which extends the spindle 24, also referred to as an input shaft 24. The rotary lock 48 is electrically joined to the control 44 by a conventional electrical line 50 for selectively locking and unlocking, or releasing, the input shaft 24. FIG. 3 illustrates in more particularity the rotary lock 48. The lock 48 includes an annular tubular stationary housing 52 which is conventionally fixedly secured to the motor 42 by bolts, for example (not shown). The housing 52 receives an end portion of the shaft 24 and is coaxial therewith. At least one arm 54 and preferably a pair of arms 54, including a first arm designated 54 and a second arm designated 54b, are fixedly joined to the shaft 24 in the housing 52. More specifically, each of the arms 54 includes a proximal end 56,56b fixedly joined to the shaft 24 by an integral hub 57 prevented from rotating by a conventional key 57b. The arms each further include a distal end 58,58b extending radially outwardly from the shaft 24 and substantially perpendicular thereto; a top surface 60,60b; an opposite, bottom surface 62,62b; and a side surface 64,64b as shown in FIG. 4. At least one pin 66 and preferably a plurality of pins 66 each includes a proximal portion 68 joined to the housing 52 as described in more detail below for circumferential restraint relative to the shaft 24 and centerline 28. The pin 66 also includes an intermediate portion 70, and the pin 66 is positionable in a withdrawn position 72 as shown in dashed line in FIG. 3 away from the arm top surface 60 for allowing the arm 54 to rotate with the shaft 24 without obstruction from the pin 66. The pin 66 is also positionable in a deployed position 74 as shown in solid line in FIG. 3, for allowing the pin intermediate portion 70 to contact the arm side surface 64 at the arm distal end 58 for preventing rotation of the arm 54 and shaft 24. Means are provided for selectively positioning the pins 66 in the withdrawn and deployed positions as indicated generally in FIG. 3 at 76. More specifically, the pin positioning means 76 includes means for urging the pins 66 into one of the withdrawn and deployed positions. And, an actuator, such as for example an electric solenoid 78, is energizable for moving the pins 66 into the other of the withdrawn and deployed positions. The solenoid 78 is a conventional coil of wire suitably mounted around the shaft 24 to a support plate 80 conventionally fixedly secured in the housing 52. A magnetic plunger 82 is provided for magnetic cooperation with the solenoid 78 and is slidably disposed in the solenoid 78 and joined to the pins 66 for moving the pins 66 when the solenoid is energized. The plunger 82 is preferably tubular and includes a central bore 84 disposed coaxially about the shaft 24 for longitudinal movement relative thereto, an annular base plate 86 having a diameter D.sub.1 slightly less than the inside diameter of the housing 52 for allowing longitudinal movement therein. The base plate 86 is disposed at one end of the plunger 82 adjacent to the arm top surfaces 60,60b. The plunger 82 also includes an intermediate portion 88 extending from the base plate 86, and is preferably integral therewith, and through the solenoid 78. An annular flange 90 is conventionally fixedly joined to the intermediate portion 88, for example by screws or by being formed integrally therewith, at an opposite end of the plunger 82. The base plate 86 includes a plurality of circumferentially spaced guide holes 92, as additionally shown in FIG. 5, which face the arm top surfaces 60,60b with each of the guide holes 92 including a tubular pin recess 94 extending outwardly away from the arms 54,54b. The pins 66 extend from the recesses 94 through the plunger guide holes 92 toward the arms 54,54b with each pin 66 including an annular flange 96 fixedly joined to the pin proximal portion 68 and disposed in the pin recess 94. The diameter of the pin recess is predeterminedly greater than the diameter of the guide hole 92, with the diameter of the flange 96 being predeterminedly less than the diameter of the guide recess 94 to allow the pin 66 to translate longitudinally in the recess 94 and through the guide hole 92, with the flange 96 preventing the pin 66 from being ejected from the recess 94 through the guide hole 92. A plurality of compression pin springs 98 are disposed in respective ones of the pin recesses 94 for providing a spring force on the pin flange 96 for urging the pin flange 96 against the guide hole 92 for positioning the pin intermediate portion 70 outwardly from the plunger base plate 86 toward the arms 54 in an extended position 100 as illustrated in solid line for the right pin illustrated in FIG. 3. The pin extended position 100 is effective for allowing the pin intermediate portion 70 to contact the arm side surface 64,64b for example, when the plunger 82 and the pins 66 are disposed in the deployed position 74. In the preferred embodiment, the means for urging the pins 66 to one of the withdrawn and deployed positions includes the shaft 24 and the plunger 82 being oriented vertically with the plunger 82 being disposed above the shaft arms 54 for allowing gravity to pull the plunger 82 toward the arms 54 in the deployed position when the solenoid 78 is de-energized. It is possible that when the plunger 82 is deployed, one of the pins 66 will contact an arm top surface 60 as shown in the left-hand side of FIG. 3. Without the pin 66 being disposed in the recess 94, the pin 66 would merely rest upon the arm 54 and would not lock rotation of the shaft 24 until the shaft 24 begins to rotate, allowing the pin 66 to drop between adjacent ones of the shafts 54. Accordingly, by spring mounting the pins 66 in the plunger 82, if one of the pins 66 contacts one of the arm top surfaces 60 in the deployed position 74, that one pin 66 will be pressed into the recess 94 against the spring 98, thusly positioning the pin 66 in a retracted position 102 as illustrated in the left-hand side of FIG. 3. The longitudinal length of the pin recess 94, therefore, is preselected for allowing the pin 66 to be urged into the recess 94 against the spring 98 in the retracted position 102 when a pin distal portion 104 contacts the arm top surface 60 in the deployed position. The plunger 82 is positionable in the deployed and withdrawn positions 74,72 for deploying and withdrawing the pin 66 by de-energizing and energizing the solenoid 78, respectively. The solenoid 78 is predeterminedly sized so that when it is energized it is effective for magnetically drawing upwards the plunger 82 away from the arms 54 into the withdrawn position 72. When the solenoid 78 is de-energized, gravity simply drops the plunger 82 into the deployed position 74. In an alternate embodiment, a plunger compression spring 106 may be disposed around the plunger intermediate portion 88 and between the base plate 86 and the solenoid 78 for urging the plunger base plate 86, including the plurality of pins 66, toward the arms 54 in the deployed position 74 when the solenoid is de-energized. The spring 106 may be used in addition to gravity or as an alternate to gravity where the entire rotary lock 48 is oriented, for example upside down to that shown in FIG. 3. The solenoid 78 is predeterminedly sized for being effective for disposing the plunger 82 in the withdrawn position 72 compressing the plunger compression spring 106 when the solenoid 78 is energized. The plunger intermediate portion 88 in this embodiment has a diameter D.sub.2 which is substantially less than the diameter D.sub.1 of the base plate 86 and larger than the diameter of the shaft 24 for providing room for positioning the spring 106. The spring 106 is predeterminedly initially compressed between the support plate 80 and a cover plate 108 suitably fixedly secured to the top end of the base plate 86, by screws for example. The cover plate 108 provides access for assembling the springs 98 and the pins 66 into the recesses 94 and guide holes 92. As illustrated in more particularity in FIG. 5, the plunger base plate 86 includes an outer perimeter 110 which includes at least one and preferably two longitudinal guide recesses 112 therein. The housing 52 includes a corresponding pair of longitudinal guide rails 114 suitably fixedly joined to the housing 52, by brazing for example, which extend parallel to and into the complementary guide recesses 112. The guide rails 114 are slidably disposed in the recesses 114 for allowing the plunger 86 to move longitudinally along the shaft 24 between the deployed and withdrawn positions while circumferentially restraining the plunger 86, and the pins 66 disposed therein, from rotating about the shaft 24 and longitudinal axis 28 when at least one of the pins 66 contacts the arm side surface 64,64b in the deployed position. Torque transmitted from the arms 54,54b is transferred through the at least one pin 66 contacting a respective arm and in turn is transferred through the plunger base plate 86 and to the stationary housing 52 through the guide recess 112 and rails 114 for providing a positive lock for preventing rotation of the shaft 24. By predeterminedly selecting the number of pins 66 and the widths of the pins 66 and arms 54,54b as measured in the circumferential direction, the amount of maximum rotation of the shaft 24 upon deployment of the pins 66 may be minimized. For example, in the preferred embodiment illustrated in FIG. 4, there are eleven equiangularly spaced pins 66 spaced 36.degree. apart. The arms 54,54b are selected for having a width which is smaller than the spacing between adjacent pins 66 to ensure that the pins 66 may be deployed without interference with the arm top surfaces 60. Accordingly, when the pins 66 are deployed, the arms 54,54b are allowed to rotate only until one of the arms 54 contacts an adjacent pin 66 and thus prevents further rotation of the shaft 24. In the preferred embodiment, each of the arms 54,54b includes an arcuate, preferably semicircular, notch 116 disposed in respective side surfaces 64,64b of the arms and aligned with the pins 66, against which pins 66 the notches 116 contact when the pins 66 are deployed. The notch 116 is preferred for distributing the load between the pins 66 applied to the arms 54,54b. The pins 66 are predeterminedly sized for particular torque-resisting applications to ensure that they are adequate for accommodating reaction torque without bending or misalignment thereof. The pins 66 may be simply cantilevered from the guide holes 92 with the pin distal portions 104 suitably being unsupported in space. However, in accordance with a preferred embodiment of the present invention, an annular restraint plate 118, as illustrated in FIG. 3, is predeterminedly spaced from the bottom surfaces 62,62b of the arms 54 and suitably fixedly connected to the housing 52, by brazing or welding for example. The restraint plate 118 includes a plurality of circumferentially spaced restraint holes 120, also shown in FIG. 4, which are longitudinally, coaxially aligned with respective ones of the pins 66 to ensure that the pins 66 may be longitudinally deployed from the guide holes 92 and into respective, opposing ones of the restraint holes 120. The pins are predeterminedly longitudinally sized so that the distal portions 104 are disposed into the restraint holes 120 when the pins 66 and the plunger 82 are positioned in the deployed position 74. The guide recesses 112 and rails 114, as illustrated in FIGS. 3 and 5, are predeterminedly positioned for longitudinally aligning the pins 66 with respective restraint holes 120. Accordingly, the guide rails 114 allow the plunger 82 to translate longitudinally and parallel to the centerline 28; restrain circumferential rotation of the plunger 82 for accommodating the torque transmitted from the shaft 24 through the pins 66; and longitudinally align the pins 66 with respective restraint holes 120 to ensure that the pins 66 are received in the respective restraint holes 120 when in the deployed position 74. By circumferentially restraining the arms 54,54b from rotation using the plunger 82 and additionally the restraint plate 118, relatively large torque from the shaft 24 may be accommodated since the pins 66 are supported at both ends. Although one arm 54 may be used for restraining and locking rotation of the shaft 24 against a respective pin 66, the two arms 54,54b described above are preferred. More specifically, by using two arms 54,54b, the arms, pins 66, and restraint holes 120, may be predeterminedly configured so that at least one of the pair of arms 54,54b is always positioned between adjacent ones of the pins 66 for allowing the distal portions 104 of the adjacent ones of the pins 66 to be positioned in respective restraint holes 120 in the deployed position without obstruction by the at least one arm. For example, where the pins 66 in the restraint holes 120 are equiangularly spaced from each other, the second arm 54b is disposed obliquely to the first arm 54 at an angle A as shown in FIG. 4 for allowing the adjacent pins 66 to enter the respective restaint holes 120 without obstruction by the at least one arm 54b. As shown in FIG. 4, for example, the second arm 54b is positioned circumferentially between adjacent pins 66 whereas the first arm 54 blocks one of the pins 66 from entering its respective restraint hole 120. Alternatively, when the first arm 54 is positioned between adjacent pins 66 to prevent obstruction between the pins 66 and the restraint holes 120, the second arm 54b would be positioned longitudinally in line with one of the pins 66, thereby obstructing the pin 66 from entering its complementary restraint hole 120 (not shown). Various alternate arrangements may be used for ensuring that at least one arm 54 is initially positioned between adjacent pins 66 at the time of deployment for locking the shaft 24. The pins 66 and corresponding restraint holes 120 may be unequally circumferentially angularly spaced and one, two or more arms 54 may be utilized. It is preferred that at least one arm 54 is positioned between adjacent pins upon deployment to ensure a positive lock between the arm 54 and one of the pins 66 with a minimum amount of initial rotation of the shaft 24 before locking occurs. In an alternate embodiment, one arm 54 could be used, for example, the first arm 54 illustrated in FIG. 4, which could initially be circumferentially aligned with one of the pins 66 for preventing its deployment from the recess 94. In such an embodiment, when the plunger 82 is in the deployed position 74, all of the pins 66 will be received in their respective restraint holes 120 except for the one pin 66 which would be obstructed by the arm 54 as illustrated in FIGS. 3 and 4. In such a situation, upon initial rotation of the shaft 24 and the arm 54 in the backflow occurrence described above, the shaft 24 would rotate until the arm 54 contacts the next adjacent pin 66 disposed in its deployed position which would then stop the arm 54 and the shaft 24. Such initial rotation is equal to about the angular spacing between adjacent pins 66 which allows momentum to build up in the arm 54 prior to contacting the next adjacent pin 66. Although this momentum may be suitably accommodated by predeterminedly sized pins 66 in respective restraint holes 120, for example, the pins 66 may be made smaller where the buildup in momentum may be reduced. By utilizing the two arms 54,54b as preferably spaced as described above, the maximum amount of initial rotation of the shaft 24 and the arm 54 is limited to less than the angular spacing between adjacent pins 66. This reduces momentum buildup in the arm 54 which allows for correspondingly smaller pins 66 and related support structure for minimizing the cost and complexity of the rotary lock 48. Illustrated in FIGS. 6 and 8 is an alternate embodiment of a rotary lock 48b in accordance with the present invention. In this embodiment, at least one and preferably two solenoids 78b are fixedly supported by an annular support plate 80b fixedly joined to the housing 52b. A tubular plunger 82b is conventionally fixedly joined to a proximal end 68b of a respective pin 66b by being welded thereto or formed integrally therewith, for example. The tubular plunger 82b includes a central bore 122 and a solenoid compression spring 124 is disposed in the central bore 122 and the solenoid 78b for moving the pin 66b to the deployed position 74 when the solenoid 78b is de-energized. When the solenoid 78b is energized, it is effective for moving the plunger 82b disposed therein and the pin 66b to the withdrawn position 72. Both solenoids 78b in the embodiment illustrated in FIG. 6 are connected through the electrical lines 50 to the control 44 for simultaneous and coordinated actuation. A restraining plate 118b is suitably fixedly connected to the housing 52b and spaced from the bottom surfaces of the shaft arms 54,54b. A respective restraint holes 120b is longitudinally, coaxially aligned with a respective one of the pins 66b, and the pins 66b are predeterminedly sized for positioning distal portions 104b into respective restraint holes 120b when in the deployed position 74. As illustrated in FIG. 7, since only two solenoids 78b with two respective pins 66b are utilized, only two restraint holes 120b are required for receiving the respective pins 66b. In the preferred embodiment, the pins 66b are colinear with a line passing through the centerline axis 28. In this embodiment, the first and second arms 54,54b are preferably disposed obliquely to each other to ensure that at least one of the arms 54, 54b is not initially obstructing deployment of one of the pins 66b. This arrangement minimizes the amount of momentum buildup in the arms 54,54b upon initial rotation of the shaft 24 prior to the arms 54,54b contacting one of the pins 66b for locking the shaft 24. In the event the shaft 24 is initially stopped and obstructs deployment of one of the pins 66b, that pin 66b will merely be urged by the solenoid 78b against the arm top surface 60 without adverse effect. As soon as the shaft 24 rotates sufficiently to unobstruct the pin 66b, the solenoid 78b will complete its travel of the plunger 82b for positioning the pin 66b in its respective restraint hole 120b. In both the embodiments described above, the motor 42 and the solenoids 78,78b are energized simultaneously during normal operation to allow the motor 42 to rotate the shaft 24 for positioning the control rod 40 without obstruction by the rotary lock 48,48b. Upon completion of the desired rotation of the motor 42 and positioning of the control rod 40, the motor 42 is de-energized and stopped, and the solenoid 78,78b is de-energized substantially simultaneously so that the pins 66 engage the restraint plate 118. If the shaft 24 then begins to unintentionally rotate, such as for example by the backflow occurrence described above, the shaft 24 will only undergo a limited rotation until one of the arms 54 contacts a pin 66 and is stopped. The pins 66, as described above, provide a positive lock for the shaft 24 to prevent undesirable rotation thereof, including unintentional withdrawal of the control rod 40 from the reactor vessel 10. The rotary locks 48 as described above provide a positive lock of the shaft 24 to prevent ejection of the control rod 40 from the vessel 10 and allow for relatively simple testing of the rotary locks 48. More specifically, the lock 48 may be simply tested by de-energizing the solenoid 78 for positioning the pins 66 in the deployed position 74 and then energizing the motor 42 to allow the arms 54 to strike a respective pin 66 for preventing further rotation of the shaft 24. Since the motor 42 will be unable to rotate the shaft past the holding pin 66, the motor 42 will stall, which may be conventionally observed by the control 44 for indicating the effective operation of the rotary lock 48. If the rotary lock 48 is unable to prevent rotation of the shaft 24 during testing, the control 44 can provide a suitable indication thereof, which will then result in manual inspection of the rotary lock 48 for correcting any problem that might exist. Accordingly, the above-described rotary lock 48 in accordance with the present invention provides a positive restraint which locks the shaft 24 from unintentional rotation providing an improvement over a conventional friction-type brake which may allow for slippage. Furthermore, the rotary lock 48 may be tested remotely as described above so that access to the environment adjacent to the nuclear reactor vessel 10 is not required. Yet further, the rotary lock 48 may be tested relatively quickly during normal operation of the reactor vessel 10 without requiring down time of the reactor solely for testing purposes. While there have been described herein what are considered to be preferred embodiments of the present invention, other modifications of the invention shall be apparent to those skilled in the art from the teachings herein, and it is, therefore, desired to be secured in the appended claims all such modifications as fall within the true spirit and scope of the invention. Accordingly, what is desired to be secured by Letters Patent of the United States is the invention as defined and differentiated in the following claims: |
claims | 1. An electron beam measurement apparatus, which measures, based on information on an image, patterns formed on a sample, comprising:an electron optical system that has a lens and a deflector and scans a predetermined observation region on the sample with an electron beam emitted from an electron source, a detector for detecting a charged particle secondarily generated from the sample by irradiation with the electron beam, and a means for forming an image including forming a secondary electron image using a secondary electron and a reflective electron image using a reflective electron based on the detected charged particle,wherein the patterns are delineated in a single layer present on a substrate, andfurther including a means for classifying patterns, which are arranged in an image acquired by the irradiation with the electron beam irradiated on the patterns on the sample, into at least one of first and second groups based on the secondary electron image and the reflective electron image. 2. The electron beam measurement apparatus according to claim 1, whereinan image processing parameter or a waveform processing parameter is used for each of the groups to obtain the size of a portion of the pattern included in the group or a position of the contour of the portion of the pattern, wherein the parameter used varies depending on the group. |
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description | The present application claims the benefit of U.S. Provisional Patent Application No. 61/828,017 filed May 28, 2013, and is a continuation-in-part of International Patent Application No. PCT/US13/42070 filed May 21, 2013, which claims of benefit of U.S. Provisional Patent Application No. 61/649,593 filed May 21, 2012, the entireties of which are incorporated herein by reference. The present invention relates nuclear reactors, and more particularly to a passive confine system for use in the event of a loss-of-coolant accident and a reactor shutdown. The containment for a nuclear reactor is defined as the enclosure that provides environmental isolation to the nuclear steam supply system (NSSS) of the plant in which nuclear fission is harnessed to produce pressurized steam. A commercial nuclear reactor is required to be enclosed in a pressure retaining structure which can withstand the temperature and pressure resulting from the most severe accident that can be postulated for the facility. The most severe energy release accidents that can be postulated for a reactor and its containment can generally be of two types. One thermal event of potential risk to the integrity of the containment is the scenario wherein all heat rejection paths from the plant's nuclear steam supply system (NSSS) are lost, forcing the reactor into a “scram.” A station black-out is such an event. The decay heat generated in the reactor must be removed to protect it from an uncontrolled pressure rise. Loss-of-Cooling Accident (LOCA) is another type of thermal event condition in which a breach in the pressure containment boundary of reactor coolant system (RCS) leads to a rapid release of flashing water into the containment space. The reactor coolant (primary coolant), suddenly depressurized, would violently flash resulting in a rapid rise of pressure and temperature in the containment space. The in-containment space is rendered into a mixture of air and steam. LOCA events are usually postulated to occur due to a failure in an RCS system pipe containing the primary coolant water. The immediate consequence of a LOCA is rapid depressurization of the RCS and spillage of large quantities of the primary coolant water until the pressure inside the RCS and in the containment reach equilibrium. Nuclear plants are designed to scram immediately in the wake of the RCS depressurization which suppresses the reactor's criticality and stops the chain reaction. However, the large enthalpy of the primary coolant water spilling from the RCS into the containment and the ongoing generation of decay heat in the core are sources of energy that would cause a spike in the containment pressure which, if sufficiently high, may threaten its pressure retention capacity. More recently, the containment structure has also been called upon by the regulators to withstand the impact from a crashing aircraft. Containment structures have typically been built as massive reinforced concrete domes to withstand the internal pressure from LOCA. Although its thick concrete wall could be capable of withstanding an aircraft impact, it is also unfortunately a good insulator of heat, requiring pumped heat rejection systems (employ heat exchangers and pumps) to reject its unwanted heat to the external environment (to minimize the pressure rise or to remove decay heat). Such heat rejection systems, however, rely on a robust power source (off-site or local diesel generator, for example) to power the pumps. The station black out at Fukushima in the wake of the tsunami is a sobering reminder of the folly of relying on pumps. The above weaknesses in the state-of-the-art call for an improved nuclear reactor containment system. What is needed is an efficient energy expulsion system to bring the internal pressure in the containment in the wake of a LOCA to normal condition in as short a time as possible. To ensure that such a system would render its intended function without fail, it is further desirable that it be gravity operated (i.e., the system does not rely on an available power source to drive any pumps or motors). A passive nuclear reactor cooling system for use in the event of a loss-of-coolant accident (LOCA) and complete reactor shutdown is provided that overcomes the foregoing drawbacks. The cooling system is configured to create a completely passive means to reject the reactor's decay heat without any reliance on and drawbacks of pumps and motors requiring an available electric power supply. In one embodiment, the cooling system relies entirely on gravity and varying fluid densities to extract and induce flow of cooling water through the system which includes a heat exchanger. The cooling system is engineered to passively extract decay heat from the reactor in the event of a LOCA station black out or another postulated accident scenario wherein the normal heat rejection path for the nuclear fuel core is lost such as via a ruptured pipe in the primary coolant piping or other event. In one configuration, the passive cooling system utilizes the reserve cooling water in the reactor well as a vehicle to extract and reject, decay heat from the reactor via a heat exchanger attached to the reactor containment vessel walls. The cooling water flows via gravity in a closed flow loop between the reactor well and the heat exchanger to reject heat through the containment vessel walls to an external heat sink. In one embodiment, the heat sink may be an annular reservoir filled with cooling water that surrounds the containment vessel. In further embodiments, as further described herein, an in-containment auxiliary reservoir (e.g. storage tank) of cooling water may be provided which is fluidly coupled to the reactor well to provide a supplemental source or reserve of cooling water. The cooling system closed flow loop may circulate cooling water between both the reactor well and auxiliary reservoir heat exchanger and the heat exchanger. In one embodiment, a passive reactor cooling system usable after a loss-of-coolant accident includes a containment vessel in thermal communication with a heat sink, a reactor well disposed in the containment vessel, a reactor vessel disposed at least partially in the reactor well, the reactor vessel containing a nuclear fuel core which heats primary coolant in the reactor vessel, a water storage tank disposed in the containment vessel and in fluid communication with the reactor well, the tank containing an inventory of cooling water, and a heat exchanger disposed in the containment vessel, the heat exchanger in fluid communication with the reactor well via a closed flow loop. Following a loss of primary coolant, the tank is configured and operable to flood the reactor well with cooling water which is converted into steam by heat from the fuel core and flows through the closed flow loop to the heat exchanger. In one embodiment, the steam condenses in the heat exchanger forming condensate, and the condensate flows via gravity back to the reactor well. The heat exchanger comprises an array of heat dissipater ducts integrally attached to the containment vessel in one embodiment. In another embodiment, a passive reactor cooling system usable after a loss-of-coolant accident includes a containment vessel in thermal communication with a heat sink, a reactor well disposed in the containment vessel, a reactor vessel disposed at least partially in the reactor well, the reactor vessel containing a nuclear fuel core and primary coolant heated by the fuel core, a water storage tank disposed in the containment vessel and in fluid communication with the reactor well, the tank containing an inventory of cooling water, and a heat exchanger disposed in the containment vessel, the heat exchanger in fluid communication with the reactor well via an atmospheric pressure closed flow loop. Following a loss of primary coolant, the tank is configured and operable to flood the reactor well with cooling water. The cooling water in the flooded reactor well is heated by the fuel core and converted into steam, the steam flows through the Closed flow loop to the heat exchanger and condenses forming condensate, and the condensate flows back to the reactor well. The heat exchanger comprises an array of heat dissipater ducts integrally attached to the containment vessel in one embodiment. A method for passively cooling a nuclear reactor after a loss-of-coolant accident is provided. The method includes: locating a reactor vessel containing a nuclear fuel core and primary coolant in a reactor well disposed inside a containment vessel; at least partially filling a water storage tank fluidly coupled to the reactor well with cooling water; releasing cooling water from the water storage tank into the reactor well; heating the cooling water with the fuel core; converting the cooling water at least partially into steam; accumulating the steam in the reactor well; flowing the steam through a heat exchanger; condensing the steam forming condensate in the heat exchanger; and returning the condensate to the reactor well, wherein the coolant steam and condensate circulates through a closed flow loop between the heat exchanger and reactor well. In one embodiment, the steam is produced within an insulating liner assembly disposed on an outside surface of the reactor vessel, the liner assembly being fluidly coupled to the reactor well via flow-hole nozzles disposed at the bottom and top portions of the reactor vessel. The liner assembly may comprise a plurality of spaced apart liners. The condensing step may further include the heat exchanger rejecting heat from the steam to an annular reservoir holding water that surrounds the containment vessel. The heat exchanger may comprises an array of heat dissipater ducts integrally attached to the containment vessel adjacent the annular reservoir. According to other aspects of the disclosure, the present invention further provides nuclear reactor containment system that overcomes the deficiencies of the foregoing arrangements for rejecting heat released into the environment within the containment by a thermal event. The containment system generally includes an inner containment vessel which may be formed of steel or another ductile material and an outer containment enclosure structure (CES) thereby forming a double walled containment system. In one embodiment, a water-filled annulus may be provided between the containment vessel and the containment enclosure structure providing an annular cooling reservoir. The containment vessel may include a plurality of longitudinal heat transfer fins which extend (substantially) radial outwards from the vessel in the manner of “fin”. The containment vessel thus serves not only as the primary structural containment for the reactor, but is configured and operable to function as a heat exchanger with the annular water reservoir acting as the heat sink. Accordingly, as further described herein, the containment vessel advantageously provides a passive (i.e. iron-pumped) heat rejection system when needed during a thermal energy release accident such as a LOCA or reactor scram to dissipate heat and cool the reactor. In one embodiment according to the present disclosure, a nuclear reactor containment system includes a containment vessel configured for housing a nuclear reactor, a containment enclosure structure (CES) surrounding the containment vessel, and an annular reservoir formed between the containment vessel and containment enclosure structure (CES) for extracting heat energy from the containment space. In the event of a thermal energy release incident inside the containment vessel, heat generated by the containment vessel is transferred to the annular reservoir which operates to cool the containment vessel. In one embodiment, the annular reservoir contains water for cooling the containment vessel. A portion of the containment vessel may include substantially radial heat transfer fins disposed in the annular reservoir and extending between the containment vessel and containment enclosure structure (CES) to improve the dissipation of heat to the water-filled annular reservoir. When a thermal energy release incident occurs inside the containment vessel, a portion of the water in the annulus is evaporated and vented to atmosphere through the containment enclosure structure (CES) annular reservoir in the form of water vapor. Embodiments of the system may further include an auxiliary air cooling system including a plurality of vertical inlet air conduits spaced circumferentially around the containment vessel in the annular reservoir. The air conduits are in fluid communication with the annular reservoir and outside ambient air external to the containment enclosure structure (CES). When a thermal energy release incident occurs inside the containment vessel and water in the annular reservoir is substantially depleted by evaporation, the air cooling system becomes operable by providing a ventilation path from the reservoir space to the external ambient. The ventilation system can thus be viewed as a secondary system that can continue to cool the containment ad infinitum. According to another embodiment, a nuclear reactor containment system includes a containment vessel configured for housing is nuclear reactor, a containment enclosure structure (CES) surrounding the containment vessel, a water filled annulus formed between the containment vessel and containment enclosure structure (CES) for cooling the containment vessel, and a plurality of substantially radial fins protruding outwards from the containment vessel and located, in the annulus. In the event of a thermal energy release incident inside the containment vessel, heat generated by the containment vessel is transferred to the water filled reservoir in the annulus through direct contact with the external surface of the containment vessel and its fins substantially radial thus cooling the containment vessel. In one embodiment, when a thermal energy release incident occurs inside the containment vessel and water in the annulus is substantially depleted by evaporation, the air wane system is operable to draw outside ambient air into the annulus through the air conduits to cool the heat generated in the containment (which decreases exponentially with time by natural convection. The existence of water in the annular region completely surrounding the containment vessel will maintain a consistent temperature distribution in the containment vessel to prevent warping of the containment vessel during the thermal energy release incident or accident. In another embodiment, a nuclear reactor containment system includes a containment vessel including a cylindrical shell configured for housing a nuclear reactor, a containment enclosure structure (CES) surrounding the containment vessel, an annular reservoir containing water formed between the shell of the containment vessel and containment enclosure structure (CES) for cooling the containment vessel, a plurality of external (substantially) radial fins protruding outwards from the containment vessel into the annulus, and an air cooling system including a plurality of vertical inlet air conduits spaced circumferentially around the containment vessel in the annular reservoir. The air conduits are in fluid communication with the annular reservoir and outside ambient air external to the containment enclosure structure (CES). In the event of a thermal energy release incident inside the containment vessel, heat generated by the containment vessel is transferred to the annular reservoir via the (substantially) radial containment wall along with its internal and external fins which operates to cool the containment vessel. Advantages and aspects of a nuclear reactor containment system according to the present disclosure include the following: Containment structures and systems configured so that a severe energy release event as described above can be contained passively (e.g. without relying on active components such as pumps, valves, heat exchangers and motors); Containment structures and systems that continue to work autonomously for an unlimited duration (e.g. no time limit for human intervention); Containment structures fortified with internal and external ribs (fins) configured to withstand a projectile impact such as a crashing aircraft without losing its primary function (i.e. pressure & radionuclide (if any) retention and heat rejection); and Containment vessel equipped with provisions that allow for the ready removal (or installation) of major equipment through the containment structure. All drawings are schematic and not necessarily to scale. References herein to a single drawing figure (e.g. FIG. 22) which has associated sub-parts (e.g. FIGS. 22A and 22B) shall be construed as a reference to the figure and sub-parts unless otherwise indicated. The features and benefits of the invention are illustrated and described herein by reference to illustrative embodiments. This description of illustrative embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such illustrative embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the nominal orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a rigorously specific orientation denoted by the term. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Referring to FIGS. 1-15, a nuclear reactor containment system 100 according to the present disclosure is shown. The system 100 generally includes an inner containment structure such as containment vessel 200 and an outer containment enclosure structure (CES) 300 collectively defining a containment vessel-enclosure assembly 200-300. The containment vessel 200 and containment enclosure structure (CES) 300 are vertically elongated and oriented, and define a vertical axis VA. In one embodiment, the containment vessel-enclosure assembly 200-300 is configured to be buried in the subgrade at least partially below grade (see also FIGS. 6-8). The containment vessel-enclosure assembly 200-300 may be supported by a concrete foundation 301 comprised of a bottom slab 302 and vertically extending sidewalls 303 rising from the slab forming a top base mat 304. The sidewalls 303 may circumferentially enclose containment vessel 200 as shown wherein a lower portion of the containment vessel may be positioned inside the sidewalls. In some embodiments, the sidewalls 303 may be poured after placement of the containment vessel 200 on the bottom slab 302 (which may be poured and set first) thereby completely embedding the lower portion of the containment vessel 200 within the foundation. The foundation walls 303 may terminate below grade in some embodiments as shown to provide additional protection for the containment vessel-enclosure assembly 200-300 from projectile impacts (e.g. crashing plane, etc.). The foundation 301 may have any suitable configuration in top plan view, including without limitation polygonal (e.g. rectangular, hexagon, circular, etc.). In one embodiment, the weight of the containment vessel 200 may be primarily supported by the bottom slab 302 on which the containment vessel rests and the containment enclosure structure (CES) 300 may be supported by the base mat 304 formed atop the sidewalls 303 of the foundation 301. Other suitable vessel and containment enclosure structure (CES) support arrangements may be used. With continuing reference to FIGS. 1-15, containment structure vessel 200 may be an elongated vessel including a hollow cylindrical shell 204 with circular transverse cross-section defining an outer diameter D1, a top head 206, and a bottom head 208. In one embodiment, the containment vessel 200 (i.e. shell and heads) may be made from a suitably strong and ductile metallic plate and bar stock that is readily weldable (e.g. low carbon steel). In one embodiment, a low carbon steel shell 204 may have a thickness of at least 1 inch. Other suitable metallic materials including various alloys may be used. The top head 206 may be attached to the shell 204 via a flanged joint 210 comprised of a first annular flange 212 disposed on the lower end or bottom of the top head and a second mating annular flange 214 disposed on the upper end or top of the shell. The flanged joint 210 may be a bolted joint, which optionally may further be seal welded after assembly with a circumferentially extending annular seal weld being made between the adjoining flanges 212 and 214. The top head 206 of containment vessel 200 may be an ASME (American Society of Mechanical Engineers) dome-shaped flanged and dished head to add structural strength (i.e. internal pressure retention and external impact resistance); however, other possible configurations including a flat top head might be used. The bottom head 208 may similarly be a dome-shaped dished head or alternatively flat in other possible embodiments. In one containment vessel construction, the bottom head 208 may be directly welded to the lower portion or end of the shell 204 via an integral straight flange (SF) portion of the head matching the diameter of shell. In one embodiment, the bottom of the containment vessel 200 may include a ribbed support stand 208a or similar structure attached to the bottom head 208 to help stabilize and provide level support for the containment vessel on the slab 302 of the foundation 301, as further described herein. In some embodiments, the top portion 216 of the containment vessel shell 204 may be a diametrically enlarged segment of the shell that forms a housing to support and accommodate a polar crane (not shown) for moving equipment, fuel, etc. inside the containment vessel. This will provide crane access to the very inside periphery of the containment vessel and enable placement of equipment very close to the periphery of the containment vessel 200 making the containment vessel structure compact. In one configuration, therefore, the above grade portion of the containment vessel 200 may resemble a mushroom-shaped structure. In one possible embodiment, the enlarged top portion 216 of containment vessel 200 may have an outer diameter D2 that is larger than the outer diameter D1 of the rest of the adjoining lower portion 218 of the containment vessel shell 204. In one non-limiting example, the top portion 216 may have a diameter D2 that is approximately 10 feet larger than the diameter D1 of the lower portion 218 of the shell 204. The top portion 216 of shell 204 may have a suitable height H2 selected to accommodate the polar crane with allowance for working clearances which may be less than 50% of the total height H1 of the containment vessel 200. In one non-limiting example, approximately the top ten feet of the containment vessel 200 (H2) may be formed by the enlarged diameter top portion 216 in comparison to a total height H1 of 200 feet of the containment vessel. The top portion 216 of containment vessel 200 may terminate at the upper end with flange 214 at the flanged connection to the top head 206 of the containment vessel. In one embodiment, the diametrically enlarged top portion 216 of containment vessel 200 has a diameter D2 which is smaller than the inside diameter D3 of the containment enclosure structure (CES) 300 to provide a (substantially) radial gap or secondary annulus 330 (see, e.g. FIG. 4). This provides a cushion of space or buffer region between the containment enclosure structure (CES) 300 and containment vessel top portion 216 in the advent of a projectile impact on the containment enclosure structure (CES). Furthermore, the annulus 330 further significantly creates a flow path between primary annulus 313 (between the shells of the containment enclosure structure (CES) 300 and containment vessel 200) and the head space 318 between the containment enclosure structure (CES) dome 316 and top head 206 of the containment vessel 200 for steam and/or air to be vented from the containment enclosure structure (CES) as further described herein. Accordingly, the secondary annulus 330 is in fluid communication with the primary annulus 313 and the head space 318 which in turn is in fluid communication with vent 317 which penetrates the dome 316. In one embodiment, the secondary annulus 330 has it smaller (substantially) radial width than the primary annulus 313. Referring to FIGS. 1-4, the containment enclosure structure (CES) 300 may be double-walled structure in some embodiments having sidewalk 320 formed by two (substantially) radially spaced and interconnected concentric shells 310 (inner) and 311 (outer) with plain or reinforced concrete 312 installed in the annular space between them. The concentric shells 310, 311 may be made of any suitably strong material, such as for example without limitation ductile metallic plates that are readily weldable (e.g. low carbon steel). Other suitable metallic materials including various alloys may be used. In one embodiment, without limitation, the double-walled containment enclosure structure (CES) 300 may have a concrete 312 thickness of 6 feet or more which ensures adequate ability to withstand high energy projectile impacts such as that from an airliner. The containment enclosure structure (CES) 300 circumscribes the containment vessel shell 204 and is spaced (substantially) radially apart from shell 204, thereby creating primary annulus 313. Annulus 313 may be a water-filled in one embodiment to create a heat sink for receiving and dissipating heat from the containment vessel 200 in the case of a thermal energy release incident inside the containment vessel. This water-filled annular reservoir preferably extends circumferentially for a full 360 degrees in one embodiment around the perimeter of upper portions of the containment vessel shell 204 lying above the concrete foundation 301. FIG. 4 shows a cross-section of the water-filled annulus 313 without the external (substantially) radial fins 221 in this figure fir clarity. In one embodiment, the annulus 313 is filled with water from the base mat 304 at the bottom end 314 to approximately the top end 315 of the concentric shells 310, 311 of the containment enclosure structure (CES) 300 to form an annular cooling water reservoir between the containment vessel shell 204 and inner shell 310 of the containment enclosure structure (CES). This annular reservoir may be coated or lined in some embodiments with a suitable corrosion resistant material such as aluminum, stainless steel, or a suitable preservative for corrosion protection. In one representative example, without limitation, the annulus 313 may be about 10 feet wide and about 100 feet high. In one embodiment, the containment enclosure structure (CES) 300 includes a steel dome 316 that is suitably thick and reinforced to harden it against crashing aircraft and other incident projectiles. The dome 316 may be removably fastened to the shells 310, 311 by a robust flanged joint 318. In one embodiment, the containment enclosure structure (CES) 300 is entirely surrounded on all exposed above grade portions by the containment enclosure structure (CES) 300, which preferably is sufficiently tall to provide protection for the containment vessel against aircraft hazard or comparable projectile to preserve the structural integrity of the water mass in the annulus 313 surrounding the containment vessel. In one embodiment, as shown, the containment enclosure structure (CES) 300 extends vertically below grade to a substantial portion of the distance to the top of the base mat 304. The containment enclosure structure (CES) 300 may further include at least one rain-protected vent 317 which is in fluid communication with the head space 318 beneath the dome 316 and water-filled annulus 313 to allow water vapor to flow, escape, and vent to atmosphere. In one embodiment, the vent 317 may be located at the center of the dome 316. In other embodiments, a plurality of vents may be provided spaced (substantially) radially around the dome 316. The vent 317 may be formed by a short section of piping in some embodiments which is covered by a rain hood of any suitable configuration that allows steam to escape from the containment enclosure structure (CES) but minimizes the ingress of water. In some possible embodiments, the head space 31 between the dome 316 and top head 206 of the containment vessel 200 may be filled with an energy absorbing material or structure to minimize the impact load on the containment enclosure structure (CES) dome 316 from a crashing (falling) projecting (e.g. airliner, etc.). In one example, a plurality of tightly-packed undulating or corrugated deformable aluminum plates may be disposed in part or all of the head space to form a crumple zone which will help absorb and dissipate the impact forces on the dome 316. Referring, primarily to FIGS. 1-5 and 8-17, the buried portions of the containment vessel 200 within the concrete foundation 301 below the base mat 304 may have a plain shell 204 without external features. Portions of the containment vessel shell 204 above the base mat 304, however, may include a plurality of longitudinal external (substantially) radial ribs or fins 220 which extend axially (substantially) parallel to vertical axis VA of the containment vessel-enclosure assembly 200-300. The external longitudinal fins 220 are spaced circumferentially around the perimeter of the containment vessel shell 204 and extend (substantially) radially outwards from the containment vessel. The ribs 220 serve multiple advantageous functions including without limitation (1) to stiffen the containment vessel shell 204, (2) prevent excessive “sloshing” of water reserve in annulus 313 in the occurrence of a seismic event, and (3) significantly to act as heat transfer “fins” to dissipate heat absorbed by conduction through the shell 204 to the environment of the annulus 313 in the situation of a fluid/steam release event in the containment vessel. Accordingly, in one embodiment to maximize the heat transfer effectiveness, the longitudinal fins 220 extend vertically for substantially the entire height of the water-filled annulus 313 covering the effective heat transfer surfaces of the containment vessel 200 (i.e. portions not buried in concrete foundation) to transfer heat from the containment vessel 200 to the water reservoir, as further described herein. In one embodiment, the external longitudinal fins 220 have upper horizontal ends 220a which terminate at or proximate to the underside or bottom of the larger diameter top portion 216 of the containment vessel 200, and lower horizontal ends 220b which terminate at or proximate to the base mat 304 of the concrete foundation 301. In one embodiment, the external longitudinal fins 220 may have a height H3 which is equal to or greater than one half of a total height of the shell of the containment vessel. In one embodiment, the upper horizontal ends 220a of the longitudinal fins 220 are free ends not permanently attached (e.g. welded) to the containment vessel 200 or other structure. At least part of the lower horizontal ends 220b of the longitudinal fins 220 may abuttingly contact and rest on a horizontal circumferential rib 222 welded to the exterior surface of the containment vessel shell 204 to help support the weight of the longitudinal fins 220 and minimize stresses on the longitudinal welds. Circumferential rib 222 is annular in shape and may extend a full 360 degrees completely around the circumferential of the containment vessel shell 204. In one embodiment, the circumferential rib 222 is located to rest on the base mat 304 of the concrete foundation 301 which transfers the loads of the longitudinal fins 220 to the foundation. The longitudinal fins 220 may have a lateral extent or width that projects outwards beyond the outer peripheral edge of the circumferential rib 222. Accordingly, in this embodiment, only the inner portions of the lower horizontal end 220b of each rib 220 contacts the circumferential rib 222. In other possible embodiments, the circumferential rib 222 may extend (substantially) radially outwards far enough so that substantially the entire lower horizontal end 220b of each longitudinal rib 220 rests on the circumferential rib 222. The lower horizontal ends 220b may be welded to the circumferential rib 222 in some embodiments to further strengthen and stiffen the longitudinal fins 220. The external longitudinal fins 220 may be made of steel (e.g. low carbon steel), or other suitable metallic materials including alloys which are each welded on one of the longitudinally-extending sides to the exterior of the containment vessel shell 204. The opposing longitudinally-extending side of each rib 220 lies proximate to, but is preferably not permanently affixed to the interior of the inner shell 310 of the containment enclosure structure (CES) 300 to maximize the heat transfer surface of the ribs acting as heat dissipation fins. In one embodiment, the external longitudinal fins 220 extend (substantially) radially outwards beyond the larger diameter top portion 216 of the containment vessel 200 as shown. In one representative example, without limitation, steel ribs 220 may have a thickness of about 1 inch. Other suitable thickness of ribs may be used as appropriate. Accordingly, in some embodiments, the ribs 220 have a radial width that is more than 10 times the thickness of the ribs. In one embodiment, the longitudinal fins 220 are oriented at an oblique angle A1 to containment vessel shell 204 as best shown in FIGS. 2-3 and 5. This orientation forms a crumple zone extending 360 degrees around the circumference of the containment vessel 200 to better resist projectile impacts functioning in cooperation with the outer containment enclosure structure (CES) 300. Accordingly, an impact causing inward deformation of the containment enclosure structure (CES) shells 210, 211 will bend the longitudinal fins 220 which in the process will distribute the impact forces preferably without direct transfer to and rupturing or the inner containment vessel shell 204 as might possibly occur with ribs oriented 90 degrees to the containment vessel shell 204. In other possible embodiments, depending on the construction of the containment enclosure structure (CES) 300 and other factors, a perpendicular arrangement of ribs 220 to the containment vessel shell 204 may be appropriate. In one embodiment, referring to FIGS. 6-8, portions of the containment vessel shell 204 having and protected by the external (substantially) radial fins 220 against projectile impacts may extend below grade to provide protection against projectile strikes at or slightly below grade on the containment enclosure structure (CES) 300. Accordingly, the base mat 304 formed at the top of the vertically extending sidewalls 303 of the foundation 301 where the fins 220 terminate at their lower ends may be positioned a number of feet below grade to improve impact resistance of the nuclear reactor containment system. In one embodiment, the containment vessel 200 may optionally include a plurality of circumferentially spaced apart internal (substantially) radial fins 221 attached to the interior surface of the shell 204 (shown as dashed in FIGS. 2 and 3). Internal fins 221 extend (substantially) radially inwards from containment vessel shell 204 and longitudinally in a vertical direction of a suitable height. In one embodiment, the internal (substantially) radial fins 221 may have a height substantially coextensive with the height of the water-filled annulus 313 and extend from the base mat 304 to approximately the top of the shell 204. In one embodiment, without limitation, the internal fins 221 may be oriented substantially perpendicular (i.e. 90 degrees) to the containment vessel shell 204. Other suitable angles and oblique orientations may be used. The internal fins function to both increase the available heat transfer surface area and structurally reinforce the containment vessel shell against external impact (e.g. projectiles) or internal pressure increase within the containment vessel 200 in the event of a containment pressurization event (e.g. LOCA or reactor scram). In one embodiment, without limitation, the internal fins 221 may be made of steel. Referring to FIGS. 1-15, a plurality of vertical structural support columns 331 may be attached to the exterior surface of the containment vessel shell 204 to help support the diametrically larger top portion 216 of containment vessel 200 which has peripheral sides that are cantilevered (substantially) radially outwards beyond the shell 204. The support columns 331 are spaced circumferentially apart around the perimeter of containment vessel shell 204. In one embodiment, the support columns 331 may be formed of steel hollow structural members, for example without limitation C-shaped members in cross-section (i.e. structural channels), which are welded to the exterior surface of containment vessel shell 204. The two parallel legs of the channels may be vertically welded to the containment vessel shell 204 along the height of each support column 331 using either continuous Or intermittent welds such as stitch welds. The support columns 331 extend vertically downwards front and may be welded at their top ends to the bottom/underside of the larger diameter top portion 216 of containment vessel housing the polar crane. The bottom ends of the support columns 331 rest on or are welded to the circumferential rib 222 which engages the base mat 304 of the concrete foundation 301 near the buried portion of the containment. The columns 331 help transfer part of the dead load or weight from the crane and the top portion 216 of the containment vessel 300 down to the foundation. In one embodiment, the hollow space inside the support columns may be filled with concrete (with or without rebar) to help stiffen and further support the dead load or weight. In other possible embodiments, other structural steel shapes including filled or unfilled box beams, I-beams, tubular, angles, etc. may be used. The longitudinal fins 220 may extend farther outwards in a (substantially) radial direction than the support columns 331 which serve a structural role rather than a heat transfer role as the ribs 220. In certain embodiments, the ribs 220 have a (substantially) radial width that is at least twice the (substantially) radial width of support columns. FIGS. 11-15 show various cross sections (both longitudinal and transverse) of containment vessel 200 with equipment shown therein. In one embodiment, the containment vessel 200 may be part of a small modular reactor (SMR) system such as SMR-160 by Holtec International. The equipment may generally include a nuclear reactor vessel 500 disposed in a wet well 504 and defining an interior space housing a nuclear fuel core inside and circulating primary coolant, and as steam generator 502 fluidly coupled to the reactor and circulating a secondary coolant which may form part of a Rankine power generation cycle. Such a system is described for example in PCT International Patent Application No. PCT/US13/66777 filed Oct. 25, 2013, which is incorporated herein by reference in its entirety. Other appurtenances and equipment may be provided to create a complete steam generation system. Steam generator 502 is more fully described in International PCT Application No. PCT/US13/38289 filed Apr. 25, 2013, which is incorporated herein by reference in its entirety. As described therein and shown in FIGS. 11, 12, and 23 of the present application, the steam generator 502 may be vertically oriented and axially elongated similarly to submerged bundle heat exchanger 620. The steam generator 502 may be comprised of a set of tubular heat exchangers arranged in a vertical stack configured to extract the reactor's decay heat from the primary coolant by gravity-driven passive flow means. The circulation flow loops of primary coolant (liquid water) and secondary coolant (liquid feedwater and steam) through the reactor vessel and steam generator during normal operation of the reactor and power plant with an available electric supply produced by the station turbine-generator (T-G) set is shown in FIG. 23 herein. The primary coolant flow between the fluidly coupled steam generator 502 and reactor vessel 500 forms a first closed flow loop for purposes of the present discussion. In one embodiment, the primary coolant flow is gravity-driven relying on the change in temperature and corresponding density of the coolant as it is heated in the reactor vessel 500 by nuclear fuel core 501, and then cooled in the steam generator 502 as heat is transferred to the secondary coolant loop of the Rankine cycle which drives the turbine-generator set. The pressure head created by the changing different densities of the primary coolant (i.e. hot—lower density and cold—higher density) induces flow or circulation through the reactor vessel-steam generating vessel system as shown by the directional flow arrows. In general with respect to a pressurized closed flow loop, the primary coolant is heated by the nuclear fuel core 501 and flows upwards in riser column 224. The primary coolant from the reactor vessel 500 then flows through the primary coolant fluid coupling 273 between the reactor vessel 500 and steam generator 502 and enters the steam generator. The primary coolant flows upward in the centrally located riser pipe 337 to a pressurizer 380 at the top of the steam generator. The primary coolant reverses direction and flows down through the tube side of the steam generator 502 and returns to the reactor vessel 500 through the fluid coupling 273 where it enters an annular downcomer 222 to complete the primary coolant flow loop. The steam generator 502 may include three vertically stacked heat transfer sections—from bottom up a preheater section 351, steam generator section 352, and superheater section 350 (See, e.g. FIGS. 11, 12, and 23). Secondary coolant flows on the shellside of the steam generator 502 vessel. Secondary coolant in the form of liquid feedwater from the turbine-generator (T-G) set of the Rankine cycle enters the steam generator at the bottom in the preheater section 351 and flows upwards through the steam generator section 352 being converted to steam. The steam flows upwards into the superheater section 350 and reaches superheat conditions. From there, the superheated steam is extracted and flows to the T-G set to produce electric power. Auxiliary Heat Dissipation System Referring primarily now to FIGS. 2-3, 16, and 18, the containment vessel 200 may further include an auxiliary heat dissipation system 340 comprising a discrete set or array of heat dissipater ducts 341 (HDD). In one embodiment, the auxiliary heat dissipation system 340 and associated heat dissipater ducts 341 may form part of a passive reactor core cooling system described in further detail below and shown in FIGS. 22 and 23. Heat dissipater ducts 341 include a plurality of internal longitudinal ducts (i.e. flow conduits) circumferentially spaced around the circumference of containment vessel shell 204. Ducts 341 extend vertically parallel to the vertical axis VA and in one embodiment are attached to the interior surface of shell 204. The ducts 341 may be made of metal such as steel and are welded to interior of the shell 204. In one possible configuration, without limitation, the ducts 341 may be comprised of vertically oriented C-shaped structural channels (in cross section) or half-sections of pipe/tube positioned so that the parallel legs of the channels or pipe/tubes are each seam welded to the shell 204 for their entire height to define a sealed vertical flow conduit. The fluid (liquid or steam phase) in the heat dissipater ducts in this embodiment therefore directly contacts the reactor containment vessel 200 to maximize heat transfer through the vessel to the water in the annular reservoir (primary annulus 313) which forms a heat sink for the reactor containment vessel 200 and the heat dissipater ducts. Other suitably shaped and configured heat dissipater ducts 341 may be provided for this type construction so long as the fluid conveyed in the ducts contacts at least a portion of the interior containment vessel shell 204 to transfer heat to the water-filled annulus 313. In other possible but less preferred acceptable embodiments, the heat dissipater ducts 341 may be formed from completely tubular walled flow conduits (e.g. full circumferential tube or pipe sections rather than half sections) which are welded to the interior containment vessel shell 204. In these type constructions, the fluid conveyed in the ducts 341 will transfer heat indirectly to the reactor containment vessel shell 204 through the wall of the ducts first, and then to the water-filled annulus 313. Any suitable number and arrangement of ducts 341 may be provided depending on the heat transfer surface area required for cooling the fluid flowing through the ducts. The ducts 341 may be uniformly or non-uniformly spaced on the interior of the containment vessel shell 204, and in some embodiments grouped clusters of ducts may be circumferentially distributed around the containment vessel. The ducts 341 may have any suitable cross-sectional dimensions depending on the flow rate of fluid carried by the ducts and heat transfer considerations. The open upper and lower ends 341a, 341b of the ducts 341 are each fluidly connected to a common upper inlet ring header 343 and lower outlet ring header 344. The annular shaped ring headers 343, 344 are vertically spaced apart and positioned at suitable elevations on the interior of the containment vessel 200 to maximize the transfer of heat between fluid flowing vertically inside ducts 341 and the shell 204 of the containment vessel in the active heat transfer zone defined by portions of the containment vessel having the external longitudinal fins 220 in the primary annulus 313. To take advantage of the primary water-filled annulus 313 for heat transfer, upper and lower ring headers 343, 344 may each respectively be located on the interior of the containment vessel shell 204 adjacent and near to the top and bottom of the annulus. In one embodiment, the ring headers 343, 344 may each be formed of half-sections of arcuately curved steel pipe as shown which are welded directly to the interior surface of containment vessel shell 204 in the manner shown. In other embodiments, the ring headers 343, 344 may be formed of complete sections of arcuately curved piping supported by and attached to the interior of the shell 204 by any suitable means. In one embodiment, the heat dissipation system 340 is fluidly connected to a source of steam that may be generated from a water mass inside the containment vessel 200 to reject radioactive material decay heat from the reactor core. The containment surface enclosed by the ducts 341 serves as the heat transfer surface to transmit the latent heat of the steam inside the ducts to the shell 204 of the containment vessel 200 for cooling via the external longitudinal fins 220 and water filled annulus 313. In operation, steam enters the inlet ring header 343 and is distributed to the open inlet ends of the ducts 341 penetrating the header. The steam enters the ducts 341 and flows downwards therein along the height of the containment vessel shell 204 interior and undergoes a phase change from steam to liquid. The condensed steam drains down by gravity in the ducts and is collected by the lower ring header 344 from which it is returned back to the source of steam also preferably by gravity in one embodiment. It should be noted that no pumps are involved or required in the foregoing process. It will be appreciated that in certain embodiments, more than one set or array of heat dissipater ducts 341 may be provided and arranged on the inside surface of the inner containment vessel 200 within the containment space defined by the vessel. Auxiliary Air Cooling System According to another aspect of the present disclosure, a secondary or backup passive air cooling system 400 is provided to initiate air cooling by natural convection of the containment vessel 200 if, for some reason, the water inventory in the primary annulus 313 were to be depleted during a thermal reactor related event (e.g. LOCA or reactor scram). Referring to FIG. 8, the air cooling system 400 may be comprised of a plurality of vertical inlet air conduits 401 spaced circumferentially around the containment vessel 200 in the primary annulus 313. Each air conduit 401 includes an inlet 402 which penetrates the sidewalk 320 of the containment enclosure structure (CES) 300 and is open to the atmosphere outside to draw in ambient cooling air. Inlets 402 are preferably positioned near the upper end of the containment enclosure structure's sidewalls 320. The air conduits 401 extend vertically downwards inside the annulus 313 and terminate a short distance above the base mat 304 of the foundation (e.g. approximately 1 foot) to allow air to escape from the open bottom ends of the conduits. Using the air conduits 401, a natural convection cooling airflow pathway is established in cooperation with the annulus 313. In the event the cooling water inventory in the primary annulus 313 is depleted by evaporation during a thermal event, air cooling automatically initiates by natural convection as the air inside the annulus will continue to be heated by the containment vessel 200. The heated air rises in the primary annulus 313, passes through the secondary annulus 330, enters the head space 318, and exits the dome 316 of the containment enclosure structure (CES) 300 through the vent 317 (see directional flow arrows, FIG. 8). The rising heated air creates a reduction in air pressure towards the bottom of the primary annulus 313 sufficient to draw in outside ambient downwards through the air conduits 401 thereby creating a natural air circulation pattern which continues to cool the heated containment vessel 200. Advantageously, this passive air cooling system and circulation may continue for an indefinite period of time to cool the containment vessel 200. It should be noted that the primary annulus 313 acts as the ultimate heat sink for the heat generated inside the containment vessel 200. The water in this annular reservoir also acts to maintain the temperature of all crane vertical support columns 331 (described earlier) at essentially the same temperature thus ensuring the levelness of the crane rails (not shown) at all times which are mounted in the larger portion 216 of the containment vessel 200. Operation of the reactor containment system 100 as a heat exchanger will now be briefly described with initial reference to FIG. 19. This figure is a simplified diagrammatic representation of the reactor containment system 100 without all of the appurtenances and structures described herein for clarity in describing the active heat transfer and rejection processes performed by the system. In the event of a loss-of-coolant (LOCA) accident, the high energy fluid or liquid coolant (which may typically be water) spills into the containment environment formed by the containment vessel 200. The liquid flashes instantaneously into steam and the vapor mixes with the air inside the containment and migrates to the inside surface of the containment vessel 200 sidewalls or shell 204 (since the shell of the containment is cooler due the water in the annulus 313). The vapor then condenses on the vertical shell walls by losing its latent heat to the containment structure metal which in turn rejects the heat to the water in the annulus 313 through the longitudinal fins 220 and exposed portions of the shell 204 inside the annulus. The water in the annulus 313 heats up and eventually evaporates forming a vapor which rises in the annulus and leaves the containment enclosure structure (CES) 300 through the secondary annulus 330, head space 318, and finally the vent 317 to atmosphere. As the water reservoir in annulus 313 is located outside the containment vessel environment, in some embodiments the water inventory may be easily replenished using external means if available to compensate for the evaporative loss of water. However, if no replenishment water is provided or available, then the height of the water column in the annulus 313 will begin to drop. As the water level in the annulus 313 drops, the containment vessel 200 also starts to heat the air in the annulus above the water level, thereby rejecting a portion of the heat to the air which rises and is vented from the containment enclosure structure (CES) 300 through vent 317 with the water vapor. When the water level drops sufficiently such that the open bottom ends of the air conduits 401 (see, e.g. FIG. 8) become exposed above the water line, fresh outside ambient air will then be pulled in from the air conduits 401 as described above to initiate a natural convection air circulation pattern that continues cooling the containment vessel 200. In one embodiment, provisions (e.g. water inlet line) are provided through the containment enclosure structure (CES) 300 for water replenishment in the annulus 313 although this is not required to insure adequate heat dissipation. The mass of water inventor in this annular reservoir is sized such that the decay heat produced in the containment vessel 200 has declined sufficiently such that the containment is capable of rejecting all its heat through air cooling alone once the water inventor is depleted. The containment vessel 200 preferably has sufficient heat rejection capability to limit the pressure and temperature of the vapor mix inside the containment vessel (within its design limits) by rejecting the thermal energy rapidly. In the event of a station blackout, the reactor core is forced into a “scram” and the passive core cooling systems will reject the decay heat of the core in the form of steam directed to upper inlet ring header 343 of heat dissipation system 340 already described herein (see. e.g. FIGS. 16 and 18). The steam then flowing downwards through the network of internal longitudinal ducts 341 comes in contact with the containment vessel shell 204 interior surface enclosed within the heat dissipation ducts and condenses by rejecting its latent heat to the containment structure metal, winch in turn rejects the heat to the water in the annulus via heat transfer assistance provide by the longitudinal fins 220. The water in the annular reservoir (primary annulus 313) heats up eventually evaporating. The containment vessel 200 rejects the heat to the annulus by sensible heating and then by a combination of evaporation and air cooling, and then further eventually by natural convection air cooling only as described herein. As mentioned above, the reactor containment system 100 is designed and configured so that air cooling alone is sufficient to reject the decay heat once the effective water inventory in annulus 313 is entirely depleted. In both these foregoing scenarios, the heat rejection can continue indefinitely until alternate means are available to bring the plant back online. Not only does the system operate indefinitely, but the operation is entirely passive without the use of any pumps or operator intervention. Passive Reactor Cooling System According to another aspect of the invention, a passive gravity-driven nuclear reactor cooling system 600 is provided to reject the reactor's decay heat following a loss-of-coolant accident (LOCA) during which time the reactor is shutdown (e.g. “scram”). The cooling system does not rely on and suffer the drawbacks of pumps and motors which require an available electric supply. Accordingly, the reactor cooling system 600 can advantageously operate during a power plant blackout situation. Referring to FIGS. 20 and 21, the passive reactor cooling system 600 in one embodiment is an atmospheric pressure closed loop flow system in one embodiment comprised of three major fluidly coupled parts or sub-systems, namely (i) a reactor well 620, (ii) a discrete set or array of heat dissipater ducts 341 (HDD) integrally connected to the inner wall of the containment structure (described in detail above), and (iii) an in-containment reactor water storage tank 630 filled with a reserve of cooling water. The reactor cooling system 600 is configured to utilize cooling water flooded into the reactor well 620 from the storage tank to extract the thermal energy generated by the fuel core during a reactor shutdown and LOCA that can continue indefinitely in the absence of an available source of electric power, as further described herein. Although FIGS. 20 and 21 shows the reactor well 620 in the flooded condition, it should be noted that the reactor well is dry and empty during the normal power generation operating mode of the reactor prior to a LOCA event. Referring to FIGS. 20-23, the reactor vessel 500 containing the nuclear core 501 is disposed in reactor well 620 defined by a large concrete monolith 621. The monolith 621 is formed inside the inner containment vessel 200 (best shown in FIG. 21). Reactor vessel 500 is generally funned by a vertically elongated cylindrical shell (sidewall) and a closed bottom head 505. Accordingly, the reactor vessel 500 is vertically oriented with a majority of the height or length of the reactor vessel being positioned inside the reactor well as shown. The reactor well 620 is an annular vacant space surrounding the reactor vessel 500 and may be dry and unfilled during normal power generation operation of the reactor. The bottom head 505 of the reactor vessel 500 is spaced above the bottom of the reactor well 620. The top of the reactor well 620 may be partially or completely closed by a closure structure. In one embodiment, the closure structure may be formed at least in part by a ring-shaped reactor support flange 632 that extends circumferentially around the perimeter of the reactor vessel 500. The annular support flange may be supported by the concrete monolith 621. Additional structural and other elements (e.g. metal, concrete, seals/gaskets, etc.) may be provided to supplement the support flange 632 and to seal the top of the reactor well 630 if it is to be completely sealed for better capturing steam present in the reactor well which is directed to the auxiliary heat dissipation system 340, as further described herein. The outer wall of the reactor well 620 may be insulated by one or more layers of stainless steel liners 700 with small interstitial space or air gap formed between them (see, e.g. FIGS. 22, 22A, 22B). For additional cooling of the reactor well space, cold water may be circulated in the inter-liner spaces in some embodiments. The stainless steel liners 700 serve to block extensive heating of the concrete monolith 621 forming the reactor well. Referring to FIGS. 20 and 22 (including sub-parts A and B), the outside surface of the reactor vessel 500 may also be insulated by a liner assembly comprised of one or more layers of metal liners 701 with small interstitial spaces or air gaps therebetween which serve to retard the outflow of heat generated by the reactor core 501 during normal reactor operation. In some non-limiting examples, the liners may preferably be stainless steel or aluminum; however, other suitable metals for a reactor well environment may be used. Preferably, in one embodiment, the liners 701 may extend completely around the circumference and the entire height of the reactor vessel 500 that is positioned within the reactor well 620 including under the bottom head 505 of the reactor vessel. The entire perimeter of the reactor vessel 500 lying within the reactor well may therefore include the liners 701 such that a plurality of liners is disposed between the outside surface of the reactor vessel 500 and outermost liner 510. The insulating liner assembly comprised of liners 701 may include an array of one or more flow-holes which may be formed by top flow-hole nozzles 702 disposed in the upper sidewall (shell) region of the reactor vessel 500 and reactor well 620, preferably below the first pipe penetration into the reactor vessel in one embodiment. The nozzles 702 are in fluid communication with the air gaps (interstitial spaces) in the insulating liner assembly and space formed within the reactor well 620. The top flow-hole nozzles 702 are therefore disposed on the outside surface of the reactor vessel sidewall, but are not in fluid communication with the interior of the reactor vessel 500 and primary coolant therein. Although in some embodiments the nozzles 702 may be attached to outside surface of the reactor vessel for support, the nozzles are instead configured to be in fluid communication with the air gaps formed in the side liner 701 assembly on the outside of the reactor vessel as noted above. In one embodiment, for example, this may be accomplished by providing a plurality of lateral holes in the nozzles 702 adjacent the air gaps between the liners 701. The top flow-hole nozzles 702 are configured and operable to evacuate steam flowing within the liner assembly and discharge the steam to the reactor well, as further described herein. The top flow-hole nozzles 702 may be circumferentially spaced around the reactor vessel. In one non-limiting embodiment, four top flow-hole nozzles 702 may be provided at approximately the same elevation. Other arrangements and numbers of top flow-hole nozzles 702 may be provided. One or more bottom flow-hole nozzles 703 may also be provided for the vessel liners 701 adjacent the bottom bead 505 of the reactor vessel 500. In one embodiment, a single larger nozzle 703 may be provided which is concentrically aligned with the centerline CL of the reactor vessel 500 at the lowest point on the arcuate bottom reactor vessel head 505. The nozzle 703 may be supported, configured, and arranged to form fluid communication with the air gaps (interstitial spaces) between the bottom liners 701 and reactor well 620 in similar fashion as the top flow-hole nozzles 702. Nozzle 703 may therefore be constructed and operate similarly to top flow-hole nozzles 702 being supported by, but not in fluid communication with the interior of the reactor vessel 500 and primary coolant therein. The bottom flow-hole nozzle 703 is configured and operable to admit cooling water in the reactor well from the water storage tank 630 into the lower portion of the insulating liner assembly, as further described herein. The top flow-hole nozzles 702 may have provisions such as closure flaps 704 which are designed to remain closed during normal operation of the reactor when the gaps between the reactor vessel 500 and the liners 701 are filled with air (see, e.g. FIG. 22A). The flap and nozzle combination forms a flap valve. The flaps 704 are each pivotably movable and connected to its respective nozzle 702 at a top end by a pivot 705. Any suitable type of pivot may be provided, such as without limitation a pinned joint or self-hinge wherein the flap is made of a flexible material such as a high temperature withstanding polymer. The flaps 704 may be made of any suitable metallic or non-metallic material. The vertical orientation and weight of the flap 704 holds it in the closed position against the free end of nozzle 702 by gravity. In other embodiments, a commercially available flap valve comprising a valve body and flap may instead be mounted on the free end of the top flow-hole nozzles 702 to provide the same functionality. The bottom flow-hole nozzles 703 are also normally each closed by a flap 706 during normal operation of the reactor when the gaps between the reactor vessel 500 and the liners 701 are filled with air (see, e.g. FIG. 22B). In one embodiment, the flaps 706 may be held closed via a float device including a buoyant float 709 rigidly connected to one end of the flap by a linkage arm 708. The flap 706 and linkage arm 708 assembly is pivotably coupled to a bottom nozzle 703 by a pivot 707, such as without limitation a pinned joint in one embodiment. Flap 706 is preferably made of a rigid metallic or non-metallic material in order to maintain its shape and seal against the free end of nozzle 703 when in its closed position. In operation, gravity acts downward on the float 709 when the reactor well 620 is empty during normal operation of the reactor. This rotates the float 709 and the flap 706 assembly in a counter-clockwise direction to force the flap against the free end of nozzle 703. When water floods the reactor well 620 from storage tank 630 during a LOCA event as further described herein, the rising water will cause the float 709 to rotate upwards now in a clockwise direction. This simultaneously rotates the flap clockwise and downward opening the nozzle 703 admitting water into the air gaps between the reactor vessel 500 metal shell wall and the stainless steel liners 701. When the cooling water W from water storage tank 630 enters the air gaps between the liners 701 and comes in contact with the metal reactor vessel 500 wall after the passive reactor cooling system 600 is activated, the water vaporizes producing steam which raises the pressure in the gap. This buildup of pressure forces the flaps 704 of the top flow-hole nozzles 702 to open and relieve the steam build up into the reactor well 620 which is subsequently routed to the heat dissipation ducts 341 of the auxiliary heat dissipation system 340, as further described herein. Accordingly, the cooling water W therefore enters the liners 701 through the open flap(s) 706 of the bottom flow-hole nozzle(s) 703 and is evacuated from the liner assembly through the top flow-hole nozzles 702 in the form of steam. Referring now to FIGS. 20 and 21, the concrete monolith 621 further defines a large in-containment cooling water storage tank 630 (i.e. within the inner containment vessel 200 also variously shown in FIGS. 1-19). The water tank 630 holds a reserve of cooling water W and is fluidly coupled and positioned to dump its contents into the reactor well 620 in the event of a LOCA. In one embodiment, water storage tank 630 is fluidly coupled to the reactor well 620 by an upper and lower flow conduit 633 in which dump valves 631 are positioned to control flow. At least one flow conduct 633 with dump valve 634 may be provided; however, in some embodiments more than two flow conduits with dump valves may be provided. The dump valve may be operated in a fully opened or closed mode, or alternatively if needed throttled in a partially open mode. During normal power generation operation of the reactor, the dump valves are normally closed to prevent cooling water W from flooding into the reactor well 620 through the flow conduits. The dump valves 631 may be automatically operated via electric or pneumatic valve operators. In one embodiment, the dump valves 631 may be configured to operate as “fail open” when power supply is lost to the valves to automatically flood the reactor well 620 with cooling water W. In some preferred non-limiting embodiments, the cooling water tank 630 has a volumetric capacity at least as large as or larger than the capacity of the reactor well 620 to optimize cooling the reactor core and replenishing any cooling water W in the reactor well which might be lost as steam to the containment space in designs where the top of the reactor well is either not intentionally fully enclosed amid/or tightly sealed or may be damaged. A method for operating the passive reactor cooling system 600 will now be described with primary reference to FIGS. 20-22. As mentioned earlier in this disclosure, in the case of a LOCA, the pressure and temperature in the containment will rise. When the containment pressure (or temperature) reaches a pre-set threshold value, then the dump valves 631 connecting the water storage tank 630 and reactor well 620 are opened causing a rapid transfer of cooling water W and filling of the reactor well. The insulating liners 701 on the reactor vessel 500 protect it from rapid quenching (and high thermal stresses). After the water in the reactor well 620 reaches near the top flow-hole nozzle 702 in the liner 701 assembly (until then the reactor vessel is undergoing limited cooling that the heat transfer across the liners to the reactor well water), then the cold cooling water W begins to fill the interstitial spaces between the liners and the reactor vessel thus significantly accelerating the extraction of decay heat from the reactor core 501 and reactor vessel. After some time, the temperature of the pool of deposited water in the reactor well 620 reaches the boding point temperature and begins to boil. The steam thus produced rises by buoyancy action through inlet piping 603 to the bank of heat dissipater ducts 341 of the auxiliary heat dissipation system 340, as described above and shown in FIGS. 16, 18, and 21. These ducts 341 condense the steam generated in the reactor well pool and return the condensate to the reactor well 620 via outlet piping 603 with the latent heat of steam delivered to the external annular reservoir 313 holding water having a temperature lower than the steam to form a heat sink in thermal communication with the containment vessel 200. Accordingly, the heat from the spilled reactor cooling system primary coolant water (e.g. via a primary coolant piping failure) is thus removed by the containment, albeit less efficiently, as the water/air mixture rises and contacts the internal surface of the containment (which is equipped with large external and internal fins 220, 221 shown in FIG. 3 and described above) to facilitate the heat extraction. It should be noted that the flow of steam and condensate between the heat dissipater ducts 341 and reactor well 620 is advantageously driven solely by gravity due to the changing densities of the steam and condensate, without need for pumps and an available power supply. The heat dissipater ducts 341 are therefore preferably positioned on the inner containment vessel 200 wall at higher location than the reactor well 630 and the extraction point of steam from the reactor well. Flow of steam and condensate through the inlet and outlet piping 603 to and from the array of heat dissipater ducts 341 may be controlled by suitable valves 625 (see FIG. 20), which may be operated in an on/off mode, or throttled. Valves 625 may be configured to operate as “fail open” when power supply is lost to the valves which may have electric or pneumatic valve operators. This automatically opens and actuates the closed flow loop of the reactor cooling system 600 between the heat dissipater ducts 341 and reactor well 620. The inlet steam piping 603 to the heat dissipater ducts 341 may be fluidly coupled to the top portion of reactor well 620 to optimally capture the accumulating steam. The outlet condensate return piping 603 may be fluidly coupled to the top portion of water storage tank 630 to optimally capture the accumulating steam. The atmospheric closed flow loop of the reactor cooling system 600 between the reactor well 620 and heat dissipater ducts 341 may therefore flow through the water storage tank 630 (see FIG. 21). In the event of a LOCA, as the water inventory in the annular reservoir 313 between the inner containment vessel 200 and outer containment enclosure structure 300 evaporates, it may be readily replenished. However, if replenishment is not possible, then the receding water inventory in the reservoir 313 will actuate rejection of heat to the air by ventilation action using the passive air cooling system 400 described above. Once all the water has evaporated in the reservoir 313, the containment structure will continue to reject heat by air cooling alone. Air cooling after a prolonged period of water cooling is ideally sufficient to remove all the decay heat. This also passive gravity driven heat expulsion process driven by changing air densities can continue as long as necessary to cool the reactor. It will be appreciated that numerous variations of the foregoing method for operating the passive reactor cooling system 600 are possible. While the foregoing description and drawings represent some example systems, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made. One skilled in the art will further appreciate that the invention may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the invention, which are particularly adapted to specific environments and operative requirements without departing from the principles of the present invention. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive, the scope of the invention being defined by the appended claims and equivalents thereof, and not limited to the foregoing description or embodiments. Rather, the appended claims should be construed broadly, to include other variants and embodiments of the invention, which may be made by those skilled in the art without departing from the scope and range of equivalents of the invention. |
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039986918 | claims | 1. A process for producing radioactive iodine (I-131) comprising: heating a compound selected from the group consisting of: telluric acid and tellurium trioxide, at a temperature in the range of about 500.degree. to 560.degree. C to obtain a tellurium oxide intermediate having a composition TeO.sub.3.nTeO.sub.2, where n is in the range from about 3.5 to about 4.5, irradiating said tellurium oxide intermediate with a neutron flux to form I-131 in said intermediate, and converting the irradiated intermediate to tellurium dioxide by pumping a carrier gas therethrough while heating said irradiated intermediate at a temperature in the range of its decomposition temperature up to about 650.degree. C, thereby releasing said I-131. 2. The process according to claim 1, in which the conversion is effected at a temperature in the range of about 560.degree. - 650.degree. C. 3. The process according to claim 1, in which the irradiation is effected to a degree sufficient to convert a substantial amount of the Te-130 contained in the tellurium oxide intermediate to Te-131. 4. The process according to claim 1, in which the irradiation is effected in a nuclear reactor. |
abstract | A control rod spider assembly connection between a connecting finger and a rodlet. An upper end plug of the rodlet is secured to the inner bore of the hollow connecting finger with a mating one of a right hand or left hand thread interfacing between the interior of the bore and the circumference of the rodlet. The upper end of the rodlet is captured by a second fastener mechanism having the other of the right hand or the left hand thread. The second fastener mechanism is anchored to one or both of the connecting finger or the upper end plug to secure the connection. |
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056420145 | abstract | The invention provides a self-powered device having at least one substrate, at least one radioactive power source formed over the substrate, and integrated circuits formed over the substrate. The radioactive power source includes a first active layer of a first conductivity type, a second active layer of a second conductivity type. The first and second active layers form a depletion layer. A tritium containing layer is provided which supplies beta particles that penetrates the depletion layer generating electron-hole pairs. The electron-hole pairs are swept by the electric field in the depletion layer producing an electric current. |
claims | 1. A method of flow control management in a nuclear reactor, the method comprising:positioning a control rod guide tube in a lower plenum of the nuclear reactor,defining a flow channel and one or more ports in the control rod guide tube;fluidly coupling an upper end portion of the control rod guide tube to a fuel support;defining a cavity in the fuel support:fluidly coupling the fuel support to a lumen on a lower tie plate of a fuel assembly:receiving a fluid flow into the flow channel of the control rod guide tube through the one or more ports;providing the received fluid flow from the flow channel to the cavity of the fuel support; andproviding the fluid flow from the fuel support cavity to the lumen on the lower tie plate. 2. The method of claim 1, wherein receiving the fluid flow into the flow channel includes receiving the fluid flow from the one or more ports defined by the control rod guide tube for reducing flow asymmetries. 3. The method of claim 1, wherein the one or more ports are positioned at a substantial distance from the fuel support. 4. The method of claim 1, wherein providing the received fluid flow from the flow channel to the fuel support cavity includes ensuring that a cross-sectional area of the flow channel is less than or equal to a cross-sectional area of the fuel support cavity to minimize a pressure increase of the fluid flow between the flow channel and the fuel support cavity. 5. The method of claim 1, wherein the defining of the flow channel and the one or more ports includes ensuring that the one or more ports has a total cross-sectional area greater than a cross-sectional area of the flow channel, andwherein the cross-sectional area of the flow channel is less than or equal to a cross-sectional area of the fuel support cavity. 6. The method of claim 1, wherein the defining of the flow channel and one or more ports further comprises:defining a plurality of flow channels in the control rod guide tube;reducing a pressure drop of the fluid flow across the control rod guide tube by providing a plurality of ports in the control rod guide tube, for each flow channel. 7. The method of claim 1, further comprising:configuring the fuel support cavity to modify the fluid flow through the fuel support cavity. 8. The method of claim 1, wherein the control rod guide tube includes:a body having an axial length defining a lower end portion and the upper end portion of the control rod guide tube;a cavity within a substantial length of the body including one or more orifices at the upper and lower end portions of the body;a control rod chamber within the cavity for receiving a control rod;a plurality of ports coupled to the cavity and positioned at a substantial length from the upper end portion of the body; andat least two flow channels within the cavity extending a substantial portion of the axial length of the body. 9. The method of claim 8, wherein the control, rod chamber has a cruciform shape. 10. The method of claim 8, wherein each flow channel is fluidly coupled to one or more of the ports for receiving fluid flow from outside the body, andwherein each flow channel is fluidly coupled to an outlet proximate to the upper end portion of the body for providing the received fluid flow. 11. The method of claim 8, wherein the upper end portion of the body is configured to couple to a bottom of a fuel support for fluidly coupling each of the flow channels to an orifice of a fuel assembly cavity defined by the fuel support. 12. The method of claim 11, wherein a cross-sectional area of each flow channel is less than or equal to a cross-sectional area of the coupled orifice of the fuel assembly cavity. 13. The method of claim 10, wherein a combined cross-sectional area of the ports coupled to each flow channel is greater than a cross-sectional area of the coupled flow channel. 14. The method of claim 8, further comprising:an insert positioned within the cavity;wherein the insert defines the control rod chamber within the cavity, andwherein the insert defines each of the flow channels in conjunction with an inner surface of the body defining the cavity. 15. The method of claim 14, wherein the insert is rotatable within the cavity. 16. The method of claim 14, wherein the insert is fixedly attached to the inner surface of the body. 17. The method of claim 1, wherein the control rod guide tube includes:a body having a wall defining the upper end portion, a lower end portion, a cavity defined by an inner surface of the wall and extending from the upper end portion to the lower end portion, and a plurality of ports positioned axially along the wall between the upper end portion and the lower end portion for providing fluid flow into the cavity; andan insert dimensioned for positioning within the cavity, the insert having an upper end portion and a lower end portion and including a control rod chamber configured to receive a control rod and a plurality of channel fixtures that, at least partially, define one or more flow channels within the body cavity for receiving a fluid flow through one or more of the body ports; for channeling the received fluid flow within the body cavity between the lower end portion and the upper end portion, and for providing the fluid flow to the upper end portion of the body. 18. The method of claim 17, wherein each flow channel is configured to receive fluid flow from two or more ports. 19. The method of claim 17, wherein the insert is fixedly attached to the inner surface of the wall, andwherein each flow channel is substantially enclosed by a portion of the inner surface and a portion of the insert. 20. The method of claim 17, wherein the insert is rotatable within the body cavity. |
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047284820 | description | DETAILED DESCRIPTION The present method provides a method for inservice inspection of a pressure vessel wall with the lower internals maintained in position within the pressure vessel. A pressurized water nuclear reactor 1 is illustrated in FIG. 1, having a pressure vessel 3 which is of a generally cylindrical shape having a cylindrical wall 5, closed at the bottom by a bottom wall 7 of a hemispherical contour. The vessel is closed at the top by a flanged dome shaped head 9, which is secured, such as by bolts, to the top edge 11 of the pressure resistance wall 5, through the flanged portion 13, and is removable for refueling and inspection. The pressure resistant wall 5 has a plurality of inlet nozzles 15 and outlet nozzles 17, only one of each being shown, distributed about its periphery, a pair of each of such nozzles usually being provided. A nuclear core 19 is supported in the lower region of the pressure vessel 3, the core being supported in spaced relationship to the bottom wall 7 by a core barrel 21. The core barrel has a flange 23 which rests on a ledge 25 in the top inner surface of the pressure resistant wall 5. A core former 27 is situated about the lower region of the core barrel 21. The core includes a series of fuel assemblies 29 and thimbles 31 for receiving control rods, not shown, with at least one such thimble 31 adapted for insertion therein of an instrument for monitoring the operation of the core. The fuel assemblies and thimbles are mounted between a lower core plate 33 and an upper core plate 35. The control rods, as is known, may contain rod clusters of high and lower absorption cross-section for neutrons, and serve to reduce the thermal power of the reactor, or otherwise control the same, through monitoring by use of the instrument in the dedicated thimble therefor, or to shutdown the reactor. A lower core support plate 37 is provided with support columns 39, and a secondary core support 40 is also provided, as illustrated. These components comprise the lower internals 41. In the upper region of the pressure vessel 3, vertical guides 43 for the control rods and vertical guides for water displacement rods are provided, which generally comprise the upper internals 45. The lower internals 41, containing the core 19, and upper internals 45 are mounted generally coaxially within the pressure vessel 3. An annulus 47 between the core barrel 21 and the pressure resistant wall 5 provides for communication between the inlet nozzles 15 and the lower end of the core 19. Drive rods 49 from the control rods extend through seals 51 in the head 9. Drive mechanisms (not shown) are used to properly position the control rods, axially. Affixed about the periphery of the core barrel 21 there are neutron shields 53, and a plurality of radiation specimen pockets 55, with specimens (not shown) for monitoring radiation, insertable into said pockets. In practice, the width of the annular chamber 47, between the pressure vessel wall 5 and the core barrel 21 is about six inches (15.24 cm), while the width is narrowed in the area of the specimen pockets 55, the distance between the outer wall of the specimen pocket and the pressure resistant wall being about four inches (10.16 cm). The outer circumference of the pressure vessel wall 5 is about 540 inches (13.716 m), which gives an indication of the size of the pressure vessel and area that requires periodic inspection. In operation of the pressurized water reactor, coolant enters through the inlet nozzles 15 and flows downwardly through the annulus 47 to the bottom wall 7 and then upwardly through the core 19, into upper internals 45 and then transversely to, and outwardly from, the outlet nozzles 17, as indicated by the arrows shown in FIG. 1. The pressure vessel 3 is constructed from a plurality of sectional units which are integrally connected together by welds. Such welds, illustrated as welds 57, may comprise welds between the bottom head and wall 7 and the cylindrical wall 5, and welds between sections of the cylindrical wall 5, such as about nozzles 15 and 17, and intermediate welds. It is to the internal inspection of these welds and of the inner surface 59 of the pressure vessel 3 to which the present method is directed. As best illustrated in FIG. 2, the core barrel 21 has a cylindrical wall section 61 from which flange 23 extends outwardly. The flange 23 has a plurality of apertures 63 therethrough which communicate with the annulus 47 when the core barrel is positioned within the pressure vessel 3. The apertures 63 are normally closed with a removable plug 65. The purpose of the apertures 63, which normally have a diameter of about 2.25 inches (5.72 cm), is normally to enable the removal of specimens from the radiation specimen pockets 55 which are located about the outer periphery of the core barrel cylindrical wall section, eight of such apertures being illustrated in FIG. 2, although more or less than this number of apertures may be provided. As illustrated in FIG. 3, the removal of the head of the pressure vessel and the upper internals can be affected while leaving the core barrel 21 and core 19 intact in the pressure vessel 3. The present invention enables inspection of the welds 57 and internal surface 59 of the pressure vessel wall 5 while the reactor is in such a partially dismantled condition. When the head of the pressure vessel and upper internals have been removed, the vessel is normally flooded with water to protect against radioactivity, and the present method can be carried out under such flooded conditions. In the present method, access to the annular chamber is provided by either removing a plug from an existing aperture through the flange of the core barrel or by forming an additional aperture, which could be about 3.5 inches (8.89 cm) in diameter, through said flange, which additional aperture is subsequently provided with a plug, which may also be removable. A means for inspecting a weld or the inner surface of the pressure vessel wall is inserted through the access into the annular chamber and positioned proximate the area to be inspected. The means for inspecting, which is inserted through the access, may be an ultrasonic testing device, a visual examining means such as an optical scanning device, or other inspecting means, dependent upon the type of inspection desired. Ultrasonic testing, which is normally used, involves the injection of pulses of high frequency sound into the component to be tested. Any internal defects reflect sound back to the transmitting transducer, which then acts as a receiver. Such ultrasonic testing has a high sensitivity for cracks and other planar defects and can measure both length and height of a defect. Such testing is conventionally used on various components. The method is schematically illustrated in FIG. 4, wherein inspection of a weld 71 in the pressure vessel wall 5 is effected. As illustrated, the head 9 and upper internals 45 have been removed from the reactor but the lower internals 41 remain in position. After removal of any plug in the aperture 63 illustrated, an inspecting means 69, on the end of a positioning tool 67, such as a sonic tester, is inserted through the aperture 63 into the annular chamber 47, and is positioned proximate the inner surface of the weld 71. The actual positioning of the inspecting means relative to the area to be inspected will vary dependent upon the particular inspecting means used. The inspecting means is positioned at a location that is proximate the area, i.e. at a location that is sufficient to enable the desired examination or testing of the area by the inspecting means. The sonic tester, or sensor, 69 is then activated to inspect the weld 71, with such inspection being effected while the lower internals 41, including the core barrel 21, are still positioned within the pressure vessel 3. After the inspection, the inspecting means is retrieved through the aperture and the aperture plugged. The inservice inspections, of predetermined selected areas of the inner vessel surface and welds, according to the present method, may be made during normal plant outages such as refueling shutdowns or maintenance shutdowns occurring during a scheduled interval without the need for removing the lower internals from the pressure vessel. Since removal of the lower reactor internals is not required, labor savings are achieved and a reduction of man REM exposure is also achieved, as compared with previous processes that require the removal of the lower reactor internals. |
062787586 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT A preferred embodiment according to the present invention will be hereinafter described by reference to the accompanying drawings. FIG. 1 is a partial top view of a support grid 40 for a fuel assembly according to the present invention, and a plurality of straps 50, 60 are assembled in a crossed manner as described later so as to define a plurality of grid cells 41 positioned in a rectangular arrangement. In other words, the disposition of the grid cells 41 is a square arrangement such as 14.times.14, 15.times.15 and 17.times.17. Then, fuel rods 9 are individually placed through these grid cells 41 as shown by the dash-and-two-dot Sines and resiliently supported as in conventional grids. Shape of the blanks for the straps 50, 60 are partially shown in FIGS. 2a and 2b, respectively. Describing the structure of the strap 50 by reference to FIG. 2a, slits 51 are formed along an upper side edge of the strap 50 at intervals of a length corresponding to a distance between opposite sides of the grid cell 44. The slits 51 are designed to receive the straps 60 as the other member and extend perpendicular to a longitudinal axis of the strap 50. At an open-end side of the slit 51, a welding tab 57 is protrudingly formed, and another welding tab 59 is formed at the opposite side edge. Mixing vanes 53, 55 are formed at the both sides of the opening of alternate slits 51. Furthermore, welding apertures 53a, 55b are formed in the base end on either side of the slits 51 and outer side ends 53b, 55b are shaped in a curved outline. At the bend line 54 shown by dash-and-two-dot lines, the mixing vanes 53, 55 are individually bent in opposite directions so as to place the outer side ends 53b, 55b close to fuel rods 9 with narrow gaps as shown in FIG. 1. The structure of the strap 60 to be joined to the strap 50 is depicted in FIG. 2b. Slits 61 similar to the slits 51 are formed at intervals along a lower side edge thereof. Further, welding tabs 67 are protrudingly formed at the side of open ends of the slits 61. As can be understood from FIG. 1, the slits 51, 61 are each positioned at the crossing portion between the straps 50, 60 and mixing vanes 63, 65 are protrudingly formed at the upper side edge of the strap 60 in alignment with slits 61 corresponding to the slits 51 without the mixing vanes 53, 55. The mixing vanes 63, 65 each have the same shape as that of the mixing vanes 53, 55 and have a weld aperture 63a, 65a and curved outer side ends 63b, 65b and are each to be bent in opposite directions at the bend line 64 during assembly. This state is shown in FIG. 1. The support grids 40 of the structure as shown in FIG. 1 are joined into a fuel assembly, which is in turn loaded in a nuclear reactor core. During operation of the nuclear reactor, the coolant flows upwards (from the lower portion of the illustration to the top portion in FIGS. 2a and 2b) between the fuel rod 9 and the straps 50, 60 and a portion thereof impinges onto the slanted mixing vanes 53, 55, 63, 65 and caused to swirl, thereby promoting mixing of the coolant. In the aforementioned structure, the bend lines 54, 64 parallel to the longitudinal axis are located above the welding apertures 53a, 55a, 63a, 65a and so the apertures do not show in the horizontal plane of projection as is clear in FIG. 1. In other words, the coolant impinging area of the mixing vanes 53, 55, 63, 65 are larger than in the conventional structure, thereby promoting and increasing agitation and mixing of the coolant. Next, another embodiment according to the present invention will be described making reference to FIG. 3 and FIGS. 4a and 4b. FIG. 3 is a partial top view of a support grid 140 for a fuel assembly and a plurality of straps 150, 160 are similarly assembled in a crossed manner so as to define a plurality of grid cells 141 in a square arrangement. Moreover, the fuel rods 9 are individually placed through and resiliently supported in these grid cells 141 as shown by dash-and-two-dot lines. In FIGS. 4a and 4b, the blank shapes of the straps 150, 160 are partially shown. Describing the structure of the strap 150 by reference to FIG. 4a, slits 151 are formed along the upper side edge and each of them extends vertically in the drawing. Welding tabs 157, 159 are formed in a similar pattern. Mixing vanes 153, 155 are formed at both sides of the opening of alternate slits 151 and welding apertures 153a, 155a are defined in the base end of the vanes 153, 155 at either side of the slits 151. A gap 152 between the mixing vanes 153, 155 is slightly larger than that in the structure shown in FIGS. 2a and 2b. Moreover, at the bend lines 154 shown by dash-and-two-dot lines, the mixing vanes 153, 155 are bent in opposite directions, respectively, so as to be adjacent to the fuel rods 9 as shown in FIG. 3. The structure of the straps 160 to be combined with the straps 150 is depicted in FIG. 4b. Slits 161 similar to the slits 151 are defined at intervals along the lower side edge of the strap 160. Moreover, welding tabs 167, 169 are protrudingly formed as in the strap 60. In alignment with the slits 161 corresponding to the slits 151 without the mixing vanes 153, 155, mixing vanes 163, 165 are protrudingly formed at the upper side edge of the strap 160. In addition, as apparent from a comparison of FIG. 4a with FIG. 4b, the mixing vanes 163, 165 are larger in length than the mixing vanes 153, 155. They are bent in opposite directions at bend lines 164 that are slanted to avoid the welding apertures 163a, 165a during assembly. This state is illustrated in FIG. 3. The straps 150, 160 of the aforementioned structure are assembled to become the support grid 140 after the mixing vanes 153, 155, 163, 165 are bent from their blank state. In this situation, the mixing vanes 163, 165 slightly overlap adjacent grid cells 141 as shown in FIG. 3 while the mixing vanes 153, 155 do not overlap adjacent grid cells because the gap 152 is relatively large and so the assembly of the straps is not obstructed by the mixing vanes 153, 155. The support grid 140 is also, in a way similar to one for the support grid 40, joined into a fuel assembly, which is loaded in a nuclear reactor core. During the operation of the nuclear reactor, the coolant flows upwards between the fuel rod 9 and the straps 150, 160 and a portion of the coolant impinges on the slanted or bent mixing vanes 153, 155, 163, 165 to be agitated, thereby promoting mixing. In the abovementioned structure, since the length of the mixing vanes 163, 165 is larger than conventional ones, the total area of the projected plane of the mixing vanes 153, 155, 163, 165 becomes larger thereby making the coolant impinging surface larger than that in the prior art, increasing the effects of agitation and mixing. As described above, according to the present invention, since the area of the slanted portion of the mixing vanes protrudingly formed at side edges of the straps constituting a support grid in a horizontal projected plane is increased by displacement of the bend line from which the slanted portion begins or the increased length of all the mixing vanes, the effects of agitation and mixing in the coolant can also be increased. |
summary | ||
abstract | A method and a system evaluate an efficacy of a test suite in a model-based diagnostic testing system to determine a revision of the test suite. The evaluation comprises suggesting a test to be added to the test suite based on probabilities of a correct diagnosis and an incorrect diagnosis. The evaluation either alternatively or further comprises identifying a test to be deleted from the test suite based on probabilities of a correct diagnosis for the test suite and for a modified test suite that does not include a test. An efficacy value of each test in the test suite is computed. The system has a computer program comprising instructions that implement an evaluation of diagnostic efficacy of the test suite. The system either is a stand-alone system or is incorporated into the testing system. |
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description | This application is continuation-in-part and claims priority from U.S. application Ser. No. 13/441,109 filed Apr. 6, 2012, which is a divisional of U.S. application Ser. No. 12/449,087 filed on Jul. 8, 2009 (abandoned), which is a continuation-in-part from U.S. application Ser. No. 11/877,003 filed on Oct. 23, 2007 (abandoned), which is a continuation of U.S. application Ser. No. 11/094,304 filed Mar. 31, 2005 (abandoned). These applications are incorporated by reference. The United States Government has rights in this invention pursuant to Contract No. W-31-109-ENG-38 between the U.S. Department of Energy and the University of Chicago. The present invention relates to autonomous or nearly autonomous liquid-metal-cooled fast reactors for under-developed countries or for remote locations where the skilled labor pool is small. The lead-cooled fast reactor is one of six Generation IV nuclear systems selected by the Department of Energy for development. Design features that provide for near autonomous operation of the reactor also enhance safety and reduce costs, both Gen-IV goals. This invention relates to control and safety systems for near autonomous operation which exploit inherent feedback mechanisms to regulate power during both load change and upsets so that temperatures remain within safe limits with minimal need for active control system action. By way of definition, increasing autonomy is marked by a decreasing number of actuators and a migration of these actuators toward the balance of plant. The reactor design studied in this work originally appeared as the Secure Transportable and Autonomous Reactor-Liquid Metal (STAR-LM) concept proposed under the Nuclear Energy Research Initiative (NERI). The plant equipment is described in B. W. Spencer, “An Advanced Modular HLMC Reactor Concept Featuring Economy, Safety, and Proliferation Resistance,” Proceedings of the 8th International Conference on Nuclear Engineering, Apr. 2-6, 2000, Baltimore, Md. The primary features are a natural circulating primary system and an ultra-long life reactor core. The core lattice has a large coolant fraction resulting in a low pressure drop which enhances natural circulation. A combination of reduced power density and internal conversion allow for a core life of 10-15 years. These and other objects, aspects, and advantages of the present disclosure will become better understood with reference to the accompanying description and claims. An object of the present invention is to provide a highly autonomous modular nuclear plant in which the power of the nuclear reactor is controlled by the demand of the electric grid to which it is connected. The present invention provides a source of reliable nuclear generated electricity for under-developed countries or for remote locations where the skilled labor pool is small. Another object of the present invention is to provide a nuclear reactor coupled to a naturally circulating coolant loop. This is linked to a water/steam loop by means of a steam generating system. The water/steam loop consists of the steam generator, an electric power generating unit, together with a steam/water recirculating and steam control system. The electric power generator is coupled to supply power to a sink. Such exemplary sinks are an isolated external electric grid, an electric grid with multiple power suppliers, and one or more users of the power. As the power requirements of the sink change, a controller linked to a three way valve either increases or decreases the steam flow to the expansion turbine to boost or retard the power output. The three way valve is also in communication with a feedwater heater. This steam routing changes the pressure and temperature in the water/steam system which through the steam generator alters the flow rate and temperature of the coolant through the reactor coolant loop. The change in coolant flow alters the cooling of the reactor core which responds by increasing or decreasing its power output to restore a state of equilibrium to the nuclear power plant. The invention consists of certain novel features and a combination of parts hereinafter fully described, illustrated in the accompanying drawings, and particularly pointed out in the appended claims, it being understood that various changes in the details may be made without departing from the spirit, or sacrificing any of the advantages of the present invention. The invention is directed to a system for regulating nuclear reactor core reactivity. The nuclear reactor contains a nuclear reactor core as the source of thermal energy. The reactor core is the portion of the nuclear reactor containing the nuclear fuel components where the nuclear reactions take place generating fuel power. In a preferred embodiment, the core is nitride fueled. The nuclear core of the present invention has a negative temperature reactivity coefficient. A negative temperature reactivity coefficient relates to the nuclear core such that as the temperature of the nuclear core increases, the reactivity hence fuel power of the core decreases. The negative temperature reactivity coefficient characteristic is advantageous in that core reactivity establishes a reactivity and thermal equilibrium dependent upon coolant temperature and rate of flow through the coolant loop. Consequently, as the coolant temperature at an coolant inlet into the reactor drops and/or rate of coolant flow to the core increases, the core temperature decreases causing an increase in core reactivity and fuel power. Inversely, as coolant temperature increases and/or rate of coolant flow to the core decreases, the core reactivity and fuel power decreases. Accordingly, the invention comprises a naturally circulating nuclear reactor. Natural circulation relates to the ability of the core's coolant to cycle through the coolant loop unrestricted. In a preferred embodiment, the coolant is a liquid metal. In a more preferred embodiment, the coolant is a lead alloy. Most preferentially, the coolant is a lead bismuth eutectic. The coolant loop is unrestricted in that there are no physical pumps positioned or employed in the coolant loop to provide for or assist in the flow circulation. Instead, flow circulation is a function of removal of heat and the resulting coolant density change in the core coolant. The difference in density in the coolant establishes a thermal driving head which drives the coolant through the coolant loop as the relatively warm, less dense, coolant rises and the relatively cool, more dense, coolant drops. Necessarily, the coolant loop requires a reactor core to act as a heat source to heat the coolant and a steam generator as a heat sink to cool the coolant, where the heat sink is positioned above the heat source. As is illustrated in FIG. 1, the steam generator 113 is above the nuclear reactor core 111 in that the steam generator 113 is displaced relative to the nuclear reactor core 111 at a vector positive to a direction parallel to a gravity vector. In practice, reactivity in the fuel heats the core, which in turn heats the coolant. The heated and less dense coolant exits the reactor via a nuclear reactor coolant outlet. When the coolant flows through the coolant loop, the heat exchange from the coolant to the water/steam within the thermally coupled steam generator causes the coolant temperature to fall and density to rise. The relatively cool and dense coolant then drops through the coolant loop via gravity until it returns to reenter the reactor at the nuclear reactor coolant inlet which is positioned lower than the nuclear reactor coolant outlet. The cycle then repeats as the coolant is reheated. As stated above, the coolant loop is in thermal communication with a steam generator. Fundamentally the steam generator is a heat exchanger to transfer heat from the coolant loop to the steam/water piping system. The steam generator has a saturated liquid space and a steam space. At the interface of the saturated liquid space and steam space is an area which may be thought of as two phase region where the liquid in the saturated liquid space transitions to steam. The most significant heat transfer occurs in the two-phase region so the larger the two-phase region the greater the heat transfer between the coolant and the water/steam system. In operation, the steam generator is in fluid communication with a feedwater header and a steam piping system. Liquid enters the steam generator from the feedwater header, transitions to steam, and then exits the steam generator through the steam piping system. The steam generated in the steam generator exits the steam generator into the steam piping system. The steam piping system is in fluid communication with a three way valve at a three way valve inlet port. The three way valve may be any flow splitter type valve known in the art. As such, the three way valve has a three way valve first outlet and a three way valve second outlet. The three way valve is engineered in such a way that the steam entering the inlet port is divided between the first outlet and second outlet. The division of steam flow is directed by the three way valve such that the valve increases steam flow through the first outlet while concomitantly decreasing steam flow to the second outlet. Conversely, the three way valve can divide the steam such that steam flow to the first outlet is decreased with a concomitant increase in steam flow to the second outlet. An expansion turbine is in fluid communication with the three way valve at the three way valve first outlet port. Steam flow from the steam generator is controlled by the three way valve and enters the expansion turbine at a desired flow rate where it expands as required to perform work. The expansion turbine is also in fluid communication with a condenser. Once steam enters the expansion turbine and expands to do work, it continues to flow to the condenser. The condenser operates to remove heat from the steam after exiting the expansion turbine. Once heat is removed from the steam the steam condenses to water. A feedwater pump having a pump inlet is in fluid communication via a pump header with the condenser at the pump inlet. Water produced from steam in the condenser flows through the pump header to the feedwater pump, where it is accelerated then discharged through a pump discharge. In one embodiment, the pump is a constant speed pump. In a preferred embodiment, the pump is a centrifugal pump. A feedwater header is in fluid communication with the feedwater pump at the pump discharge. The feedwater header is also in fluid communication with the saturated liquid space of the steam generator. Water discharged by the feedwater pump is carried through the feedwater header to the steam generator to complete a water/steam circuit. A feedwater heater is in fluid communication at a heater inlet with the three way valve second outlet. Steam flow from the steam generator not directed to the expansion turbine is directed by three way valve to enter the feedwater heater at a desired steam flow rate. The feedwater is then in thermal communication with the feedwater header. Fundamentally the feedwater heater is a heat exchanger to transfer heat from the steam carried by the feedwater heater to the liquid in the feedwater header. The feedwater heater is then in fluid communication with the condenser at a heater outlet such that fluid exiting the feedwater heater enters the condenser where excess heat may be removed and the fluid is added back to the steam/water cycle. In an alternative embodiment, the feedwater heater is in fluid communication with the water/steam loop downstream of the condenser, such that fluid exiting the feedwater heater may be instead mixed directly with water forwarded to the steam generator rather than entering the condenser. Within this alternative embodiment, the fluid exiting the feedwater heater may enter into the water/steam loop in a fluid connection to the pump header or a fluid connection to the feedwater header. An electric generator is mechanically driven by the expansion turbine and is electrically connected to an electric-energy sink such as an electrical grid, such that mechanical energy produced by the expansion turbine is converted by the electric generator to electric energy to supply the sink. A controller is in data communication with the electric generator and the three way valve. The controller is programmed to respond to power drawn electric generator by directing the three way valve to increase or decrease steam shunted between the expansion turbine and the feedwater heater. The controller may function through several means known in the art. In one embodiment, an increase in electrical demand from the sink causes an increase in the electric power extracted from the generator. The controller senses the increase in electric power drawn from the generator and an error signal is form as the difference between the current demanded from the generator and the generator output. The error signal is sent to a feedback mechanism such as a proportional-integral controller which adjusts the three way valve in a way that increases steam flow to the turbine such that an increase in turbine power is realized to equalize the current demand from the sink and the current produced by the generator. In another embodiment, the controller senses a drop in voltage supplied from the generator as demand from the sink increases. An error signal is formed as the difference between the actual voltage generated and a desired value. The error signal is then sent to a proportional-integral controller which adjusts the three way valve in a way that increases steam flow to the turbine such that an increase in turbine power is realized for the generator to match the voltage output of the generator to the desired value. In another embodiment, the controller senses a drop in frequency when there is a decrease in generator rotational speed caused by an increase in mechanical load on the generator as more power is demanded from the sink. An error signal is formed as the difference between the actual frequency supplied by the generator and a predetermined frequency, for example 60 Hz. In response, the error signal is sent to a proportional-integral controller which adjusts the three way valve in a way that increases steam flow to the turbine such that an increase in turbine power is realized to return the generator speed back to that required to produce 60 Hz. The mode of operation of the invention relies on a relationship between the temperature and flowrate of the coolant flowing into the core and the core power. This relationship may be quantified in the reactivity balance where the reactivity D of the core is related to core flow rate, core inlet temperature, and externally imposed reactivity throughD=(P−1)A+(P/W−1)B+C*Tpc+δDext where P is normalized power, W is normalized reactor flowrate, *Tpc is change in core inlet coolant temperature, δDext is externally imposed reactivity, and A, B, C are integral reactivity feedback parameters that are measurable at the full power operating point. The values of A, B, C are such that a decrease in inlet temperature or an increase in liquid-metal flowrate increases reactivity. An increase in reactivity D increases core power. The object is for reactor power to follow the electric grid demand in a way that maintains temperatures within limits acceptable for normal operation. This is achieved in part through a lengthening of the steam generator two phase region and an elevating of the midpoint of this region in response to an increase in steam flow to the expansion turbine. Reactor coolant inlet temperature is lowered by increased two-phase heat transfer and reactor coolant flowrate is increased by greater buoyancy induced by greater separation of the core and steam generator thermal centers. If the integral inlet temperature coefficient, C, is negative and/or if the integral feedback flow coefficient, B, is negative, then the decrease in cold leg temperature and increase in coolant flow in response to the original increase in steam power add reactivity to the core. As a result, the core power increases. Ultimately the core finds a new steady-state condition for which the power is in equilibrium with the increase in steam power. This mode of operation is termed inherent boiling zone control since the only active control actions are a change in the three way valve position that led to the steam power increase and realignment of the condenser cooling water flowrate to match the new plant power. The natural circulation design option requires the coolant loop layout to support removal of reactor heat at normal operation with acceptable core outlet temperature. It is within the skill of the art to show that for unit core temperature rise, flowrate goes up linearly with the separation distance between steam generator and reactor thermal centers. Thus, the power that can be removed for fixed outlet temperature is proportional to the separation distance. The natural circulation design option then is limited in power to the extent the thermal separation is limited by costs or otherwise. Additionally, the feedwater temperature and flowrate to the steam generator are made to behave in a way that matches steam generator power to turbine power. The feedwater temperature will increase with a decrease in steam power to the expansion turbine when, as the first three way valve outlet is increasingly closed, steam is bypassed to the feedwater heater. The feedwater heater raises the temperature of the liquid flowing through the feedwater header into the steam generator. Consequently, the saturated liquid space shrinks in length causing the elevation of the two phase region in the steam generator to drop and there is a resulting coolant flow rate reduction, through reduced gravity head, which lowers reactor power. The feedwater flowrate will decrease as steam pressure increases when a centrifugal pump or other constant feed pump is used to deliver the feedwater. Referring now to FIG. 1, there is disclosed a schematic representation of a liquid-metal cooled nuclear powered electric generating system 110 powered by a nuclear reactor having a core 111. Liquid metal coolant is employed to cool the reactor core 111. The liquid metal coolant exits the core 111 through a line 112 and is transported to the steam generator 113 shown in dotted line and positioned above the reactor core. The water/steam portion 114 of the steam generator 113 is comprised of three phase regions: the steam space 115, the two-phase region 116, and the saturated liquid space 117 where the relative volume of these regions varies in relation to the heat flow from the liquid-metal side 118. The most significant heat transfer occurs in the two-phase region 116 so the larger the two-phase region the greater the heat transfer between the liquid-metal side 118 and the water side 114. Moreover, since the flow rate of the liquid-metal through the reactor is determined by the gravity head, the vertical position of the two-phase region 116 affects the flow of liquid metal through the reactor core 111 thereby controlling the heat transfer and power output. Steam from the steam generator 113 exits the steam generator through a steam piping system 119 and a portion is directed toward the expansion turbine 120. The flow rate to the steam turbine is controlled by the three way valve 121. After the steam expands as required to perform the work in the expansion turbine 120, it exits through a line 122 to a condenser 123 where heat is removed and water exits the condenser 123 through a pump header 124 to the feedwater pump 125. The feedwater pump 125 then returns water to the steam generator 113 via a feedwater header after passing through a thermally coupled feed water heater 126. The mechanical output from the expansion turbine 120 is directed to an electric generator 127 which is in turn is connected to a sink such as the electric grid 128. A sink such as the electric grid 128 demands power from the electric generator 127. A controller 129 linked to the three way valve 121 is able to sense the demand for increased power from the generator 127. As the demand for power from the electric grid 128 increases, the controller 129 responds by sending more steam to the expansion turbine 120 and less steam to the feedwater heater 126. This steam portioning increases the power output supplied by the generator 127 to the grid 128 and decreases the feedwater temperature as it recycles to the steam generator. In a reciprocal manner, as the power demand decreases, the controller 129 responds by altering the setting of the three way valve 121 to send less steam to the turbine 120 and shunt more steam to the feedwater heater 126. It should be noted that alteration of steam division changes the volume of the two phase region 116, ultimately leading to an increase in reactor power and a reestablishment of a reactor core equilibrium at a higher level. For the low power situation, the changes in physical parameters would be reversed leading to a decrease in the volume of the two phase region 116 of the water/steam line and subsequently a decrease in reactor power. Referring now to FIG. 2, a three way valve and controller function to regulate generator output in the presence demand from a sink. In the illustration, there is an increase in mechanical load on an expansion turbine 211 resulting when an electric grid 212 increases demand for electrical power from the electric generator 213. The controller unit 214 senses a drop in output frequency as generator speed drops due to the added mechanical load from the increase in demand. An error signal is formed as the difference between the actual output frequency and a desired frequency of 60 Hz. The error signal is sent to a feedback mechanism such as a proportional-integral controller 215 which moves a valve shaft 216 of a three way valve 217 in a way that concomitantly increases steam flow from a steam generator 218 to the expansion turbine 211 and decreases steam flow to a feedwater heater 219, such that an increase in turbine power is realized to return the generator speed back to that required to achieve a frequency of 60 Hz and power production from a reactor core reaches equilibrium with the demands of the electric grid 212. It is to be understood that the above-described arrangements are only illustrative of the application of the principles of the present invention and it is not intended to be exhaustive or limit the invention to the precise form disclosed. Numerous modifications and alternative arrangements may be devised by those skilled in the art in light of the above teachings without departing from the spirit and scope of the present invention. It is intended that the scope of the invention be defined by the claims appended hereto. In addition, the previously described versions of the present invention have many advantages, including but not limited to those described above. However, the invention does not require that all advantages and aspects be incorporated into every embodiment of the present invention. All publications and patent documents cited in this application are incorporated by reference in their entirety for all purposes to the same extent as if each individual publication or patent document were so individually denoted. |
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claims | 1. A device for stabilizing a dryer assembly in a reactor pressure vessel of a nuclear reactor, comprising:a reaction arm having a first end shaped and positioned to couple to a hold down bracket of a top head of the reactor pressure vessel, the reaction arm being rotatable about the first end; anda spring coupled to a second end of the reaction arm, the spring including an end configured to transfer a stabilizing force to a steam dryer support bracket of the dryer assembly relative to the reactor pressure vessel, the first end of the reaction arm, the second end of the reaction arm, and the end of the spring being substantially coplanar in a plane substantially perpendicular to an axis of rotation of the first end of the reaction arm. 2. The device of claim 1 wherein the spring is preloaded with a force of approximately 20,000 pounds for providing a continuous stabilizing preload to the dryer assembly during normal operations. 3. The device of claim 2 wherein the spring is preloaded for providing a downward preloaded force to the dryer assembly and against a steam dryer support bracket of the dryer assembly upon placement of the top head onto the reactor pressure vessel. 4. The device of claim 1 wherein the spring is positioned between the dryer assembly and a top head of the reactor pressure vessel and is configured for applying a downward force to the dryer assembly only during upward vertical movement of the dryer assembly. 5. The device of claim 1 wherein the spring is configured to be positioned between a hold down bracket of a top head of the reactor pressure vessel and a lifting rod of the dryer assembly and provides a downward stabilizing force to the lifting rod from the hold down bracket, and further configured to provide a stabilizing gap dimensioned to enable the spring to flex upwards into the stabilizing gap. 6. The device of claim 1, wherein the reaction arm includes a first portion configured to couple to the hold down bracket and a second portion configured to couple to the spring, the reaction arm configured for positioning the spring for providing a downward stabilizing force relative to the hold down bracket. 7. A device for stabilizing a dryer assembly in a reactor pressure vessel of a nuclear reactor, comprising:a reaction arm having a first end configured to couple to and rotate about a hold down bracket of a top head of the reactor pressure vessel, the reaction arm being rotatable about the first end; anda spring coupled to a second end of the reaction arm and positioned for preloading the dryer assembly with a force of approximately 20,000 pounds downward against a steam dryer support bracket of the reactor pressure vessel. 8. The device of claim 7 wherein the spring is a spiral spring. 9. The device of claim 7, wherein the spring is configured to be positioned within a gap between the dryer assembly and the hold down bracket when coupled to the second end of the reaction arm, the spring dimensioned to provide the gap for enabling the spring to flex upward and the dryer assembly to flex upward against the preloaded spring. 10. The device of claim 7 wherein the spring is configured for providing a downward force to the dryer assembly upon placement of the top head onto the reactor pressure vessel. 11. The device of claim 7 wherein the spring is configured to be positioned against a lifting rod of the dryer assembly and preloading the lifting rod downward. 12. The device of claim 11 wherein the lifting rod includes a lifting rod eye positioned about a top end of the lifting rod, and wherein the spring is positioned against the lifting rod eye and preloads the lifting rod eye downward. 13. The device of claim 7 wherein the reaction arm is configured to couple to the hold down bracket with a first fastener and to the spring with a second fastener. 14. The device of claim 13 wherein the first fastener is a bolt and the second fastener is a torque reaction lug. 15. The device of claim 14 wherein the reaction arm includes a first reaction arm portion and a second reaction arm portion positioned generally in parallel with the first reaction arm portion to form the reaction arm, the first reaction arm portion configured to be positioned on a first side of the hold down bracket and a first side of the spring and the second reaction arm portion configured to be positioned on a second side of the hold down bracket and a second side of the spring. 16. The device of claim 7, wherein the reactor pressure vessel includes four reaction arms, each reaction arm including support bracket coupled to one of four hold down brackets within the top head of a reactor pressure vessel and each support bracket being coupled to a one of four spiral springs. |
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042785006 | abstract | The primary coolant fluid is circulated within at least one primary loop between a steam generator and a pressure vessel. Within the steam generator, the primary fluid flows through a tube bundle and exchanges heat with a secondary fluid which is admitted in the form of water and discharged in the form of steam. After expansion within turbines and recovery in a condenser, the steam is then returned into the generator. The primary fluid circulating pumps are each driven by a turbine which is fed with steam taken directly from the steam generator or from the main outlet duct of the generator. The essential objective of the invention is to dispense with the use of an electric motor for driving each primary fluid circulating pump. |
claims | 1. An apparatus for processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, comprising:generation means for generating the image components due to the grid on the basis of data of the radiation image, comprisinganalysis means for analyzing the radiation image to obtain a spatial frequency and optionally an angle of a periodic pattern that the image components form,extraction means for extracting predetermined components containing the image components from the radiation image on the basis of an analysis result from said analysis means, andprocessing means for processing the predetermined components obtained by said extraction means to obtain the image components; andremoving means for removing the image components obtained by said processing means from the radiation image,wherein said processing means executes a process for estimating and processing an unsteady portion of the predetermined components from steady portions before and after the unsteady portion as the process of the predetermined components, andwherein said processing means estimates an amplitude and phase of a sine wave corresponding to the image components on the basis of the steady portions of the predetermined components before and after the unsteady portion, and the spatial frequency, and mends the unsteady portion on the basis of the estimation result. 2. The apparatus according to claim 1, wherein said generation means generates the image components on the basis of a feature that the image components superposed on the radiation image are steady. 3. The apparatus according to claim 1, wherein said extraction means executes filtering for extracting components having the spatial frequency obtained by said analysis means from the radiation image. 4. The apparatus according to claim 1, wherein said processing means detects the unsteady portion on the basis of envelope information of the predetermined components. 5. The apparatus according to claim 1, wherein said processing means obtains the image components by executing another filtering of the mended predetermined components. 6. The apparatus according to claim 1, wherein said processing means obtains the image components by substituting an unsteady portion that satisfies a predetermined condition in the unsteady portion by a predetermined value. 7. The apparatus according to claim 1, wherein said generation means executes the generation process of the image components for a predetermined line selected from the radiation image. 8. The apparatus according to claim 7, wherein said generation means executes the generation process for a resulted line data obtained by averaging a plurality of lines selected from the radiation image. 9. The apparatus according to claim 7, wherein said generation means executes the generation process of the image components for the predetermined line which is representative of a plurality of lines of the radiation image. 10. The apparatus according to claim 1, further comprising:image extraction means for extracting a partial image corresponding to a radiation irradiation field from the radiation image, andwherein said generation means generates the image components of the partial image obtained by said image extraction means. 11. The apparatus according to claim 1, further comprising:detection means for detecting if the grid is used for radiography, andwherein said generation means executes the process on the basis of a detection result of said detection means. 12. The apparatus according to claim 1, further comprising:image storage means for storing a radiation image obtained by removing the image components generated by said generation means from the radiation image. 13. The apparatus according to claim 1, further comprising:image sensing means for capturing the radiation image by a solid-state image sensing element having an image-receiving surface with a size within which a spatial distribution of radiation, at the image-receiving surface, corresponding to the object to be captured falls. 14. The apparatus according to claim 1, further comprising:image component storage means for storing the image components generated by said generation means. 15. A system for processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, said system being built by connecting a plurality of apparatuses intercommunicatably, comprising:generation means for generating the image components due to the grid on the basis of data of the radiation image, comprisinganalysis means for analyzing the radiation image to obtain a spatial frequency and optionally an angle of a periodic pattern that the image components form,extraction means for extracting predetermined components containing the image components from the radiation image on the basis of an analysis result from said analysis means, andprocessing means for processing the predetermined components obtained by said extraction means to obtain the image components; andremoving means for removing the image components obtained by said processing means from the radiation image,wherein said processing means executes a process for estimating and processing an unsteady portion of the predetermined components from steady portions before and after the unsteady portion as the process of the predetermined components, andwherein said processing means estimates an amplitude and phase of a sine wave corresponding to the image components on the basis of the steady portions of the predetermined components before and after the unsteady portion, and the spatial frequency, and mends the unsteady portion on the basis of the estimation result. 16. A method of processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, comprising:a generation step of generating the image components due to the grid on the basis of data of the radiation image, comprisingan analysis step of analyzing the radiation image to obtain a spatial frequency and optionally an angle of a periodic pattern that the image components form,an extraction step of extracting predetermined components containing the image components from the radiation image on the basis of an analysis result from said analysis step, anda processing step of processing the predetermined components obtained in said extraction step to obtain the image components; anda removing step of removing the image components obtained in said processing step from the radiation image,wherein said processing step includes executing a process for estimating and processing an unsteady portion of the predetermined components from steady portions before and after the unsteady portion as the process of the predetermined components,wherein said processing step includes estimating an amplitude and phase of a sine wave corresponding to the image components on the basis of the steady portions of the predetermined components before and after the unsteady portion, and the spatial frequency, and mending the unsteady portion on the basis of the estimation result, andwherein said steps are performed using a processor and a memory. 17. A computer-readable storage medium storing, in executable form, a program for making a computer execute a method of processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, said method comprising:a generation step of generating the image components due to the grid on the basis of data of the radiation image, comprisingan analysis step of analyzing the radiation image to obtain a spatial frequency and optionally an angle of a periodic pattern that the image components form,an extraction step of extracting predetermined components containing the image components from the radiation image on the basis of an analysis result from said analysis step, anda processing step of processing the predetermined components obtained in said extraction step to obtain the image components; anda removing step of removing the image components obtained in said processing step from the radiation image,wherein said processing step includes executing a process for estimating and processing an unsteady portion of the predetermined components from steady portions before and after the unsteady portion as the process of the predetermined components, andwherein said processing step includes estimating an amplitude and phase of a sine wave corresponding to the image components on the basis of the steady portions of the predetermined components before and after the unsteady portion, and the spatial frequency, and mending the unsteady portion on the basis of the estimation result. 18. An apparatus for processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, comprising:analysis means for analyzing the radiation image to obtain a spatial frequency of grid stripes generated on the radiation image due to the grid;first extraction means for executing frequency band limitation processing on the radiation image on the basis of the spatial frequency to extract the processed radiation image as first grid stripes components;calculation means for calculating an amplitude value of the first grid stripes components;second extraction means for substituting a value of the first grid stripes components corresponding to an image portion where the amplitude value is out of a predetermined range with a value estimated from a value of the first grid stripes components corresponding to an image portion where the amplitude value is within the predetermined range to extract the substituted image as second grid stripes components; andremoving means for removing the second grid stripes components from the radiation image. 19. A system for processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, said system being built by connecting a plurality of apparatuses intercommunicatably, comprising:analysis means for analyzing the radiation image to obtain a spatial frequency of grid stripes generated on the radiation image due to the grid;first extraction means for executing frequency band limitation processing on the radiation image on the basis of the spatial frequency to extract the processed radiation image as first grid stripes components;calculation means for calculating an amplitude value of the first grid stripes components;second extraction means for substituting a value of the first grid stripes components corresponding to an image portion where the amplitude value is out of a predetermined range with a value estimated from a value of the first grid stripes components corresponding to an image portion where the amplitude value is within the predetermined range to extract the substituted image as second grid stripes components; andremoving means for removing the second grid stripes components from the radiation image. 20. A method of processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, comprising:an analysis step of analyzing the radiation image to obtain a spatial frequency of grid stripes generated on the radiation image due to the grid;a first extraction step of executing frequency band limitation processing on the radiation image on the basis of the spatial frequency to extract the processed radiation image as first grid stripes components;a calculation step of calculating an amplitude value of the first grid stripes components;a second extraction step of substituting a value of the first grid stripes components corresponding to an image portion where the amplitude value is out of a predetermined range with a value estimated from a value of the first grid stripes components corresponding to an image portion where the amplitude value is within the predetermined range to extract the substituted image as second grid stripes components; anda removing step of removing the second grid stripes components from the radiation image,wherein said steps are performed using a processor and a memory. 21. A computer-readable storage medium storing a program for making a computer execute a method of processing a radiation image of an object obtained by radiography using a grid which is used to remove scattered radiation from the object, to remove image components due to the grid, said method comprising:an analysis step of analyzing the radiation image to obtain a spatial frequency of grid stripes generated on the radiation image due to the grid;a first extraction step of executing frequency band limitation processing on the radiation image on the basis of the spatial frequency to extract the processed radiation image as first grid stripes components;a calculation step of calculating an amplitude value of the first grid stripes components;a second extraction step of substituting a value of the first grid stripes components corresponding to an image portion where the amplitude value is out of a predetermined range with a value estimated from a value of the first grid stripes components corresponding to an image portion where the amplitude value is within the predetermined range to extract the substituted image as second grid stripes components; anda removing step of removing the second grid stripes components from the radiation image. |
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046438459 | abstract | A method of and an apparatus for cutting a high-activity solid waste, such as a used channel box and a used control rod, to reduce same in size to facilitate its disposal. The channel box is cut axially through opposing corners to produce elongated split portions of an L-shaped in cross section, and the control rod is cut axially through a central portion to produce elongated split portions of an L-shape in cross section. The portions obtained by cutting the channel box and control rod are substantially similar in shape and facilitate storing. |
summary | ||
047056622 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 diagrammatically shows the integrated primary circuit of a fast neutron nuclear reactor constructed in accordance with the invention. In per se known manner for an integrated fast neutron nuclear reactor, FIG. 1 shows that the complete primary circuit of the reactor according to the invention is placed within a vertically axed cylindrical vessel 10, called the main vessel, which is sealed at its upper end by a concrete-filled sealing slab 12. The main vessel 10 and its sealing slab 12 are placed in a vessel well or shaft 14 formed in a concrete enclosure 16. In per se known manner, vessel 10 can be supported either by placing it on the bottom of the vessel shaft 14, or by suspending the vessel on the upper part of the vessel shaft, as illustrated in FIG. 1. The reactor core 18 is placed in the central part of vessel 10 and rests on the bottom of the latter via a support structure having a support member 20, which also serves to supply the core with liquid sodium. Vessel 10 is filled up to a level N with a certain volume of liquid metal 22, generally constituted by sodium. The sodium 22 is surmounted by a covering of neutral gas 24, generally constituted by argon. An inner vessel 26, e.g. having a step, defines within the main vessel 10 a hot collector 28 positioned above core 18 and vessel 26 and an annular cold collector 30 surrounding core 18. According to the invention, a certain number of steam generators 32 traverse the sealing slab 12 and pass into the peripheral part of the main vessel 10, whilst passing through the inner vessel 26. The hot sodium leaving the reactor core 18 and entering the hot collector 28 penetrates the steam generators 32. On passing through the latter, it cools by heat exchange with the water of the water/steam circuit, before passing out into the cold collector 30. In per se known manner, the reactor also comprises a certain number of primary pumps 34 placed within the cold collector 30. These primary pumps 34 suck the relatively cold primary sodium transferred into collector 30 by steam generators 32 and then deliver it to the supply support member 20 of reactor vessel 18 by means of pipes 36. As is more particularly illustrated in FIG. 2, the aforementioned primary circuit can comprise three primary pumps 34, six steam generators being associated with each of these pumps. However, this construction is obviously not limitative and a different number of pumps and steam generators can be considered. As in existing integrated reactors, the components within the reactor vessel, i.e. the steam generators 32 and pumps 34 can be supported either by suspending these components on the sealing slab 12, or by having them rest on the bottom of vessel 10, as is diagrammatically shown in FIG. 1. With reference to FIG. 3, a description will now be given of a preferred embodiment of the steam generator 32. Thus, it can be seen in FIG. 3 that the steam generator 32 has a generally vertical configuration and passes through a passage 38 formed in slab 12, so that the actual exchange zone is introduced directly into the sodium 22 contained in the reactor vessel, whereas the connection of the steam generator to the water/steam circuit of the reactor and to a neutral gas secondary circuit is brought about by a part of the generator positioned above slab 12. Said part of the steam generator positioned above slab 12 and which constitutes the head of said component comprises, starting from the top, a water intake chamber 40, a steam discharge chamber 42 and a helium discharge chamber 44, in the considered example where helium constitutes the chemically neutral gas ensuring the heat transfer between the primary sodium and the water of the water/steam circuit. However, the chemically neutral gas can be of a different nature and can even be a gaseous mixture. The water intake chamber 40 is formed between the upper dome 46 of the steam generator and an upper horizontal tube plate 48. Dome 46 carries a water intake pipe 50, which is used for connecting the chamber 40 to the water/steam circuit, as will be shown hereinafter. The upper tube plate 48 supports a group of vertical straight tubes 52, which are open at their lower ends and extend freely downwards from plate 48 to the interior of the reactor vessel. The steam discharge chamber 42 is formed within a cylindrical wa11 51 extending dome 46 downwards between the upper tube plate 48 and an intermediate horizontal tube plate 54. A steam discharge pipe 56 is provided in wall 51 to permit the connection of the steam discharge chamber 42 to the water/steam circuit. The intermediate tube plate 54 supports a group of vertical straight tubes 58, called intermediate tubes and which are arranged coaxially around each of the inner tubes 52, so as to define with the latter annular passages 60. As is shown in FIG. 3, the intermediate tubes 58 are sealed at their lower ends, in the vicinity of which issue the inner tubes 52. A circulation of water and then steam takes place from the water intake chamber 40 to the steam discharge chamber 42. Thus, the water admitted into chamber 40 descends within the inner tube 52, before rising through annular spaces 60 up to the steam discharge chamber 42, as illustrated by the arrows F.sub.1 in FIG. 3. The helium discharge chamber 44 is also formed within a cylindrical wall 62 positioned in the extension of wall 51 and is defined between the intermediate tube plate 54 and a horizontal lower tube plate 64. A discharge pipe 66 formed in wall 62 makes it possible to connect the helium discharge chamber 44 to the helium circuit which will be described hereinafter. The lower tube plate 64 supports a third group of vertical straight tubes 68 positioned coaxially around each of the intermediate tubes 58, so as to define with the latter a second series of annular spaces 70. As is shown in FIG. 3, the outer tubes 68 are extended downwards slightly beyond the bottom of the intermediate tubes 58, so as to be connected to helium intake tubes 72 having a reduced diameter and used for introducing the helium through the bottom of each of the tubes 68, so that this neutral gas rises through the annular spaces 70 into the helium discharge chamber 44. Although it is theoretically possible to envisage the introduction of the helium by tubes 72 preferably passing through the bottom of the reactor vessel, the most realistic solution shown in FIG. 2 consists of raising the tubes 72, outside the actual exchange group, to above slab 12, in order to ensure the independence of the generator from the main vessel. In practice, FIG. 3 shows that the steam generator has for this purpose a widened area defined by a widened wall 74, which extends downwards from the lower tube plate 64 to the vicinity of the upper face of sealing slab 12. A certain number of helium intake pipes 76 (12 in the embodiment shown in FIG. 6) are installed on wall 74 and support small tube plates 78, to each of which are connected a certain number of tubes 72. These tubes 72 are then grouped and extend by a vertical straight portion 72a down and along the exchange group. This straight portion 72a is extended below the lower end of the outer tubes 68. It is connected to the lower end of the outer tubes 68 by a substantially horizontal portion 72b located below the lower end of tubes 68. Portion 72b permits a certain differential expansion between outer tube 68 and the helium intake tubes 72. The circulation of helium within the steam generator is diagrammatically represented by arrow F.sub.2 in FIG. 3. The exchange structures of the steam generators 32 used in the reactor according to the invention are consequently constituted by a group of elementary vertical tubular cells introduced into the liquid sodium 22 contained in the reactor vessel 10. In this structure, tubes 52, 58 and 68 constituting each of the said tubular cells are independent of one another, particularly from the standpoint of the differential expansions between the tubes and between the group and its outer casing. This feature is of fundamental importance for the mechanical reliability of the tubes and the steam generators overall. Associated with the use of a chemically neutral intermediate fluid, such as helium circulating between the sodium and the water/steam and at a temperature between that of these two fluids, it makes it possible to obtain steam generators with high mechanical integrity and reliability. Moreover, this so-called plunging tube structure makes it possible to obtain an acceptable thermal efficiency for an exchange height permitting the installation of steam generators within the reactor vessel. Thus, for a reactor of 1500 MWe with steam generators integrated into a main vessel with an internal diameter below or equal to 25 m and a height below or equal to 18.5 m, for each generator there is a heat exchange height between 15 and 17 m and an external diameter of the group of approximately 1.75 m. The tube length would be reduced if the inner tubes 58 had fins placed in annular spaces 60 by which the steam rises. In order to provide a good understanding of the structure of the steam generator used in the reactor according to the invention and particularly in order that the circulations of the different fluids in the tubes are more readily apparent, FIG. 3 shows a single elementary tubular cell, by considerably increasing the diameter of the tubes forming said cell compared with the dimensions of the steam generator. Obviously, in practice, there are numerous elementary cells of the type shown in FIG. 3. These different cells define the actual exchange zone of the steam generator. In order to channel the outflow of sodium 22 around the various tubular cells of the steam generator, the latter also has a ferrule 80 encircling the group of tubular cells, whereby said ferrule can e.g. be connected to the widened part 74 of the head of the steam generator. Ferrule 80 tightly traverses the inner reactor vessel 26, via an argon chamber sealing system 82. The sodium contained in the hot collector 28 enters the ferrule 80 via intakes 84 formed therein. Preferably, the intakes 84 are positioned above the level N of sodium 22 in the reactor vessel, in such a way that the sodium level N' within the steam generator is higher than said level N. Thus, it is possible to significantly increase the heat exchange length between the sodium and the water of the water/steam circuit, which makes it possible, as stated hereinbefore, to limit the height of the reactor vessel to a reasonable value. The use of gilled or finned steam-helium tubes makes it possible to maintain the entry level of the sodium into the steam generator immediately below the main vessel, with a view to ensuring a better thermal protection of the latter. The circulation of sodium from the hot collector 28 to the inlets 84 takes place within a collar 86 introduced into sodium 22 and surrounding the ferrule 80 to above the inlets 84. In the lower part of the steam generator, the sodium flows out into the cold collector 30 directly via the lower open end of ferrule 80. The sodium circulation in the steam generator is diagrammatically represented by the arrows F.sub.3 in FIG. 3. In addition, the circulation of helium in the intermediate space 70 can make it possible to discharge residual heat dissipated in the reactor core, in the case of a stoppage of primary pumps 34 (FIG. 1). The cooling exchangers of the reactor when shut down and which are conventionally used can consequently be eliminated. In order that there can be a circulation of sodium 22 by natural convection within ferrule 80, on the latter and below sodium level N is provided a normally closed entry window 85. The opening of this window is automatically controlled when the pumps 34 are stopped. As a variant, calibrated windows only permit the passage of a reduced primary sodium flow under normal operating conditions and can also be used. FIG. 3 diagrammatically shows a first possible way of supporting steam generator 32, in which the latter rests directly on the bottom of reactor vessel 10. This supporting action, which makes it possible to prevent overloading of the sealing slab 12, takes place by means of a ferrule or legs 88 fixed to the lower end of the ferrule 80 and resting directly or indirectly on the bottom of the reactor vessel. In order to take account of the differential expansion, the sealing at the point where the steam generator traverses slab 12 is e.g. brought about with the aid of a bellows system 90, whose ends are respectively fixed to the slab and to collar 86. Preferably and as is also illustrated in FIG. 3, the upper part of the steam generator carrying the inner and intermediate tubes 52, 58 is dismantlable, so that it is easier to control and inspect the outer tubes 68 which are the only ones which are in direct contact with the primary sodium. For this purpose, the intermediate tube plate 54 is extended outwards beyond wall 52, in order to form a flange 96 resting on a flange 98 formed at the upper end of wall 62. These two flanges are fixed tightly to one another by per se known fixing means and in particular by bolts 100 traversing holes regularly distributed over the circumference of flanges 96, 98. FIG. 4 is a view comparable to that of FIG. 3 showing the configuration of the exchanger after removing the upper dismantlable part incorporating the water intake chamber 40 and the steam discharge chamber 42, as well as the inner and intermediate tubes 52, 58 associated therewith. FIG. 4 shows that there is a completely free access to outer tubes 68. This possibility of dismantling the steam generator also has the advantage of facilitating the manufacture of the generator in the factory, its transportation, its on site installation, together with subsequent checks and repairs. Moreover, an inspection of each of the tube plates is also made possible by the access ports 41, 43 and 45 respectively formed in the dome 46 and in walls 51 and 62. It was stated hereinbefore that the water of the water/steam circuit descends by inner tubes 52 before rising again into the annular spaces 60 defined around the said tubes. Thus, there is a possibility of a parasitic heat exchange between the relatively cold water entering by tubes 52 and the steam rising again via spaces 60. In order to greatly reduce this parasitic heat exchange, outer tube 52 has two coaxial walls 52a, 52b defining between them an annular zones 52c. In the embodiment of FIG. 3, this annular zone is sealed at its upper end level with the tube plate 48 and is open at its lower end. It is consequently filled with stagnant steam and water ensuring a satisfactory thermal insulation, particularly in the upper part of the group, in which such insulation is particularly necessary. In the preferred embodiment shown in FIG. 3, the outer wall 52b of the inner tube 52 is extended downwards beyond its inner wall 52a, which enables the water reaching the lower end of the tubes to undergo a speed reduction. The latter is favourable to the change of direction of the water occurring in the bottom of the intermediate tubes 58. In a constructional variant shown in FIG. 5, the inner tubes 52 also have an inner wall 52a and an outer wall 52b separated by an annular zone 52c. However, this annular zone is sealed in this case at the lower end of inner tube 52a, which is also positioned above the lower end of outer tube 52b. However, zone 52c is open towards the top at the tube plate 48. In this case, the steam generator dome 60 is eliminated and replaced by a cylindrical ferrule 146 extending wall 51 in the upwards direction and opening at its upper end into the reactor building. Thus, zone 52c is filled with a gas at atmospheric pressure and generally by the air above the steam generator. Ferrule 146 supports tube plates 151 to which are connected intake pipes 150. The inner walls 52a of the inner tubes 52 are extended above the upper tube plate 48 and are connected to tube plates 151, a certain number of walls 52a being fixed to each of the plates. FIG. 5 also shows a second variant for the supporting of the steam generator, in which the latter is suspended on slab 12. In this case, the widened wall 74 has at its lower end a flange 92 resting on a flange 94 fixed to slab 12 around opening 38. In practice and although this is shown in the drawings, it is desirable to have a centering of the inner tubes 52 within the intermediate tubes 58, as well as a centering of the latter within the outer tubes 68. This centering can be brought about by any appropriate known means and particularly as a result of helical devices fixed to the outside of tubes 52, 58, at least in the lower part thereof, permitting the upward flow of the steam and the helium and improving the heat exchange. The following remarks can be made regarding the dimensioning of tube plates 48, 54, 64. The upper tube plate 48 is traversed by small diameter inner tubes 52. It is at a moderate temperature corresponding to the temperature of the feed water on entry, i.e. approximately 260.degree. C. Finally, it is subject to a service pressure corresponding to the pressure drop of the water/steam circuit. Thus, plate 48 has a moderate thickness. The intermediate tube plate 54 is traversed by intermediate tubes 58 having modest dimensions compared with the spacing of the system. This plate is subject to a mean operating pressure corresponding to the pressure difference between helium and steam, for a moderate temperature of approximately 450.degree. C. corresponding to the steam discharge temperature. It is consequently slightly thicker than tube plate 48. Finally, the lower tube plate 64 is the only one which is perforated with the diameter of the outer tube 68. It is exposed to temperature and pressure conditions closed to those of the intermediate tube plate 54 (differential pressure approx. 7 MPa and helium discharge temperature substantially below 500.degree. C.). Thus, it should be noted that the helium temperature drops by about 50.degree. between the sodium level and the tube plate 64, due to the helium/steam heat exchange taking place over close to 3 meters. This phenomenon is very favourable for the behaviour of the tube plate 64. It has already been pointed out that the steam generator 32 used in the reactor according to the invention has a generally vertically axed cylindrical configuration. However, it is apparent from the preceding description, that there is a problem of fitting the vertical parts 72a of the helium supply tube 72 relative to the generator tubular cells. In a preferred, but not limitative configuration according to FIG. 6, this is brought about by providing within the ferrule 80 which has a generally cylindrical configuration, three passage zones 83 for the tubes 72 extending over the entire height of ferrule 80 and at 120.degree. from one another. The zones 83, which have a substantially triangular cross-section, are defined by the cylindrical ferrule 80 and by vertical partitions 81 having a V-shaped section. More specifically, the branches of the V formed in section by each of the partitions 81 are parallel to the adjacent branches of two other partitions 81. Thus, the group of tubular cells has in section the shape of a star with three branches. It should be noted that the distribution of tubes on plates 48, 54 and 64 is favourable from the dimensioning standpoint. Preferably, in order to avoid a parasitic heat exchange between the hot sodium entering the upper part of ferrule 80 and the cooled sodium leaving the lower part thereof, the zones 83 housing the helium supply tubes 72 are maintained under an atmosphere corresponding to that of the neutral gas covering 24 of the reactor. This result is obtained by sealing the lower end of each of the zones 83 by a wall tightly traversed by tubes 72. Moreover, said zones 83 acn be thermally insulated with respect to the interior of ferrule 80, as a result of appropriate, not shown thermal insulation systems. Another solution consists of allowing the sodium to flow out at a reduced speed in said zones, whilst heating the descending helium. There is scarcely any difference between the zones and other zones located in the sodium vessel. The steam generator integrated with the reactor vessel according to the invention consequently has a structure such that all the fluids descending into the main vessel are contained in tubular structures which are simple when considered in isolation, said structures being moderately stressed under well known and highly divided conditions, which considerably reduces the risks linked with cracks and possible leaks. Moreover, leak detection is easy, e.g. by checking the neutral gas atmosphere in the steam generator, above sodium level N' and which would be entered by the helium from leaks or in the neutral gas covering of the reactor. Moreover, apart from its functions of transferring heat and acting as an intermediate medium between the sodium and the steam, the helium acts as a heat absorber between these two fluids, which is important for fast neutron reactors. FIG. 7 shows diagrammatically and in a non-limitative manner, all the water/steam and helium circuits of the reactor 5 according to the invention. The water/steam circuit comprises a steam pipe 102 by which the steam discharge pipe 56 of each of the steam generators 32 communicates with the high pressure stage 104a of a steam turbine and which also has a medium pressure stage 104b and a low pressure stage 104c. The steam leaving the high pressure stage 104a by a pipe 105 traverses a drier 106, a medium pressure resuperheater 108 by drawing off steam and a resuperheater 110 by helium, before being directed to the medium pressure stage 104b of the turbine. Finally, the steam is directed to the low pressure stage 104c by a pipe 112 before being condensed in a condenser 35. The condensation water returns to the intake pipes 50 of steam generators 52 by a duct 116. According to the interesting feature of the invention, the resuperheater 110 traversed by the steam introduced into the medium pressure stage 104b of the turbine simultaneously cools the helium circulating in the intermediate circuit of the steam generators 32. To this end, FIG. 7 shows that the helium circuit comprises a duct 118 connecting the helium discharge pipes 66 of the steam generators to the intake of the resuperheater 110 and a return duct 120 connecting the outlet of the resuperheater 110 to the helium intake pipes 76 of the steam generators. The helium circulation is controlled by a blower 122 arranged in the return duct 120. A system 124 for the treatment and in particular the drying of the helium is connected in parallel on blower 122. According to a variant of the above arrangement, another use of the heat carried by the helium is proposed. This is shown in FIG. 8, where it is possible to see a neutral gas circuit integrated into the reactor building and used for a supplementary reheating of the feed water and a supplementary production of live steam. This circuit comprises a water heater 126 and an additional steam generator 128. The use of helium in the intermediate circuit of steam generators is of particular interest. Thus, this gas is chemically neutral and reacts neither with the sodium nor with the water. Moreover, it does not require the use of steels differing from those normally used at the high operating temperatures of a steam generator. Finally, it has good thermal and thermodynamic characteristics. In particular, its thermal conductivity is much higher than that of air or carbon dioxide gas and its specific heat is more than four times that of liquid sodium whereas its specific gravity is 120 to 150 times lower. Helium can be used on an industrial scale and its flow is only slightly prejudicial to structures. Moreover, in the special case of the invention in which the steam generator is integrated into the reactor vessel, it has the advantage of not absorbing neutrons. |
abstract | An eddy-current flaw detector includes a trace data calculator configured to calculate each coordinate with respect to flaw detection points on which an inspection probe is used upon performing an eddy-current testing based on an inputted condition of eddy-current flaw detection and surface shape data of an inspection-object surface measured by a profilometer, and to calculate a normal vector of each flaw detection point; a gap evaluation calculator configured to acquire an evaluation result on a gap between the inspection-object surface and the inspection probe for each flaw detection point; a flaw detection data collector configured to acquire flaw detection data of an inspection object for each flaw detection point; a flaw detection data analyzer configured to evaluate presence/absence of a flaw in the inspection-object surface based on the flaw detection data of the inspection object and the evaluation result on the gap for each flaw detection point. |
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060751762 | claims | 1. A method of immobilizing mixed low-level waste, which comprises: (a) providing iron oxide powders; (b) providing said low-level waste as a powder; (c) forming a solution comprising orthophosphoric acid; (d) mixing said waste powder, said iron oxide powders and said solution in mass % of waste powder:iron oxide powders:acid solution=30-60:15-10:55-30 to form a slurry; (e) blending said slurry to form a homogeneous mixture; and (f) curing said homogeneous mixture at room temperature to form a final product. (a) providing a mixed low-level waste powder; (b) providing a predetermined amount of magnetite powder; (c) forming an acid solution of orthophosphoric acid and ferric oxide; (d) mixing said low level waste powder, said magnetite powder, and said acid solution in mass % of waste powder:magnetite powder:acid solution=30-60:15-10:55-30 to form a slurry; (e) blending said slurry to form a homogeneous mixture; and (f) curing said homogeneous mixture at room temperature to form a final product. 2. The method according to claim 1, wherein said iron oxide powders comprise a mixture of iron oxide powders having ratios in mass % FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 =25-40:40-10:35-50. 3. The method according to claim 1, wherein said iron oxide powders comprise metallurgical cinder. 4. The method according to claim 3 wherein iron oxides are added to said cinder to adjust the ratio to be FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 =25-40:40-10:35-50. 5. The method according to claim 1, wherein said iron oxide powders comprise magnetite. 6. The method according to claim 5 wherein said solution comprising orthophosphoric acid further comprises ferric oxide. 7. The method according to claim 1, wherein said low-level waste is mixed with said iron oxide powders to form a powder phase and said orthophosphoric acid solution is added to said powder phase. 8. The method according to claim 1, wherein said low-level waste is mixed with said orthophosphoric acid solution, and said iron oxide powders are then added. 9. The method according to claim 1, wherein the amount of said low-level waste in said final product is 60 mass %. 10. The method according to claim 1, wherein the amount of said low-level waste in said final product is 30-40 mass %. 11. The method according to claim 1, wherein said mass % of the acid solution in said slurry is 50 mass %. 12. The method according to claim 1, wherein said mass % of the acid solution in said slurry is 55 mass %. 13. The method according to claim 1, wherein the orthophosphoric acid content, without water, of said slurry is 25-35 mass %. 14. The method according to claim 1, wherein the concentration of said orthophosphoric acid used in step (c) is from 50-83%. 15. A method of immobilizing mixed low-level waste, which comprises: 16. The method according to claim 15, wherein said acid solution is added to a powder phase of said waste powder and said magnetite powder in step (d). 17. The method according to claim 15, wherein said acid solution of orthophosphoric acid and ferric oxide is first mixed with said waste powder, and then is added to said magnetite powder in step (d). 18. The method according to claim 15, wherein the amount of said low-level waste in said slurry is 30 to 40 mass %. 19. The method according to claim 15, wherein the mass % of said acid mixture in said slurry is 50 to 55 mass %. |
claims | 1. A terahertz wave generation device for generating terahertz waves in a direction satisfying non-collinear phase matching conditions by projecting ultra-short pulse laser light on a non-linear optical crystal, the terahertz wave generation device comprising:a pulse light source for generating the ultra-short pulse laser light; andan irradiation unit for discretely irradiating the ultra-short pulse laser light generated by the pulse light source on terahertz wave transmission line in the non-linear optical crystal so that the ultra-short pulse laser light is in synchronous with transmission of the terahertz wave; whereinthe irradiation unit comprises a plurality of optical fibers for receiving and transmitting the ultra-short pulse laser light generated by the pulse light source and projecting such toward the terahertz wave transmission line of the non-linear optical crystal so that the ultra-short pulse laser light is in synchronous with transmission of the terahertz wave;the optical fibers have mutually differing optical path lengths;the irradiation unit further comprises a light distributor for splitting the ultra-short pulse laser light generated by the pulse light source into a plurality of ultra-short pulse laser lights and transmitting such to the plurality of optical fibers;the irradiation unit further comprises a length adjustment mechanism for adjusting the optical path lengths of the optical fibers; andthe length adjustment mechanism comprises drums around which the optical fibers are wound, and a tension changing unit for changing the tension of the optical fibers in a lengthwise direction by changing the diameter of the drums. 2. The terahertz wave generation device according to claim 1, wherein the pulse light source is provided for each optical fiber. 3. The terahertz wave generation device according to claim 2, wherein each pulse light source comprises a timing adjustment mechanism for adjusting the generation timing of the ultra-short pulse laser light. 4. The terahertz wave generation device according to claim 3, wherein at the projection unit of each of the transmission paths, a lens is provided for making desired values of the incident angle on the non-linear optical crystal and the spacing of arrival positions on the terahertz wave transmission line in the non-linear optical crystal of the ultra-short pulse laser light projected from the projection units. 5. A terahertz wave generation device for generating terahertz waves in a direction satisfying non-collinear phase matching conditions by projecting ultra-short pulse laser light on a non-linear optical crystal, the terahertz wave generation device comprising:a pulse light source for generating the ultra-short pulse laser light; andan irradiation unit for discretely irradiating the ultra-short pulse laser light generated by the pulse light source on terahertz wave transmission line in the non-linear optical crystal so that the ultra-short pulse laser light is in synchronous with transmission of the terahertz wave; whereinthe irradiation unit comprises a multi-core fiber having a plurality of cores connected to respective optical fibers as transmission paths for receiving and transmitting the ultra-short pulse laser light generated by the pulse light source and projecting such toward the terahertz wave transmission line of the non-linear optical crystal so that the ultra-short pulse laser light is in synchronous with transmission of the terahertz wave;the plurality of cores have mutually differing optical path lengths;the irradiation unit further comprises a length adjustment mechanism for adjusting the optical path lengths of the optical fibers; andthe length adjustment mechanism comprises drums around which the optical fibers are wound, and a tension changing unit for changing the tension of the optical fibers in a lengthwise direction by changing the diameter of the drums. 6. The terahertz wave generation device according to claim 5, wherein projection units of the plurality of cores are positioned parallel to the direction of the terahertz wave transmission line; andthe end surface composed of the projection units of the plurality of cores is shaped so as to have a predetermined angle, and the optical path lengths of the cores are longer toward one side in a direction parallel to the projection units. 7. The terahertz wave generation device according to claim 6, wherein the pulse light source is provided for each of the cores. 8. The terahertz wave generation device according to claim 7, wherein each pulse light source comprises a timing adjustment mechanism for adjusting the generation timing of the ultra-short pulse laser light. 9. The terahertz wave generation device according to claim 8, wherein at the projection unit of each of the transmission paths, a lens is provided for making desired values of the incident angle on the non-linear optical crystal and the spacing of arrival positions on the terahertz wave transmission line in the non-linear optical crystal of the ultra-short pulse laser light projected from the projection units. 10. A method for generating terahertz waves in a direction satisfying non-collinear phase matching conditions by making ultra-short pulse laser light incident on a non-linear optical crystal, the method comprising:a step of irradiating a pulse laser light group having a discrete wave surface composed of a plurality of ultra-short pulse laser lights of a single repeating frequency toward the non-linear optical crystal; anda step of transmitting the ultra-short pulse laser lights of the pulse laser light group to successively shifted positions on the terahertz wave transmission line so that such arrive with a time difference; whereinthe pulse laser light group is composed by the ultra-short pulse laser lights being transmitted via transmission paths comprising optical fibers and being projected from projection units of the transmission paths toward the non-linear optical crystal;the shift in arrival positions of the ultra-short pulse laser lights on the transmission line is created by the projection units being parallel in one direction; andthe difference in arrival times of the ultra-short pulse laser lights on the transmission line is created by the optical path lengths of the transmission paths being longer toward one side of the parallel direction of the projection units; andthe terahertz wave generating method further comprising a step of adjusting the optical path lengths of the optical fibers with a length adjustment mechanism; whereinthe length adjustment mechanism comprises drums around which the optical fibers are wound, and a tension changing unit for changing the tension of the optical fibers in a lengthwise direction by changing the diameter of the drums. |
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052672808 | claims | 1. Process for the treatment of used ion cartridges stored in a cartridge storage pool and constituted by a hollow internal metal structure, having at its lower end at least one filter screen and filled with ion exchange resins, use being made of a plurality of pools and at least one transfer casket equipped with a cover and making it possible to transport a cartridge from a pool to a cell, characterized in that it comprises performing the following operations: a) transfer of the cartridge from the storage pool to a discharge pool, b) placing the cartridge to be conditioned within the transfer casket, c) decontamination of the transfer casket in a decontamination cell, d) transfer of the casket to a treatment cell and e) extraction by suction means of the ion exchange resins contained in the cartridge. f) removal of the empty cartridge from the casket and decontamination of the internal metal structure of the cartridge, g) deposition of the decontaminated empty cartridge in a conditioning case ensuring the confinement of said cartridge and h) deposition of the case in a storage container. j) removal of the empty cartridge from the casket and transfer of said cartridge into the treatment cell, k) disassembly of the filter screen and washing the latter, l) decontamination and cleaning of the internal metal structure of the cartridge without its screen, m) refitting the filter screen to the empty cartridge and putting back said cartridge in place in the transfer casket and n) filling the internal metal structure of the cartridge with new ion exchange resins. 2. Process according to claim 1, characterized in that the operation a) consists of loading the cartridge to be conditioned into a transfer basket in an immersed medium for the transfer thereof from the storage pool to the discharge pool. 3. Process according to claim 1, characterized in that the transfer of the casket to the treatment cell comprises an operation of engaging the casket beneath said treatment cell. 4. Process according to claim 3, characterized in that the engagement operation consists of tightly positioning the casket beneath the treatment cell, removing the cover from the casket and transferring said cover into said cell in order to ensure an opening between the casket and the cell. 5. Process according to claim 1, characterized in that the suction means used for the extraction of the ion exchange resins have a suction pipe introduced on the one hand into the cartridge and connected on the other to a pneumatic pump, which is itself connected to a resin storage unit. 6. Process according to claim 1, characterized in that it also comprises carrying out, following the extraction of the ion exchange resins, the following operations for the conditioning of the empty cartridges: 7. Process according to claim 6, characterized in that the deposition of the cartridge in the case comprises an operation of engaging the case beneath the treatment cell. 8. Process according to claim 7, characterized in that the engagement operation consists of tightly positioning the case beneath the treatment cell, removing the cover from the case and transferring said cover into said cell in order to ensure an opening between said case and said cell. 9. Process according to claim 6, characterized in that it involves a final operation i) consisting of filling the container with concrete for the storage thereof. 10. Process according to claim 6, characterized in that it comprises, following the operation f), an operation of cutting up the metal structure in order to reduce its volume during the conditioning of said structure in the storage container. 11. Process according to claim 1, characterized in that it also comprises carrying out, following the extraction of the ion exchange resins, of the following operations for recycling the empty cartridges: 12. Process according to claim 11, characterized in that the washing of the filter screen consists of transferring said screen into an ultrasonic tank, where it is cleaned. 13. Process according to claim 11, characterized in that it involves a final operation o) of transferring the casket into a discharge pool, where the cartridge is extracted from said casket and transferred into the cartridge storage pool by a special vehicle. |
048184799 | claims | 1. A ductless core component for use in a liquid cooled, fast flux nuclear reactor core for axially and laterally constraining a plurality of elongated rods containing nuclear material comprising: an upper support plate member, an upper spacer grid member, a plurality of intermediate spacer grid members, a lower spacer grid member, a lower support plate member, a nozzle means, a flow dispersion tube means for fluidly and mechanically connecting said nozzle means to said lower support plate member, a handling socket means attached to said upper support plate member for enabling gripping of the core component, a plurality of orifice plates disposed between said lower support plate member and said nozzle means, a plurality of support rods passing through said lower spacer grid member, said intermediate spacer grid members, and said upper spacer grid member fixedly attached to said lower support plate member and to said upper support plate member, said support rods being provided with passage means axially therethrough for passing a flow of coolant, and a plurality of spacer sleeve means disposed about each of said support rods between said upper spacer grid member and the uppermost of said intermediate spacer grid members, between each of said intermediate spacer grid members and said intermediate spacer grid member immediately beneath it, and between the lowermost of said intermediate spacer grid members and said lower space grid member for maintaining a predetermined spacing between said grid members between which said spacer sleeve means are disposed, each of said spacer sleeves being bowed with respect to their axis to grip said support rods to prevent flow induced vibration damage to said support rods, each of said upper spacer grid member, said plurality of intermediate spacer grid members, and said lower spacer grid member comprising a plurality a cell members, the cell members being substantially triangular and being formed through the interconnection of a plurality of top grid strip members, a plurality of middle grid strip members, a plurality of bottom grid strip members, and outer grid strip members, each elongated rod passing through and being laterally constrained within one of the cell members, each cell member comprising three wall members, each wall member comprising a portion of a top grid strip member, a middle grid strip member, a bottom grid strip member, or an outer grid strip member. 2. The ductless core component according to claim 1 wherein each wall member of each cell member formed by interconnecting the top grid strip members, the middle grid strip members, the bottom grid strip members, and the outer grid strip members consists of a portion of a top grid strip member, a middle grid strip member, a bottom grid strip member, or an outer grid strip member. 3. The ductless core component according to claim 1, futher comprising hardstop members provided intermediate two of the wall members of each substantially triangular cell members and spring means provided intermediate a third wall member of each cell member for biasing an elongated rod passing through the cell member against the hardstop members. 4. The ductless core component according to claim 3, wherein said spring means is disposed at an angle with respect to the flow of coolant through the ductless core component, which flow is substantially along the axis of the elongated rods, which axis corresponds to the axis of the ductless core component, and is resilient, whereby the flow of coolant upwardly through the cell member causes a biasing force in said spring means, the biasing force forcing the rod against said hardstop members, the strength of the biasing force being dependent on the rate of flow of coolant through the nuclear reactor core component. 5. The ductless core component according to claim 1 wherein the top grid strip members are provided with first slot means along the bottom edges thereof for receiving middle grid strip members, the middle grid strip members are provided with the second slot means along the top edges thereof for engaging said second slot means of the top grid strip members when the top grid strip members and the middle grid strip members are interconnected, the middle grid strip members being further provided with the third slot means along the bottom edges thereof for receiving bottom grid strip members, and the bottom grid strip members are provided with third slot means along the top edges thereof for engaging the third slot means when the middle grid strip members and the bottom grid strip members are interconnected. 6. The ductless core component according to claim 5 wherein the outer grid strip member of each of said top, intermediate and lower spacer grid members is adapted to engage the lateral edges of the top grid strip members, the middle grid strip members and the bottom grid strip members interconnected to form each top, intermediate, and bottom grid member. |
046845011 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like are words of convenience and are not to be construed as limiting terms. In General Referring now to the drawings, and particularly in FIGS. 1 and 3, there is shown a nuclear fuel assembly, generally designated 10 for a boiling water nuclear power reactor (BWR), in which the improvement of the present invention is incorporated. The fuel assembly 10 includes an elongated outer tubular flow channel 12 that extends along substantially the entire length of the fuel assembly 10 and interconnects an upper support fixture or top nozzle 14 with a lower base or bottom nozzle 16. The bottom nozzle 16 which serves as an inlet for coolant flow into the outer channel 12 of the fuel assembly 10 includes a plurality of legs 18 for guiding the bottom nozzle 16 and the fuel assembly 10 into a reactor core support plate (not shown) or into fuel storage racks, for example in a spent fuel pool. The outer flow channel 12 generally of rectangular cross-section is made up of four interconnected vertical walls 20 each being displaced about ninety degrees one from the next. Formed in a spaced apart relationship in, and extending in a vertical row at a central location along, the inner surface of each wall 20 of the outer flow channel 12, is a plurality of structural ribs 22. The outer flow channel 12, and thus the ribs 22 formed therein, are preferably formed from a metal material, such as an alloy of zirconium, commonly referred to as Zircaloy. Above the upper ends of the structural ribs 22, a plurality of upwardly-extending attachment studs 24 fixed on the walls 20 of the outer flow channel 12 are used to interconnect the top nozzle 14 to the channel 12. For improving neutron moderation and economy, a hollow water cross 26 extends axially through the outer channel 12 so as to provide an open cruciform inner channel 28 for subcooled moderator flow through the fuel assembly 10 and to divide the fuel assembly into four, separate, elongated compartments 30. The hollow water cross 26 is mounted to the angularly-displaced walls 20 of the outer channel 12. Preferably, the outer, elongated lateral ends of the water cross 26 are connected such as by welding to the structural ribs 22 along the lengths thereof in order to securely retain the water cross 26 in its desired central position within the fuel assembly 10. Also, the water cross 26 has a lower flow inlet end 32 and an opposite upper flow outlet end 34 which each communicate with the inner channel 28 for providing subcoolant flow therethrough. Disposed within the channel 12 is a bundle of fuel rods 36 which, in the illustrated embodiment, number sixty-four and form an 8.times.8 array. The fuel rod bundle is, in turn, separated into four mini-bundles thereof by the water cross 26. The fuel rods 36 of each mini-bundle, such being sixteen in number in a 4.times.4 array, extend in laterally spaced apart relationship between an upper tie plate 38 and a lower tie plate 40. The fuel rods in each mini-bundle are connected to the upper and lower tie plates 38,40 and together therewith comprise a separate fuel rod subassembly 42 within each of the compartments 30 of the channel 12. A plurality of grids or spacers 44 axially spaced along the fuel rods 36 of each fuel rod subassembly 42 maintain the fuel rods in their laterally spaced relationships. The lower and upper tie plates 38,40 of the respective fuel rod subassemblies 42 have flow openings 46 defined therethrough for allowing the flow of the coolant fluid into and from the separate fuel rod subassemblies. Also, coolant flow paths provide flow communication between the fuel rod subassemblies 42 in the respective separate compartments 30 of the fuel assembly 10 through a plurality of openings 48 formed between each of the structural ribs 22 along the lengths thereof. Coolant flow through the openings 48 serves to equalize the hydraulic pressure betweent he four separate compartments 30, thereby minimizing the possibility of thermal hydrodynamic instability between the separate fuel rod subassemblies 42. Compliant Inserts In Upper Tie Plate Holes Turning now to FIGS. 4 to 9, there is shown several embodiments of a combination of features of the present invention for supporting the upper ends of the fuel rods 36 so as to avoid binding and axially loading thereof which heretofore has frequently resulted in bowing of the fuel rods. As depicted in FIG. 4, each of the fuel rods 36 is one of two types: the tie rod 36a or the standard fuel rod 36b. Each has a pair of end plugs 50a,50b (only the upper one being shown) sealing opposite ends thereof. The upper end plugs 50a,50b of the fuel rods 36a,36b have respective extension members 52a,52b thereon which extends axially outward from the end plugs and have respective diameters less than that of each fuel rod. Also, as seen in FIG. 4, the upper tie plate 38 disposed adjacent the upper ends of the fuel rods 36a,36b has a plurality of holes 54 defined by endless sidewalls 56 formed therethrough between opposite upper and lower sides 58 of the upper tie plate. The holes 54 are arranged in an array which matches that of the fuel rods 36. The extension member 52a of the tie rod 36a is threaded and fastened by a nut 60 so as to limit its movement within one of the holes 54 in the upper tie plate 38, whereas the extension member 52b of the standard fuel rod 36b is slidably received within another one of the holes 54 in the tie plate. In the case of each fuel rod 36, there is a compressed coil spring 61 disposed above the extension member 52 and extending between the respective end plug 50 and the upper tie plate 38. The springs 61 force the tie plate 38 upwardly against the nut 60 on the tie rod 36a. The holes 54 are substantially larger in diameter than the respective end plug extension members 52a,52b of the fuel rods 36a,36b for accommodating insertion of a compliant insert, generally designated 62, in each of the holes 54 of the upper tie plate 38. There are three different embodiments of the insert 62 disclosed, each being modified slightly from the other. However, basically, all embodiments of the compliant insert 62 function to engage both the tie plate 38 and the respective end plug extension member 52 so as to yieldably support the extension member within the given hole 54 in spaced relationship from the hole sidewall 56. Also, each embodiment of the compliant insert 62 includes a plurality of spring members 64. Turning initially to the first embodiment depicted in FIGS. 4 and 5, it will be seen that the spring members 64a making up the compliant insert 62a are separate from one another, being angularly spaced apart approximately 120 degrees about the hole 54. Each of the spring members 64a is made from a strip of resiliently flexible material which is also creep resistant, such as Inconel. Each spring member 64a has opposite upper and lower end portions 66a and a middle portion 68a interconnecting the opposite end portions. The opposite end portions 66a in the form of tabs are disposed along the opposite sides 58 of the tie plate 38 adjacent to the respective tie plate hole 54. In addition, means are provided for securing the opposite end portions 66a of each spring member 64a to the respective sides 58 of the tie plate 38. In the first embodiment of FIG. 4, such means take the form of tackwelds 70 which interconnect the spring member end portions 66a to the tie plate 38. The elongated middle portion 68a of each spring member 64a extends through the tie plate hole 54 between the hole sidewall 56 and the respective one of extension members 52a,52b. Resilient means is defined on each spring member middle portion 68a for engaging and positioning the respective extension member 52a,52b in spaced relationship from the hole sidewall 56. In the first embodiment of FIG. 4, such resilient means is in the form of a single inwardly-protruding dimple 72 formed on each spring member 64a. In view of the above-described arrangement, the spring members 64a of the insert 62a will yield and prevent binding of the respective end plug extension member 52a,52b upon tilting of the tie plate 38 relative to the plug extension member. They will also support the extension member so as to eliminate lateral vibrations thereof, while at the same time allow the end plug extension member to freely slide relative to the upper tie plate. The alternative embodiment of the compliant insert 62b seen in FIGS. 6 and 7 is generally similar to the first embodiment of FIGS. 4 and 5. Therefore, only the difference between the two will be described. The opposite end portions or tabs 66b of the separate spring members 64b are secured to the respective sides 58 of the upper tie plate 38 by means of cutouts or indentations 74 formed therein into which turned ends 76 on the tabs 66b extend. Also, angularly-spaced recesses 78 are formed in the hole sidewall 56 in which the middle portions 68b of the spring members 64b are seated to maintain the circumferential positioning of the spring members in the hole 54. Further, the resilient means of the insert 62b is a combination of the single inwardly-protruding dimple 72 formed on one spring member middle portion 68b and a pair of tandemly-arranged inwardly-protruding dimples 80 formed on one of the other spring member middle portions 68b. The pair of bi-level dimples 80 provide more resistance to fuel rod tilting if that should be desired in a particular case. In another alternative embodiment of the compliant insert 62c seen in FIGS. 8 and 9, the spring members 64c are integrally connected to one another by spaced apart upper and lower ring portions 82. Here, the opposite end portions or tabs 66c are connected to the respective ring portions and circumferentially displaced from the middle portions 68c of the spring members 64c. The tabs 66c are also bendable between axially-extending releasing positions, seen in dotted outline form, and radially-extending securing positions, seen in full line form, in FIG. 9. The integral structure of the spring members 64c allows the insert 62c to anchor itself in the tie plate hole 54. It is thought that the invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
description | 1. Field of the Invention The present invention relates to techniques for reducing the migration of radioactive materials from a nuclear reactor to a steam and turbine system. 2. Description of the Related Art In a nuclear power plant, the reduction of exposure during operation and regular inspection is important. Various materials have been proposed, water quality control measures have been taken and the improvement of purifying facilities have been made principally for the reduction of the cobalt-60 concentration of reactor water and the migration of radioactive materials to the water systems of nuclear reactors. However, any measures to reduce the migration of radioactive materials to the main steam line and turbine system have not been taken. Modes of making radioactive materials migrate to a steam system of a present nuclear power plant system and the ratio between the radioactive material carrying effects of the modes will be explained with reference to FIGS. 1 and 2 and problems to be solved will be explained. Referring to FIG. 1, part of radioactive materials produced by a reactor core 10 contained in a reactor pressure vessel (hereinafter referred to simply as “pressure vessel”) 9 is removed by a reactor water purifying system 5. Most part of the radioactive materials adheres to devices placed in the pressure vessel 9 through which saturated water is circulated, the inner surface of the pressure vessel 9 and pipes extending on the upper side of the reactor water purifying system 5. Very small part of the radioactive materials dissolved in the reactor water in ions or molecules has a partial vapor pressure and volatilizes together with steam. On the other hand, a steam separator 4 and a dryer 3 remove most part of liquid drops. Very small part, i.e., 0.1% or below, of liquid drops migrates in small particles through a valve 1 placed at the inlet of the turbine system into a steam turbine and contaminates the steam turbine system and the associated parts. Recently, the temperature of the nuclear reactor is lowered rapidly when the nuclear reactor is shut down. When thus shutting down the nuclear reactor, water having a high radioactive concentration and discharged from the reactor water purifying system 5 is sprayed by a head spray nozzle 6 and, consequently, radioactive materials contained in the sprayed liquid drops migrate to the steam system. The behavior of liquid drops in a dryer (steam dryer) 3 closely related to the migration of liquid drops to the steam system will be described with reference to FIG. 2. Steam containing liquid drops in a wetness of 10% or below and passed through the steam separator 4 shown in FIG. 1 is dispersed by steam dispersing openings 20, flows through spaces between corrugated plates 22 and an upper part of the pressure vessel 9 into a line 11. Whereas the steam flows along the surfaces of the corrugated plates 22, liquid drops having large mass collide against the surfaces of the corrugated plates 22 and are caught by the corrugated plates 22. The liquid drops thus caught by the corrugated plates 22 flows down along the corrugated plates 22 into a drain pan 24, and returned into the reactor water system through a drain pipe 25. If the steam containing liquid drops flows at a high velocity through the spaces between the corrugated plates 22; that is, if time for which the steam stays in the spaces between the corrugated plates 22 is short, minute liquid drops flow through the spaces between the corrugated plates 22 without colliding against the corrugated plates 22. Furthermore, the steam flowing at a high velocity through the spaces between the corrugated plates 22 separates the liquid drops and radioactive materials, which have been collided against and adhering to the corrugated plates 22, off the corrugated plates 22 and carries the same away to the steam system. Thus, the ratio of migration of radioactive materials to the steam system in an actual nuclear power plant is greater than that calculated on the basis of the gas-liquid distribution ratio of the radioactive materials dissolved in water. Radioactive materials migrate to the steam system in the following three modes; (i) a first migration mode in which radioactive materials dissolved in the reactor water evaporate and migrate into the steam system, (ii) a second migration mode in which liquid drops collided against the components of a device, such as a dryer, and caught by the components of the device are separated from the components of the device in liquid drops or radioactive materials dried and adhering to the components of the device are separated from the components of the device, and the separated liquid drops or the separated radioactive materials are carried into the steam system and (iii) a third migration mode in which liquid drops sprayed by a head spray migrate to the steam system. The ratio between the respective radioactive material carrying effects of those modes is 1:3:1. Recently, the enhancement of power, temperature capacity and pressure capacity without changing the sizes of devices has been desired from the economical point of view. However, in view of the forgoing problems, it can be readily conjectured that the ability to separate liquid drops from steam of the steam separator and the dryer will become insufficient and the migration of radioactive materials to the steam system will increase when the power capacity of the nuclear reactor is increased. If the temperature capacity and pressure capacity are further enhanced, a supercritical state will result. In the supercritical state, all the particles of radioactive materials contained in supercritical water or all the radioactive materials dissolved in supercritical water migrate to the steam system unless a radioactive material separating and removing apparatus is used. Therefore a high-temperature water purifying apparatus is one of the most important apparatuses of the supercritical reactor. In this specification, the term “high-temperature water purifying apparatus” is used to denote an apparatus capable of being used under a high-temperature condition for separating and removing radioactive materials from water or steam. Various high-temperature water purifying apparatuses that do not entail heat loss have been proposed. However, none of the previously proposed high-temperature water purifying apparatuses is able to avoid performance degradation due to rise in differential pressure caused by contamination by collected corrosion products and organic substances used by the turbine system, and changes in the shape of small holes caused by the volume expansion due to the dissolution, corrosion or oxidation of a filter aid by the chemical instability of high-temperature water and those high-temperature water purifying apparatuses have a short life. Furthermore, those high-temperature water purifying apparatuses have a low trapping capacity and their purifying ability deteriorates in a short time. Consequently, those high-temperature water purifying apparatuses have not been applied to practical uses yet. (Refer to “Filter Guidebook for Pall-Generator”, Nihon Pall Ltd., P.8.) The present invention has been made in view of the foregoing circumstances and it is therefore an object of the present invention to provide a means for reducing the migration of radioactive materials from a nuclear reactor to a steam turbine system. To achieve the objective, the present invention provides a nuclear power plant system including: a nuclear reactor; a steam turbine that uses steam generated in a pressure vessel included in the nuclear reactor; and a radioactive material separating and removing apparatus placed in the pressure vessel or in a steam passage extended between the pressure vessel and an inlet of the steam turbine to separate and remove radioactive materials from steam. Preferably, the radioactive material separating and removing apparatus has a high-temperature water purifying apparatus employing a metal or a metal oxide as an ion-exchange material that exchanges ions for radioactive ions, which is stable in an environment where high-temperature water or steam exists. The present invention also provides a nuclear power plant system including: a nuclear reactor; a steam turbine that uses steam generated in a pressure vessel included in the nuclear reactor; and a radioactive material separating and removing apparatus placed in a reactor water system attached to the nuclear reactor, the pressure vessel or a steam passage extended between the pressure vessel and an inlet of the steam turbine to separate and remove radioactive materials from steam, wherein the separating and removing apparatus employs a metal or a metal oxide as an ion-exchange material that exchanges ions for radioactive ions, which is stable in an environment where high-temperature water or steam exists. The present invention further provides a method of operating a nuclear power plant system having a nuclear reactor including a pressure vessel provided with a head spray, in order to lower temperature of the pressure vessel. The method including the steps of: limiting range of scattering of water sprayed by the head spray; decreasing size of water drops of the water sprayed by the head spray; and controlling an amount of the water to be sprayed according to the amount of steam generated by heat generated after shutdown of the nuclear reactor. The present invention further provides a method of operating a nuclear power plant system having a nuclear reactor including a pressure vessel provided with a head spray. The method including the steps of: supplying water not containing any radioactive materials or water having a small radioactive material concentration and supplied from a condensate purifying apparatus or a condensate storage tank. Preferred embodiments of the present invention will be described hereinafter with reference to the accompanying drawings. There are spatial and economical difficulties in additionally installing a new radioactive material separating and removing apparatus in an existing nuclear power plant. The reduction of the amount of radioactive materials that migrate to a steam system can be achieved through the enhancement of the abilities of a steam separator 4 and a dryer 3 contained in a pressure vessel 9 by incorporating improvements therein or by replacing the steam separator 4 and the dryer 3 with those having improved abilities. Description will be given on measures to prevent the migration of liquid drops having a high radioactive material carrying effect once collided against the component members of an apparatus, such as the dryer 3, and separated from those component members to the steam system or the migration of dry radioactive materials deposited on the component members of the dryer 3 and separated from the same to the steam system. FIG. 2 shows a dryer 3 provided with improved corrugated plates 22 having surfaces coated with a coating material, such as TiO2, ZrO2 or ferrite. The surfaces of other component members other than the corrugated plates 22 of the dryer 3 also may be coated with such a coating material. Ferrite, TiO2 and ZrO2 are chemically stable in high-temperature water or steam and capable of maintaining the following functions for a long period of time. TiO2 has an ion exchanging ability and/or a superhigh hydrophilic property. TiO2 is used as an ion-exchange material at high temperatures. It is generally known that TiO2 shows a superhigh hydrophilic property when it is used in combination with an SiO2 binder. When the surfaces of the corrugated plates 22 are coated with a substance having a superhigh hydrophilic property, liquid drops fallen on the corrugated plates 22 spread over the surfaces of the corrugated plates 22 in thin liquid films. Such thin liquid films are difficult to separate from the surfaces of the corrugated plates 22 by shearing force exerted thereon by steam stream. Corrosion products contained in the liquid drops adhere firmly in flat films to the surfaces of the corrugated plates 22 after the liquid drops fallen on the surfaces of the corrugated plates 22 have dried up. Since TiO2 has an ion exchanging ability, the surfaces coated with TiO2 of the corrugated plates 22 are able to catch particles of radioactive materials. Thus, radioactive materials deposit on the corrugated plates 22 in ionized corrosion products and adhere firmly to the surfaces of the corrugated plates 22. Radioactive materials thus deposited on the corrugated plates 22 are difficult to separate from the corrugated plates 22. Principal radioactive materials, such as 60Co, 58Co and 54Mn that migrate to the steam system exist in ions in the reactor water. Therefore coating the surfaces of the corrugated plates 22 with TiO2 is effective. Ferrite and ZrO2 have an ion exchanging ability as well as TiO2. TiO2 and ZrO2 exercise a photocatalytic function to decompose organic substances when exposed to radioactive rays, such as intense gamma rays emitted by 16N, and Cerenkov radiation produced by radioactive rays. Since the turbine system uses organic materials, such as oils, the hydrophilic property and the ion exchanging property of the corrugated plates 22 can be maintained without requiring cleaning work when the surfaces of the corrugated plates 22 are coated with such a material having an ability to decompose organic substances, so that necessary maintenance work can be greatly reduced. TiO2 is an additive added to steels and is an oxide that can be easily produced on the surface of a material by a corrosive reaction. TiO2 can be produced on the surface of a material by high-temperature oxidation in an atmosphere of reduced pressure on the order of 10−4 MPa where the concentration of the air is very small. Therefore, the corrugated plates 22 having surfaces coated with a TiO2 coating can be formed by forming the corrugated plates 22 from a plate having a properly adjusted chemical composition, such as a plate of a steel containing Ti, and subjecting the corrugated plates 22 to high-temperature oxidation. When it is desired to form a TiO2 coating containing SiO2 as a binder, the corrugated plates 22 are formed, for example, from a steel plate of a steel containing Ti and Si, and the corrugated plates 22 are subjected to a high-temperature oxidation process. The surfaces of the corrugated plates 22 may be coated with very fine TiO2 and SiO2 fibers. When the surfaces of the corrugated plates 22 are coated with such very fine fibers, water is made to soak the very fine fibers by capillarity. Consequently, the water fallen on the corrugated plates 22 undergoes scarcely the shearing force of steam and is difficult to separate from the corrugated plates 22. After the water fall on the corrugated plates 22 has been dried, corrosion products contained in the water are held between the fine fibers. A TiO2 coating can be formed on the surfaces of the corrugated plates 22 by forming a layer of a material from which TiO2 can be easily produced, such as Ti or a Ti alloy, and subjecting the layer to high-temperature oxidation. The Ti or Ti alloy layer can be formed by a known physical method, such as thermal spraying, or a chemical method. When it is desired to coat the surfaces of the corrugated plates 22 with a coating of TiO2 and an SiO2 binder, the surfaces of the corrugated plates 22 are coated with, for example, a metal from which TiO2 and SiO2 can be easily produced by a physical or chemical method and the corrugated plates 22 are subjected to high-temperature oxidation. A ZrO2 coating can be formed by the same method. A ZrO2 coating can be formed by forming a layer of a material from which ZrO2 can be easily produced, such as Zr or a Zr alloy, by a physical or chemical method and subjecting the layer to high-temperature oxidation. Ferrite can be produced by subjecting a Fe-base alloy, such as a stainless steel, or a nickel-base alloy, such as Inconel, to high-temperature oxidation. Coatings of TiO2, ZrO2 and ferrite can be formed on the surfaces of corrugated plates of a dryer, which has been used in an existing nuclear power plant for the aforesaid effects. When forming such a coating on the corrugated plates of a used dryer, the surfaces of the corrugated plates are cleaned by a jet cleaning method or the like to remove n-type semiconductor oxides deposited on and comparatively loosely adhering to the surfaces of the corrugated plates and to expose a p-type oxide film firmly adhering to the surfaces of the corrugated plates. Then a TiO2 ZrO2 or ferrite coating is formed on the surfaces of the corrugated plates by a spraying method using a remotely controllable spray nozzle or a thermal spraying method. The improvement of the trapping capacity according to the above is achieved by preventing the separation of the trapped liquid drops or radioactive materials. Next, a method that improves the trapping efficiency by positively catching liquid drops or radioactive materials will be described hereinafter. Very fine water drops and charged particles, such as ions and molecules, can be easily charged because they are minute and radioactive. When a charged particle is in an Electric field E or a magnetic field H as shown in FIG. 3, a force F acts on the charged particle in a direction perpendicular to both the electric field E and the magnetic field H. Charged particles can be moved toward the corrugated plates 22 by using such effect of an electric field or a magnetic field. Trapping of radioactive materials using an electric field is applied to a radioactive ray monitor. A trapping technique using an electric field for trapping minute particles is used prevalently in the chemical engineering field. A particle moving technique using a magnetic field for moving particles is applied to charged particle accelerators. Charged particles can be trapped by the dryer 3 shown in FIG. 2 by forming fixed bars 23 holding the corrugated plates 22 of an insulating material, electrically isolating the corrugated plates 22 from the casing of the dryer, and applying a voltage across the adjacent corrugated plates 22. A power source is necessary to realize such a function. Although power may be supplied by an external power source, power is available by the following method without using any external power source. The surfaces of the corrugated plates of a used dryer are cleaned by a jet cleaning method or the like to remove n-type semiconductor oxides deposited on and comparatively loosely adhering to the surfaces of the corrugated plates and to expose a p-type oxide film firmly adhering to the surfaces of the corrugated plates. Then a TiO2 or ZrO2 coating is formed on the surfaces of the corrugated plates. Since TiO2 and ZrO2 are n-type semiconductors, the TiO2 or ZrO2 coating is excited by radioactive rays or Cerenkov radiation generated by radioactive rays. Consequently, electrons break their bonds and create holes, so that electricity is generated. The corrugated plates are able to use this electricity for scavenging charged particles; that is, power generated by photocells formed of the superposed layers of the p-type oxide film and the n-type semiconductor oxide film can be used as power sources. In the foregoing description, the p-type oxide film is supposed to be formed by high-temperature oxidation during the operation of the dryer. A p-type oxide film may be artificially formed when fabricating a new dryer. The trapping efficiency of the dryer can be improved by changing the geometrical shape of the corrugated plates 22 instead of using the physical or chemical method. When the dryer 3 is designed by the present design rule, the probability of collision of a single water molecule against the corrugated plates 22 of the dryer 3 is as small as about 5%. Therefore it is theoretically possible to make substantially 100% of water drops collide against the corrugated plates 22 if the contact area of the corrugated plates 22 are increased by twenty, provided that the condition of flows in the dryer 3 is not changed. The contact surface area of the corrugated plates 22 can be increased, when the thickness of each corrugated plate 22 is appropriately decreased on condition that the necessary mechanical strength of the corrugated plat 22 is maintained. The above method of increasing the probability of collision is effective in improving an ability of separating and removing minute water drops, ions and molecules contained in a multiphase flow and moving at a low migration speed toward the surfaces of the corrugated plates 22. Although differential pressure rises unavoidably when the contact surface area of the corrugated plates 22 is increased, increase in the contact surface area increases the probability of collision of the radioactive materials contained in water or steam against the corrugated plates 22 and the trapping efficiency of the corrugated plates 22. The above embodiment has been described in connection with the improvement of the corrugated plates 22 of the dryer 3. However, the techniques relating to the above embodiment are applicable to the walls themselves of the pressure vessel of the nuclear reactor, and are also applicable to other apparatuses arranged inside and outside the pressure vessel through which water, steam or a multiphase fluid containing water and steam flows. The surfaces exposed to water or steam of the component members of those apparatuses are coated with TiO2, ZrO2 or ferrite for the substantially the same effects as the aforesaid ones. A second embodiment of the present invention will be described. The second embodiment relates to the improvement of a high-temperature water purifying apparatus provided with a filter that can be used at a high temperature, i.e., a high-temperature filter. FIG. 5A shows a high-temperature water purifying apparatus 40 included in a nuclear power plant system in a second embodiment according to the present invention in a typical sectional view. The high-temperature water purifying apparatus 40 is suitable for purifying high-temperature water in a liquid phase, which does not mean that the high-temperature water purifying apparatus 40 is not applicable to purifying steam. As shown in FIG. 5A, the high-temperature water purifying apparatus 40 is provided with a plurality of hollow membrane pipes 30, i.e., high-temperature filters. As shown in FIGS. 4A and 4B, the hollow membrane pipe 30 is a double-wall structure having a skin 31 provided with minute pores capable of easily producing differential pressure and of catching minute particles, and a substrate 32 formed on the inner side of the skin 31 to hold the latter. The substrate 32 is provided with many minute pores greater than those of the skin 31. The substrate 32 has a tubular shape defining a hollow bore 33. Desirably, the size of the minute pores of the skin 31 is 0.45 μm or below. Sizes of most of particles of radioactive materials and corrosion products contained in reactor water are greater than 0.45 μm. Therefore the minute pores of the skin 31 are not clogged with those particles and surface filtering is possible when the sizes of the minute pores of the skin 31 are 0.45 μm or below. The hollow membrane pipes 30 are not limited to those of the double-wall structure shown in FIG. 4B, but may be those of a multiple-wall structure. The skin 31 and the substrate 32 may be formed of porous materials stable in high-temperature water, such as metals, alloys, composite materials or ceramic materials. More concretely, suitable materials for forming the skin 31 and the substrate 32 are oxides including ferrite oxides, TiO2 and ZrO2, metals capable of producing those oxides, and alloys capable of producing those oxides, including iron-base alloys, such as stainless steels, nickel-base alloys containing iron, titanium alloys and virally. Those materials are highly workable and are suitable for forming filters of appropriate hollow membrane construction having the shape of a complicated, hollow membrane. Preferably, the substrate 32 is a porous, mesh, honeycomb or monolithic structure of particles, plates, ribbons or fibers. The pores of the skin 31 are formed in sizes smaller than the particle sizes of minute particles to be caught by the skin 31 so that the pores of the skin 31 may not be clogged with the minute particles. The skin 31 can be formed by coating the substrate 32 with a thin coating of a fine ceramic material or a fine metal, and firing the thin coating. The sizes of pores of the thin coating are adjusted so that the sizes of the pores are in an appropriate range after the thin coating is corroded. The hollow membrane pipe 30 does not need necessarily to consist of a plurality of exactly divided layers as shown in FIG. 4B, but may be a single wall structure having smaller pores in outer layers and larger pores in inner layers. It is desirable, when the hollow membrane pipe 30 is thus formed, that the sizes of the pores in the outermost layer are 0.45 μm or below. The hollow membrane pipe 30 may be formed by working an original hollow membrane pipe of a metal, alloy or a composite material, into a hollow membrane pipe of cylindrical or pleated shape, and subjecting the hollow membrane pipe to corrosive oxidation in an atmosphere of high-temperature air or in an atmosphere containing steam thereby adjusting the pore's diameter at the outermost area of the membrane pipe. It is preferable to provide the outer circumference and/or the inner circumference of the hollow membrane pipe with a strainer. The strainer prevents the effluence of broken materials if small breakages are formed in the substrate 32 of the hollow membrane pipe 30 and holds a filter aid stably on the outer circumference of the hollow membrane pipe 30. Referring again to FIG. 5A, the high-temperature water purifying apparatus 40 has a vessel 40A. A water discharge port 42 and a backwashing liquid supply port 48 are formed in a part on one side and in a part on the other side, respectively, of an upper part of the vessel 40A. A drain port 47 is formed in the bottom wall of the vessel 40A. A water supply port 41 is formed in a lower part of the side wall of the vessel 40A. A water supply line (water supply pipe) 41A is connected to the water supply port 41 to supply highly contaminated water into the vessel 40A. A water discharge line (water discharge pipe) 42A is connected to the water discharge port 42 to discharge water purified by the high-temperature water purifying apparatus 40. The water supply line 41A and the water discharge line 42A are connected to a bypass line 49A, which is provided with a pre-coating pump 49B (pump for pre-coating), e.g. a mixing pump. Connected to the pre-coating pump 49B is a filter aid supply unit 49C that supplies a filter aid (described later), which is to be supplied into the vessel 40A. An upper support plate 44 is disposed horizontally in the vessel 40A on a level below that of the water discharge port 42 and the backwashing liquid supply port 48. The upper support plate 44 is provided with a plurality of openings and upper ends of the hollow membrane pipes 30 are fitted in the openings of the upper support plate 44. The upper support plate 44 separates perfectly an upper space over the upper support plate 44 and a lower space under the upper support plate 44. Thus water is able to flow between the upper and the lower space only through the hollow membrane pipes 30. A lower support plate 43 is disposed horizontally in the vessel 40A on a level below that of the water supply port 41 of the vessel 40A and above that of the drain port 47. The lower support plate 43 is provided with a plurality of openings. The lower ends of the hollow membrane pipes 30 are set on parts not provided with the openings of the lower support plate 43 so that the lower ends of the hollow membrane pipes 30 are closed. Water is able to flow between an upper space over the lower support plate 43 and a lower space under the same only through the openings of the lower support plate 43. In FIG. 5A, the upper support plate 44 and the lower support plate 43 are fixedly disposed in the vessel 40A and the hollow membrane pipes 30 are held by the upper support plate 44 and the lower support plate 43. The upper support plate 44, the lower support plate 43 and the hollow membrane pipes 30 may be assembled in a cartridge beforehand to replace an old cartridge with a new cartridge. The vessel 40A of the high-temperature water cleaning apparatus 40 may be divided into an upper part having a lower end provided with a flange, and a lower part having an upper end provided with a flange, and the upper and the lower part of the vessel 40A may be joined together by fastening together the flanges, which will facilitate replacing an old cartridge with a new one. When the vessel 40A is thus divided into the upper and the lower part, only the hollow membrane pipes 30 may be replaced with new ones. In operation, a filter aid supply unit 49C supplies a filter aid into the bypass line 49A, and the filter aid is mixed with the water in the bypass line 49A. The water mixed with the filter aid is circulated so that it is supplied into the vessel 40A through the water supply port 41, and is discharged from the water discharge port 42 to return it into the bypass line 49A. While the water mixed with the filter aid is circulated, the filter aid unable to pass through the hollow membrane pipes 30 is gathered on the surfaces of the hollow membrane pipes 30 to form filter aid precoatings 34 on the surfaces of the hollow membrane pipes 30 as shown in FIG. 5B. The filter aid must be at least stable under a condition where high-temperature water is used and capable of ion-exchanging ability. Preferable filter aids are ferrite oxides including hematite (Fe2O3), magnetite (Fe2O4) and nickel ferrite (NiFe2O4), TiO2 and ZrO2. As mentioned in connection with the description of the first embodiment, organic materials decomposing effect of a photocatalytic reaction caused by Cerenkov radiation can be expected of TiO2. Other possible filter aids include pure metals, such as Fe, Ni, Ti and Zr, which are the principal components of the foregoing oxides, alloys of those metals, such as stainless steels, and composite materials respectively containing those metals and alloys. Since the sizes of the pores of the skins of the hollow membrane pipes 30 are 0.45 μm above, it is preferable that the filter aid has a particle size of 1 μm or above to prevent clogging the minute pores of the hollow membrane pipes 30 with the filter aid. In view of the ion-exchanging ability and restriction of the rise in the differential pressure, it is preferable that the filter aid particle has a large specific surface and voidage. Since it is difficult to produce ceramic powder having a large particle size, the filter aid having a particle size on the order of 1 μm may be a porous, fine powder having a large specific surface and a large voidage, because a substance having a particle size greater than the size of the minute pores of the hollow membrane pipes 30 can be held in the vessel 40A. Such a filter aid can be produced by sintering fine particles of 1 μm or below in particle size and economically advantageous. After the filter aid precoatings 34 have been formed on the hollow membrane pipes 30, the bypass line 49A is disconnected from the water supply line 41A and the water discharge line 42A, and then contaminated water to be cleaned is supplied through the water supply port 41 into the vessel 40A. The filter aid precoatings 34 trap radiation-contaminated particles and radioactive ions contained in the contaminated water and filtered water having a reduced radioactivity penetrates the hollow membrane pipes 30. The filtered water penetrated the skins 31 and the substrates 32 of the hollow membrane pipes 30 flows through the bores 33 of the hollow membrane pipes 30 into a water collecting chamber 45, i.e., the upper apace extending over the upper support plate 44. The filtered water is discharged from the water collecting chamber 45 of the vessel 40 through the water discharge port 42 and the discharge line 42A. The thus contaminated filter aid precoatings 34 are cleaned by backwashing. Valves on the lines connected to the water supply port 41 and the water discharge port 42 are closed and backwashing liquid is supplied through the backwashing liquid supply port 48 into the vessel 40. The backwashing liquid flows through the bores 33 of the hollow membrane pipes 30, the substrates 32 and the skins 31 in that order removing the contaminated filter aid precoatings 34 and corrosion products adhering to the hollow membrane pipes 30 from the hollow membrane pipes 30. The thus contaminated, turbid backwashing liquid flows through spaces between the hollow membrane pipes 30 and the openings of the lower support plate 43 into a drain chamber 46, i.e., the lower space under the lower support plate 43. Then, the contaminated, turbid backwashing liquid is discharged through the drain port 47. When an organic acid is added to the backwashing liquid, the hollow membrane pipes 30 can be simultaneously decontaminated and cleaned. Although the foregoing description has been made on an assumption that the outer circumferences of the hollow membrane pipes 30 are coated with the filter aid precoatings 34, the outer circumferences of the hollow membrane pipes 30 do not need necessarily to be coated with the filter aid; the filter aid is sufficiently effective when the precoating liquid is stirred in the vessel 40A of the high-temperature water purifying apparatus 40 such that particles of the filter aid are suspended in the precoating liquid. Particles of the filter aid can be kept suspended in the precoating liquid provided that the velocity of the upward flow of the precoating liquid in the vessel 40A is higher than a sedimentation velocity corresponding to the Stokes radius of particles of the filter aid. A high-temperature filter of the aforesaid type is advantageous over a low-temperature filter in respect of rise in differential pressure due to the collection of particles of corrosion products. However, the high-temperature filter is very disadvantageous in respect of ion exchange because ion-exchanging materials capable of changing ions at a high rate under a high-temperature condition are unavailable. Therefore, the amount of the ion-exchanging material or the surface area must be increased to increase the ion-exchanging ability. Trade-off between the enhancement of trapping minute particles and the reduction of differential pressure must be determined properly. Thus, filtration area must be increased and the rise of differential pressure must be prevented. With an inorganic material that can be used at a high temperature, trapping rate at which ions are trapped by the surface of a material is low as compared with rate of carrying ions from a vapor. A low-temperature removing apparatus using an ion-exchange resin and a high-temperature water purifying apparatus using an inorganic ion exchanging material differ greatly from each other in that respect. The rate of ion trapping reaction is not dependent on the rate of carrying ions from a fluid to the surface of a material but is dependent on the rate of ion trapping reaction on the surface of a material. Ion trapping rate when a filter aid of a stainless steel is used will be explained. Component members of the pressure vessel of a BWR of a 1100 MW class, and devices and apparatus installed in the pressure vessel are made of stainless steels. Ferrite is formed on the surfaces of those component members. Generally, the outer surface of 5000 m2 of a reactor core corresponds to a reactor water purifying apparatus using a low-temperature ion-exchange resin of a 1%-equivalent capacity when the reactor water has a high Ni ion concentration or to a reactor water purifying apparatus using a low-temperature ion-exchange resin of a 4%-equivalent capacity when the reactor water has a low Ni ion concentration (value at 1 EFPY and removing rate is proportional to t−1/2, where t is time). When stainless steel hollow membrane pipes and a stainless steel filter aid are used and the reactor water has a high Ni ion concentration, a necessary surface area for 8%-equivalent capacity is 40,000 m2. Calculated values for a 100 μm diameter spherical shape, a 10 μm diameter spherical shape and 1 μm diameter spherical shape are 15 m3, 1.5 m3 and 0.15 m3, respectively. When the reactor water has a small Ni ion concentration, calculated values sufficient for a 100 μm diameter spherical shape, a 10 μm diameter spherical shape and 1 μm diameter spherical shape are 4 m3, 0.4 m3 and only 0.04 m3, respectively. FIG. 6 shows a high-temperature water purifying apparatus in a modification of the high-temperature water purifying apparatus shown in FIG. 5A. The high-temperature water purifying apparatus shown in FIG. 6 has composite hollow membrane pipes 49 each consisting of a hollow membrane pipe 30, a strainer 49A surrounding the hollow membrane pipe 30 so as to define an annular space between the hollow membrane pipe 30 and the strainer 49A, and filter aid particles 50 packed in the annular space. A filter aid forming the filter aid particles 50 is the same as that forming the filter aid precoatings 34 or a filter aid similar to the filter aid forming the filter aid precoatings 34. The high-temperature water purifying apparatus shown in FIG. 6, similarly to the high-temperature water purifying apparatus included in the nuclear power plant system in the second embodiment, has an upper support plate 44 and a lower support plate 44 fixedly disposed in a vessel 40A and the hollow membrane pipes 30 are extended between and held by the upper support plate 44 and the lower support plate 43. The upper support plate 44, the lower support plate 43, the hollow membrane pipes 30, the strainers 49A and the filter aid particles 50 may be assembled in a cartridge beforehand. An old cartridge can be readily replaced with a new one. The vessel 40A of the high-temperature water cleaning apparatus 40 may be divided into an upper part having a lower end provided with a flange, and a lower part having an upper end provided with a flange, and the upper and the lower part of the vessel 40A may be joined together by fastening together the flanges, which will facilitate replacing old hollow membrane pipes 30 with new ones. The high-temperature water purifying apparatus is applicable to both purifying high-temperature steam and purifying high-temperature water. A high-temperature water purifying apparatus 40 included in a nuclear power plant system in a third embodiment according to the present invention will be described with reference to FIG. 7, in which parts like or corresponding to those of the high-temperature water purifying apparatus shown in FIG. 5A are denoted by the same reference characters and the description thereof will be omitted to avoid duplication. The high-temperature water purifying apparatus shown in FIG. 6 is suitable for purifying high-temperature water in a vapor phase (i.e., steam). Referring to FIG. 7, the high-temperature water purifying apparatus 40 has a vessel 40A. A water discharge port 42 and a backwashing liquid supply port 48 are formed in a part on one side and in a part on the other side, respectively, of an upper part of the vessel 40A. A water supply port 41 is formed in the bottom wall of the vessel 40A. A drain port 47 is formed in a lower part of the side wall of the vessel 40A. A water supply line (water supply pipe) 41A is connected to the water supply port 41 to supply highly contaminated water into the vessel 40A. A water discharge line (water discharge pipe) 42A is connected to the water discharge port 42 to discharge water purified by the high-temperature water purifying apparatus 40. An upper support plate 44 is disposed horizontally in the vessel 40A on a level below that of the water discharge port 42 and the backwashing liquid supply port 48. The upper support plate 44 is provided with a plurality of openings and upper ends of hollow membrane pipes 30 are fitted in the openings of the upper support plate 44. Parts not provided with the openings of the upper support plate 44 hold upper ends of hollow water strainer pipes 51. The upper support plate 44 closes the upper ends of the water strainer pipes 51. The upper support plate 44 separates perfectly an upper space over the upper support plate 44 and a lower space under the upper support plate 44. Thus water is able to flow between the upper and the lower space only through the hollow membrane pipes 30. A lower support plate 43 is disposed horizontally in the vessel 40A on a level below that of the drain port 47 of the vessel 40A and above that of the water supply port 47. The lower support plate 43 is provided with a plurality of openings. The lower ends of the water strainer pipes 51 are fitted in the openings of the lower support plate 43. The lower ends of the hollow membrane pipes 30 are set on parts not provided with the openings of the lower support plate 43 so that the lower ends of the hollow membrane pipes 30 are closed. Water is able to flow between an upper space over the lower support plate 43 and a lower space under the same only through the water strainer pipes 51. In this embodiment, the hollow membrane pipes 30 are arranged in the shape of a hexagonal lattice and the water strainer pipes 51 are disposed at the centers of hexagonal lattices, respectively. Therefore, the ratio between the number of the hollow membrane pipes 30 and that of the water strainer pipes 51 is 2:1. Spaces between the upper support plate 44 and the lower support plate 43 are packed with filter aid particles 50. A filter aid forming the filter aid particles 50 is the same as or similar to the filter aid employed in the second embodiment. High-temperature steam supplied through the water supply port 41 into the space under the lower support plate 43 flows through the water strainer pipes 51, the filter aid particles 50 and the hollow membrane pipes 30 in that order, flows into a water collecting chamber 45 extending over the upper support plate 44 and is discharged outside through the water discharge port 42. The filter aid particles and the hollow membrane pipes 30 purify the high-temperature steam while the same flows through the vessel 40A. The specifications of the high-temperature water purifying apparatus 40 in this embodiment will be described hereinafter. The vessel 40A has a cylindrical shape. The hollow membrane pipes 30 of 25.4 mm in outside diameter and 5080 mm in length each having a filtration area of 0.405 m2 are arranged in the shape of hexagonal lattices and are extended in parallel to each other in the vessel 40A. The water strainer pipes 51 have dimensions equal to those of the hollow membrane pipes 30 and are disposed at the centers of the hexagonal lattices, respectively. If the packing ratio of the filters, i.e., the hollow membrane pipes 30 and the water strainer pipes 51, is 75%, the filters can be arranged in a density of 1500 filters per square meter. The minimum size of the filter layers, i.e., regions packed with the filter aid, between the hollow membrane pipes 30 and the water strainer pipes 51 is 2.4 mm. The reactor vessel of a BWR of a 1100 MW(E) class is 6.4 m in inside diameter, 32 m2 in sectional area and 163 m3 in volume. Therefore the reactor vessel can be packed with 48,000 or more filters. Thus 3,200 filters among the 48,000 filters, i.e., ⅔ of the 48,000 filters, are the hollow membrane pipes 30 (total filtration area: 12,960 m2). Since steal flows at 6,400 t/hr and steam is 0.036 t/m3 in specific gravity, the face velocity of steam on the filter surface is 3.8 mm/s. A filtration layer will be examined. The filtration layer is formed of spherical filter aid particles of a stainless steel arranged in an simple cubic lattice (void ratio: 48%, minimum void interval: 0.41 times the diameter of the filter aid particle) and is capable of removing 60Co ions at DF=104 calculated by using Expression (1). The radius of a void-equivalent cylinder is 0.39 μm when the diameter of the filter aid particles is 1 μm. Since the void ratio is 48%, the velocity of steam in the voids is 7.9 mm/s. Therefore, time necessary for steam to travel 1 mm is on the order of 0.13 s. Although the paths of water are assumed to be cylindrical, actually, since the filter aid particles are spherical, surface area is 1.28 times the cylindrical path. The thickness of a filter layer capable of removing ions by passing a fluid once through the filter layer will be calculated on an assumption that the chemical reaction of 60Co with steam on the surface of the filter aid particles and that of the same with water on the surface of the filter aid particles are the same. When the reactor water has a high Ni ion concentration, a filter of 5,000 m2 in surface area filters the reactor water at 60 t/hr. Since the specific weight of high-temperature water is 0.74 t/m3, high-temperature water is filtered at a volume purification rate of 81 m3/hr, which corresponds to a purifying speed of 0.0162 m/hr, i.e., 4.5 μm/s. The DF of 60Co is a function of the length L (mm) of the void cylinder of the filtration layer expressed by Expression (1)DF=e4.5×2×1.28/0.39×0.13L (1) If a filtration layer in which the distance between the water strainer pipe 51 and the hollow membrane pipe 30 is 2.4 mm is used, DF is on the order of 104. In this case, the volume of the filter aid particles is 41 m3. Water head L of a 2.4 mm thick filtration layer of 1 μm diameter spherical particles will be calculated by a method mentioned in “Physical and Chemical Processing for Waber Chemistry Control”, Asakura Shoten, p. 126. It is known from Expression (2) that the water head loss is about 200 m.h=JLν/g(1−ε)2/ε3v(σ/d)2 (2)where J (Constant of experiment): About 6 L (Thickness of filtration layer): 2.4 mm ν (Coefficient of kinematic viscosity): 0.56×10−6 m2/s g (Gravitational acceleration): 9.8 M/s2 ε (Voidage): 0.48 σ (Shape factor): About 6 v (Hollow cylinder velocity): 0.0038 m/s d (Diameter of filter aid particles): 1 μm When the reactor water has a small Ni ion concentration, the volume of the filter aid particle is something over 10 m3 and the head loss is about 50 m. In all cases, it is difficult to install a high-temperature water purifying apparatus of 104 in DF in a BWR of 285° C. Supercritical water of 650° C. react with the filter aid at a high reaction rate about 100 times a reaction rate at which water of 285° C. reacts with the filter aid. Therefore, filter aid particles of particle sizes on the order of 24 μm (volume is 0.41 m3) are satisfactory and the head loss is about 2 m and an object can be satisfactorily achieved. The amount of the filter aid particles may be increased by ten (volume of the filter aid particles is 4.1 m3 and the particle size of the same is 1 μm). In such a case, the differential pressure can be reduced by a factor of 1/10. Thus, both the increase of differential pressure and trapping capacity can be simultaneously satisfied. The high-temperature water purifying apparatus in the second or the third embodiment is installed in the pressure vessel 9 of the nuclear reactor shown in FIG. 1 or in a line extending between the pressure vessel 9 and the steam valve 1 placed at the inlet of the turbine system to separate and remove radioactive materials effectively and to reduce the migration of radioactive materials to the steam system. Since a new nuclear power plant has a large degree of freedom of design, a new high-temperature water purifying apparatus can be installed in the new nuclear power plant. In most cases, it is difficult or impossible to install a new high-temperature water purifying apparatus in an existing nuclear power plant. In such a case, improvements are incorporated into the corrugated plates of an existing dryer, for example, in a manner mentioned in connection with the first embodiment to provide the dryer additionally with the functions of a high-temperature water purifying apparatus. The high-temperature water purifying apparatus in the second or the third embodiment can be installed in the pressure vessel 9 of the nuclear reactor or in a line extending between the pressure vessel 9 and the steam valve 1 placed at the inlet of the turbine system. Usually, a steam line 11 connecting a nuclear reactor system and a steam and turbine system is provided with a stem shutoff valve, not shown, to make provision for the occurrence of a trouble in the steam and turbine system. Therefore it is highly safe to install the high-temperature water purifying apparatus at a position below the steam shutoff valve and above the steam valve 1 placed at the inlet of the turbine. When the high-temperature water purifying apparatus is installed between the pressure vessel 9 and the steam valve 1 placed at the inlet of the turbine, the line extended between the pressure vessel 9 and the steam valve 1 may be provided with a bypass line bypassing the high-temperature water purifying apparatus. Heat loss that will be caused by the high-temperature water purifying apparatus in the third embodiment is smaller than that will be caused by a reactor water purifying apparatus employing a conventional ion-exchange resin. As obvious from Table 1, the viscosity coefficient of water decreases greatly as the temperature of water rises. Therefore initial differential pressure in the high-temperature water purifying apparatus is far less than that in a low-temperature filter of the same configuration. TABLE 1VISCOSITY COEFFICIENT OF WATER AND STERMSaturatedSaturatedSupercriticalWaterwatersteamwaterTemperature25285285650(° C.)Pressure0.17.07.025(Mpa)Density10007403659(kg/m3)Viscosity891902039coefficient(μPas) The high-temperature water purifying apparatus in the embodiments employ an ion-exchange material stable in high-temperature water. Such an ion-exchange material has a long service life. The capacity of a low-temperature reactor water purifying apparatus employing the present ion-exchange resin is 2% of the flow rate of water. For example, addition of a high-temperature water purifying apparatus of an 8%-equivalent capacity can reduce the radioactive material concentration of the reactor water by a factor of ⅕ and can reduce the migration of radioactive materials to the steam system accordingly. Preferably, the iron concentration of the feed water is limited to 1 ppb or below to suppress the rise of the differential pressure in the high-temperature water purifying apparatus. Rise of the differential pressure in the high-temperature water purifying apparatus is attributable to the accumulation of particles of corrosion products in the high-temperature water purifying apparatus. Iron is the principal component of the corrosion products. Most part of iron contained in the corrosion products is contained in a leakage from a condensate purifying system. It is know from the past records of operation that the iron concentration of the feed water can be limited to 0.1 ppb or below and to about 0.02 ppb on an average by providing the condensate purifying system with a hollow fiber filter. Such an iron concentration is 1/10 or below of the iron concentration of feed water from a plant not provided with any hollow fiber filter. The high-temperature water purifying apparatus was operated experimentally in a plant not provided with any hollow fiber filter. The service life of the high-temperature water purifying apparatus was several years. It is conjectured from this fact that the service life of the high-temperature water purifying apparatus can be extended to a number of years nearly equal to the service life of the plant by suppressing the iron concentration of the feed water supplied to the high-temperature water purifying apparatus to a value not greater than a predetermined limit. Although a method of suppressing the migration of the radioactive materials to the steam system by filtering out the radioactive materials has been described, it is effective to reduce the generation of steam containing radioactive materials. It is effective in reducing the generation of steam containing radioactive materials to use water supplied from a condensate purifying apparatus or a condensate storage tank 8 shown in FIG. 1 instead of water discharged from the reactor water purifying system 5 by the head spray 6. Since the radioactive material concentration of the water supplied from the condensate purifying apparatus or the condensate storage tank 8 is smaller than that of the water discharged from the reactor water purifying system 5, steam produced by spraying water by the head spray 6 has a low radioactive material concentration. In view of a recent mode of operation to cool the nuclear reactor quickly by using a head spray when the nuclear reactor is shut down, the effect of reduction of the generation of steam containing radioactive materials is significant. It is preferable in reducing the amount of steam containing radioactive materials to direct the nozzles of the head spray toward in-pile structures so that water sprayed by the head spray may not fall directly on the pressure vessel heated at a high temperature, to interpose a cover between the inner surface of the pressure vessel and the nozzles of the head spray so as to cover the inner surface of the reactor vessel from sprayed water, and to define a region in which water is sprayed, to spray water in small particles by using appropriate nozzles or an ultrasonic spraying device and to control the amount of water to be sprayed according to the amount of steam generated by heat generated after shutdown. Preferably, the head spray 6 has a shower head provided with a plurality of nozzle holes of a diameter not greater than 1 mm arranged in an area of about 400 cm2, and water is sprayed so as to wet the dryer entirely and all the sprayed water fall on the dryer. The systems now in use uses water of a temperature nearly equal to that of the pressure vessel delivered from the reactor water purifying system in order that relatively low thermal stress may be induced in the pressure vessel when the reactor vessel is wetted with water sprayed by the head spray. However, the dropping rate of the temperature of the pressure vessel is lower than that of the temperature of the reactor water. Consequently, the temperature difference between the pressure vessel and the reactor water increases with time. Therefore, this method is able to lower the temperature of the nuclear reactor safely and efficiently by suppressing the spread of water sprayed by the head spray, spraying water in small drops and bringing steam of a low temperature into contact with the pressure vessel according to the amount of steam generated after shutdown. |
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claims | 1. A core for a nuclear reactor, the core including fissile material for generating power and having an edge beyond which no fissile material extends, wherein the core comprises:a plurality of nuclear fuel assemblies including the fissile material; anda fluence controlled nuclear fuel assembly at the edge of the core, wherein the fluence controlled nuclear fuel assembly includes the fissile material and a fluence control structure integrally within the fluence controlled nuclear fuel assembly only at the edge of the core, wherein the fluence control structure is fabricated of materials that reduce neutron flux so as to reduce neutron flux beyond the edge of the core, and wherein the plurality of nuclear fuel assemblies lack the fluence control structure and are a same configuration as, and interchangeable with, the fluence controlled nuclear fuel assembly in the core. 2. The core of claim 1, wherein the core is cylindrical, and wherein the edge is a radial perimeter of the cylindrical core. 3. The core of claim 1, wherein the fluence control structure nuclear fuel assembly is fabricated of materials that maintain their flux-reduction when exposed to radiation in the core when operating. 4. The core of claim 1, further comprising:a core shroud surrounding the nuclear fuel assemblies and the fluence controlled nuclear assembly, wherein the edge is directed toward the core shroud with no nuclear fuel assembly intervening. 5. The core of claim 3, wherein the fluence control structure consists essentially of a material having a thermal neutron absorption cross section of at least 2 barns. 6. The core of claim 5, wherein the material is an alloy containing iron, hafnium, cadmium, or combinations of any thereof. 7. The core of claim 1, wherein the fluence control structure is a member of the group including a shielding channel, a plate curtain, and a shielding fuel rod. 8. The core of claim 1, wherein the fluence controlled nuclear fuel assembly includes a plurality of nuclear fuel rods and a channel surrounding all fuel rods in the fluence controlled nuclear fuel assembly, and wherein the fluence controlled nuclear fuel assembly includes a plurality of the fluence control structures, including,a shielding channel that is the channel surrounding all fuel rods in the fluence controlled nuclear fuel assembly,a plate curtain attached to a face of the channel of the fluence controlled nuclear fuel assembly, anda shielding fuel rod that is an unfueled, solid rod with no internal cavity, wherein the shielding fuel rod is aligned with the plurality of fuel rods and surrounded by the channel in the fluence controlled nuclear fuel assembly. 9. The core of claim 7, wherein the shielding fuel rod is a member of the group including a segmented rod housing a non-fuel irradiation target, an empty rod, and a solid rod having no internal cavity. 10. The core of claim 8, wherein the plurality of the fluence control structures consist essentially of non-burnable materials maintaining a thermal neutron absorption cross section of at least 2 barns following exposure to radiation in the core while operating. |
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claims | 1. An image-guided radiation therapy (IGRT) apparatus, comprising:a medical imaging device integrated with a linear accelerator, the linear accelerator configured for emitting a radiation beam which is shaped by a beam shaper; anda gantry configured to rotate the emitted radiation beam about an axis,wherein the beam shaper is mounted to an adjustment arm configured to adjust the beam shaper between a first position and a second position, andwherein the medical imaging device is an MRI device generating a magnetic field, and the second position is a radial distance away from the magnetic field such that a magnetic field strength generated by the MRI device at the second position is lower than a magnetic field strength generated by the MRI device at the first position. 2. The IGRT apparatus as claimed in claim 1, wherein the first position is a treatment position, and the second position is a service position. 3. The IGRT apparatus as claimed in claim 1, wherein the position of the linear accelerator is moveable with the beam shaper between the first and second positions. 4. The IGRT apparatus as claimed in claim 1, wherein the beam shaper is a multi-leaf collimator. 5. The IGRT apparatus as claimed in claim 1, wherein the adjustment arm is connected to a fixed body through a linkage and operable to move the beam shaper between the first and second positions. 6. The IGRT apparatus as claimed in claim 5, wherein the fixed body is integral with the gantry, and wherein the linkage comprises a pivot operable to pivot the beam shaper from a position within the gantry to a position removed from the gantry. 7. The IGRT apparatus as claimed in claim 5, further comprising an actuator to move the adjustment arm between the first and second positions. 8. The IGRT apparatus as claimed in claim 7, wherein the actuator comprises a pneumatically or hydraulically operated component. 9. The IGRT apparatus as claimed in claim 8, wherein the actuator comprises at least one electro-mechanical actuator. 10. The IGRT apparatus as claimed in claim 9, further comprising a gearing system. 11. The IGRT apparatus as claimed in claim 5, wherein the linkage comprises a multi-axis joint. 12. The IGRT apparatus as claimed in claim 5, further comprising a locking mechanism for locking the beam shaper into position during delivery of radiation treatment. 13. The IGRT apparatus as claimed in claim 5, wherein the adjustment arm is provided with one or more joints operable to present the beam shaper in an increased number of positions and orientations. 14. The IGRT apparatus as claimed in claim 5, wherein the adjustment arm incorporates a linear actuator allowing the length of the arm to be adjusted thereby providing further flexibility in the positioning of the beam shaper when removed from the operational position. 15. The IGRT apparatus as claimed in claim 1, wherein the IGRT apparatus has a closed drum configuration. 16. The IGRT apparatus as claimed in claim 1, wherein the IGRT apparatus has an open ring configuration. 17. A radiation therapy apparatus for delivering radiation therapy to a target, the apparatus comprising:a linear accelerator configured to emit, via a radiation beam emitter, a radiation beam along a beam path toward the target;an imaging device configured to obtain an image of the target;a beam shaper configured to shape the emitted radiation beam emitted toward the target;a gantry coupled to the linear accelerator, wherein the gantry is configured to rotate the radiation beam emitter about an axis; andan adjustment arm connected to the gantry and to the beam shaper, wherein the adjustment arm is configured to adjust the beam shaper between a treatment position and a non-treatment position, andwherein the medical imaging device is an MRI device generating a magnetic field, and the non-treatment position is a radial distance away from the magnetic field such that a magnetic field strength generated by the MRI device at the non-treatment position is lower than a magnetic field strength generated by the MRI device at the treatment position. 18. The radiation therapy apparatus as claimed in claim 17, wherein the adjustment arm is connected to a fixed body through a linkage and operable to move the beam shaper between the treatment and non-treatment positions. 19. The radiation therapy apparatus as claimed in claim 18, wherein the fixed body is integral with the gantry, and wherein the linkage comprises a pivot operable to pivot the beam shaper from a position within the gantry to a position removed from the gantry. 20. The radiation therapy apparatus as claimed in claim 17, wherein the imaging device has an open ring configuration. |
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052272680 | description | DETAILED DESCRIPTION OF THE INVENTION The X-ray mask of the subject invention shall be explained, referring to the figures. FIG. 1A through FIG. 1F are production process cross-section figures of the X-ray mask of one embodiment of the subject invention. In FIG. 1A through FIG. 1F, 2 is the silicon nitride film, 3 is the tantalum film and 3a the first alignment pattern, while 6 are the alignment marks using photosensitive resin. The X-ray mask of the subject embodiment, as shown in its cross-section in FIG. 1E, has formed on the surface on the 2 micrometer thick SiN film 2, which is the X-ray permeable film, a circuitry pattern 3b and first alignment pattern 3a composed of 0.7 micrometer thick tantalum film which is the X-ray absorbant pattern. Further, via the SiN pattern 2, there are formed on the back surface of this SiN film 2 second alignment pattern 6 of photosensitive resin having the same pattern as and opposing the first alignment pattern 3a. As against the prior art X-ray mask, by forming the second alignment pattern on the laser beam side as well, the laser beam emitted from the alignment optical system does not pass through the SiN film 2 but is directly diffracted by the second alignment pattern 6 so that the attentuation of the laser beam strength is quite small. We shall now explain simply the method of manufacturing the X-ray mask of the subject embodiment, based on the manufacturing process cross-sections in FIG. 1A through FIG. 1F. As shown in FIG. 1A, a prior art X-ray mask 11 (FIG. 4) consisting of silicon support framework 1, the SiN film 2 (which is to become the X-ray permeable film) and the tantalum film 3 (which is to become the X-ray absorbant pattern) is formed. 3a is the prior art alignment pattern and 3b is the circuitry pattern. Next, on the silicon nitride film 2 a uniform coat of photosensitive material 4 (positive photoresist, for instance) is deposited using a spinner, etc. At this time, the film thickness of the photosensitive material 4 is so set to maximize the diffraction efficiency which is governed by the structure of the alignment optical system. At this point, an explanation will be given for the setting of the thickness of the photosensitive material 4, referring to FIG. 2 and FIG. 3. FIG. 2 is a cross-section figure of a lamella type diffraction grating. As shown in FIG. 2, if the incoming laser beam's wavelength is .lambda., the incoming angle is .alpha., the diffraction angle is .beta., the pitch of the diffraction grating is d, and the height of the diffraction grating is h, the diffraction efficiency .eta. can be expressed by the following expression from the lamella diffraction grating's theory: EQU .eta.=(400/m.sup.2 .pi..sup.2) . cos.sup.2 [(.delta.'+m .pi.) / 2 ] Where, .delta.'=2.pi.h/.lambda. (cos .alpha.+cos .beta.) m .lambda.=d (Sin .alpha.-Sin .beta.) and where m is the order of the diffracted beam. The diffraction efficiency .eta. changes cyclically depending on the height h of the diffraction grating and has maximum and minimum values. FIG. 3 shows the relationship when a helium neon laser (.lambda.=633 nm) is used as the laser beam with a diffraction grating with pitch d=4 micrometers and the incoming angle is set at approximately 10 degrees, between the diffraction efficiency .eta. of the zero order diffracted beam, the diffraction efficiency .eta. of the first order diffracted beam and the height h of the diffraction grating. Since the first order diffraction beam is usually used as the alignment beam, from FIG. 3 we can see that by setting the height h of the diffraction grating at 0.16 micrometer, 0.48 micrometer, etc., a maximum diffraction efficiency .eta. of approximately 40% is obtained. The height h of the diffraction grating thus obtained will be the film thickness of the photosensitive material 4. In this manner the height of the diffraction grating can be freely selected to increase the diffraction efficiency, and, since alignment marks are formed on the laser beam side of the SiN film the attenuation of the alignment beam is small, and a sufficiently strong diffracted beam can be obtained. Next, as shown in FIG. 1C, by using a light 5 to which the photosensitive material 4 is sensitive (for example, a light such as a mercury lamp) the entire surface is exposed at once using as the mask the tantalum film 3 which will be the X-ray absorbant pattern. Since the thickness of the silicon nitride film 2 which will be the X-ray permeable film is usually around 2 micrometers, the method becomes the same as contact exposure and the X-ray absorbant pattern will be accurately and equimultiply transferred to the photosensitive material 4 on the reverse side. Naturally, the first alignment pattern 3a composed of the first X-ray absorbant pattern will also be transferred without slippage in position. In this way, a second alignment pattern 6 and a second LSI circuit pattern are transferred to the opposite positions corresponding to the first alignment pattern 3a and the first LSI circuit pattern 3b respectively as illustrated in FIG. 1D through a development process. Then, before application of baking, another light exposure is applied to the unnecessary pattern of the second LSI circuit pattern only such that the second alignment pattern 6 is left, as illustrated in FIG. 1E, after development and baking. As a matter of course, instead of going through development immediately after the first entire surface exposure performed from the first main surface side of the x-ray permeable film 2 and by going through development after an exposure from the second main surface side of the x-ray permeable film 2 with the second alignment pattern 6 being masked, the second alignment pattern 6 only can be transferred at a time as shown in FIG. 1E skipping the stage of FIG. 1D. In this manner, according to the embodiment's X-ray mask, by using the second alignment pattern 6 as beam refracting beam elements, a sufficiently strong refracted beam is created when illuminated by the laser beam and accordingly a high resolution power can be obtained. That is to say, this X-ray mask would have second alignment pattern 6 having an optimized form to maximize the diffraction efficiency versus the alignment optical system. Thus a sufficient alignment signal strength is obtained, making possible the realization of a high alignment accuracy. Also, the process of forming the second alignment pattern 6 can be done just by uniformly coating the photosensitive material 4 and by exposing the entire surface at once. This does not require specialized equipment and is very easy to accomplish. Moreover, in the transfer of the X-ray absorbant pattern, as it is the same in principle as contact exposure, the pattern position distortion accompanying the transfer of the X-ray absorbant pattern is at such a low level as to be virtually insignificant and has no effect at all as a factor in alignment errors. Accordingly, as a X-ray exposure mask requiring high precision alignment it is possible to realize a high performance X-ray mask provided with second alignment pattern having sufficiently high optical characteristics. The X-ray mask of the subject invention, by forming the second alignment pattern for the mask on the other surface of the X-ray permeable film through equimultiple transfer using the first alignment patterns for the mask and substrate and the self-adjustment method, can obtain a sufficiently strong alignment signal. And, it offers great improvement in alignment accuracy and it also results in superior alignment accuracy and it also results in superior industrial productivity. Also, since the second alignment pattern are also formed on the laser beam source side with a height which increases the diffraction efficiency, a sufficiently strong alignment signal is obtained. Furthermore by using the X-ray mask of the subject invention to expose to the semiconductor wafer the circuitry pattern formed over the mask, there will be virtually no positional slippage between the mask and the wafer so that the desired circuitry pattern can also be easily exposed. |
claims | 1. An apparatus for generating a gas-cluster beam, comprising:a gas expansion nozzle mounted in a chamber to cause gas clusters from the expansion nozzle to form a beam passing through the chamber in a predetermined direction and through an aperture at an end of the chamber,wherein the chamber is formed by one or more surfaces surrounding the beam and aperture and located to deflect gas clusters and molecules from the nozzle that are not traveling within and aligned with the beam away from the beam and towards an opposing predetermined direction. 2. The apparatus of claim 1, wherein the one or more surfaces include a conical first surface coaxially surrounding the beam and angled towards the opposing predetermined direction. 3. The apparatus of claim 2, wherein the one or more surfaces include a second flat surface surrounding the aperture and facing the opposing predetermined direction. 4. The apparatus of claim 1, wherein the one or more surfaces include one or more third surfaces facing away from the beam and located immediately surrounding the beam to deflect gas molecules and clusters traveling at more than a predetermined distance from the beam away from the beam. 5. The apparatus of claim 1, further comprising a vacuum apparatus located behind the expansion nozzle for evacuating deflected gas molecules and clusters that are not part of the beam from the chamber in the opposing predetermined direction. 6. The apparatus of claim 1, wherein the gas expansion nozzle is mounted at opposing input and outlet ends using a limited number of elongated members extending from sides of the chamber to allow easy flow of gas molecules and clusters that are not part of the beam in the opposing predetermined direction. 7. The apparatus of claim 6, wherein the gas expansion nozzle is adjustably mounted at the outlet end of the nozzle to enable adjustment of the predetermined direction. 8. The apparatus of claim 7, wherein the gas expansion nozzle is tiltably mounted at the input end of the nozzle to support adjustment of the predetermined direction at the outlet end of the nozzle. 9. The apparatus of claim 1, wherein the one or more surfaces has substantially the shape of a cone or a pyramid or a elliptic paraboloid or an ellipsoid. 10. The apparatus of claim 1, further comprising a second chamber surrounding the gas cluster beam beyond the aperture and the first said chamber and having a second aperture located for allowing further flow of the gas cluster beam. 11. The apparatus of claim 10, further comprising one or more fourth surfaces facing away from the beam and located immediately surrounding the beam at the second aperture for deflecting gas molecules and clusters traveling at more than a predetermined distance from the beam away from the beam. 12. The apparatus of claim 11, wherein the gas expansion nozzle is mounted at input and outlet ends, and further wherein the outlet end is adjustably mounted to enable adjustment of the predetermined direction. 13. The apparatus of claim 10, wherein the second chamber is formed by at least one plane surface oriented at an angle of from 30° to about 60° with respect to the gas cluster beam and adapted to direct gas molecules and clusters that are not part of the beam away from the beam. 14. The apparatus of claim 1, wherein the one or more surfaces surrounds substantially all of the beam located within the chamber. 15. A gas-cluster ion-beam processing apparatus comprisingthe gas-cluster beam generator apparatus of claim 1 for generating a gas-cluster beam;an ionizer for ionizing at least a portion of the gas-cluster beam to form a gas-cluster ion-beam having a path; anda workpiece holder for supporting a workpiece in the path of the gas-cluster ion-beam. 16. The gas-cluster ion-beam processing apparatus of claim 15, further comprising a differential pumping chamber having a plane surface oriented at an angle of from about 30 degrees to about 60 degrees with respect to a gas-cluster beam trajectory and adapted to direct at least a portion of un-clustered gas into a vacuum pump. 17. A method for generating a gas-cluster beam, comprising the steps of:directing a gas expansion nozzle into a chamber to cause gas clusters from the expansion nozzle to form a beam passing through the chamber in a predetermined direction and through an aperture at an end of the chamber;deflecting gas clusters and molecules from the nozzle that are not traveling within and aligned with the beam away from the beam and towards an opposing predetermined direction using walls of the chamber that surround the beam and aperture; andcreating a vacuum behind the expansion nozzle for evacuating deflected gas molecules and clusters that are not part of the beam from the chamber. 18. The method of claim 17, wherein the step of directing includes adjustably mounting the outlet end of the nozzle and adjusting the predetermined direction. |
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description | The present invention relates generally to a light water nuclear reactor and more specifically to a method of operating a nuclear reactor based on the modeling of deposits on heat transfer surfaces. Power generators, including nuclear reactors, are used for power generation, research and propulsion. A power generation circuit generally includes a heat source such as a nuclear core or furnace and a coolant circuit. For light water reactor, respective coolant piping circuits transport the heated water or steam to either a steam generator and then a turbine, or directly to a turbine, and after going through a condenser (heat sink), carries circulating or feedwater back to the heating source. Operating temperatures and pressure may range up to or above the critical point of water. Depending on the operational conditions, the various materials used must withstand various load, environmental and radiation conditions. Material used as coolant piping and other circuit and heat source components include but are not limited to carbon steels, stainless steels, nickel-based and other alloy steels and zirconium based alloys. These materials have to withstand the high temperature and high pressure conditions. Although the materials have been carefully selected, corrosion occurs caused by the corrosive nature of the environment: high temperature, high pressure water, steam, water radiolysis, additives in water and radiation effects. Such corrosion processes limit the lifetime of the systems in contact with the coolant fluid, and include but are not limited to stress corrosion cracking, flow accelerated corrosion, crevice corrosion, erosion corrosion, generalized corrosion and nodular corrosion. Stress corrosion cracking (SCC), including intergranular stress corrosion cracking (IGSCC), is a well-known phenomenon happening to structural components in coolant circuits of a nuclear reactor, which affects the base and welding materials. SCC occurs through crack initiation, and propagation, which are caused by a combination of chemical, tensile and ductile stresses (static and dynamic). Such stresses are common in nuclear environments caused by thermal expansion and contraction, residual stresses from welding, cold working, etc. The susceptibility toward SCC is often increased by the operating coolant environment, welding, heat treatment, radiolysis and radiation. High oxygen content in the coolant fluid has been shown to accelerate SCC through higher rates of crack initiation and propagation. High oxygen content in the coolant fluid can stem from oxygen intrusion and water radiolysis processes, which create highly oxidizing species such as oxygen radical, hydrogen peroxide and many other radical species in the gamma, neutron, beta, and alpha flux. Corrosion products present in the coolant fluid ultimately accumulate on the heat transfer surface, for instance on surfaces formed of zirconium of the fuel elements of a nuclear reactor core or on internal surfaces of steam generator tubes made of stainless steel, forming a deposit layer commonly called crud. The structure of the deposit layer varies within its thickness and comprises an outer portion of low density loose crud, harboring mostly water, which is in constant exchange with the circulating reactor water, but providing a metal oxide structure capable of attracting and retaining colloidal particulates. This portion of low density loose crud is called fluffy crud. Below the portion of fluffy crud, closer to the heat transfer surface, the deposit layer comprises a inner portion of higher density crud, called tenacious crud, stuck to the heat transfer surface. The tenacious crud forms on a metal oxide layer of the heat transfer surface, which forms on heat transfer surface due to heating of heat transfer surface (i.e., general corrosion). For example, on fuel element surfaces formed of zirconium, heating results in the increase of a native zirconium oxide layer. The fraction of tenacious crud in the deposit layer increases as crud deposition increases and the crud ages. The densification is accelerated by excessive heat and prolonged exposure to reactor environment. The sponge-like nature of the deposit layer creates conditions corresponding to capillary water movement. The very low capillary velocities of fluids in crud, creating almost confined conditions, favor the water radiolysis reactions that form the molecular species, i.e. hydrogen, oxygen, hydrogen peroxide and the HO radical. Studies, such as S. Le Caër et al., Hydrogen Peroxide Formation in the Radiolysis of Hydrated Nanoporous Glasses: A Low and High Dose Study, Chem. Phys. Lett. 450 (2007) 91-95, have shown that the hydrogen in confined spaces is ineffective in facilitating the recombination reaction to water. Hence, in confined spaces the sum of the oxidizing species, i.e. oxygen, hydrogen peroxide and oxygen radical, effectively create an oxygen saturated environment. The amount and form of the deposit layer formed on the heat transfer surfaces depends on the concentrations and types of the chemical elements in the water to be converted to steam. The elements are typically in the form of particulate, colloidal and/or ionic species. As the water is converted to steam, the chemical, physical and thermodynamic processes will work in concert (interactively) to produce the evolution of the deposit layer. Over the years, there have been a number of efforts to understand and model the evolution of the deposit layer and the resulting heat transfer performance. The deposit typically evolves as a porous layer. Heat transfer through the deposit layer is primarily a combination of conduction through the deposit and water matrix and convection through water in the matrix which is converted to steam. Theories and models have focused on a concept of small capillaries within the porous matrix that conduct water to larger diameter openings called “steam chimneys,” where the water is converted to steam. The steam then travels from the steam chimney into the coolant fluid convectively transferring the heat of vaporization. A fixed diameter was used to delineate the openings that were assumed to be capillaries and those that were assumed to be steam chimneys. U.S. Pat. No. 7,420,165 teaches a method of calculating the power transfer of a nuclear component based on a number of steam chimneys in a deposit layer on the nuclear component. Under most conditions, deposits on heat transfer surfaces make the heat transfer less efficient, and increase the potential for thermal or corrosion damage of the heat transfer surface. Modeling efforts provide a better understanding of the deposition phenomenon and thus help in the development of mitigative and corrective actions. Although some of the earliest models of deposits on heat transfer surfaces treated the deposit as a layer with a modified coefficient of thermal conductivity, it was soon realized that the transfer of heat through a porous deposit layer was much more complex than simple conduction. Along these lines, the wick heat transfer model was developed. The wick heat transfer model accounted for the fact that heat transfer in a porous deposit is a combination of conductive and convective heat transfer. The conduction is through the deposit matrix and the convection is from the movement and heating of the coolant fluid within the deposit matrix. The primary convective heat transfer is from the movement of coolant fluid into the deposit matrix where it becomes steam and returns to the coolant fluid. FIG. 1 shows a version of the wick heat transfer model illustrating a deposit layer 10, which has a thickness X on a heat transfer surface 12. Solutes in the flowing coolant fluid 14 are carried into pores formed in the outer surface 16 of deposit layer 10 by a network of small diameter “capillaries” 18 which fed the fluid 14 into “steam chimneys” 20 where the water was able to absorb the latent heat of vaporization and move back into the fluid 14 as steam. The diameter of a given opening within the deposit layer 10 was used to define whether the path would serve as a capillary 18 supplying fluid 14 or as a steam chimney 20 where the fluid 14 was converted to steam. The smaller diameter openings were capable of wicking the fluid into the hotter regions of the deposit layer 10 without boiling. If the fluid from a capillary 18 connects to the larger diameter of a steam chimney 20, the larger flow opening allows the fluid 14 to flash to steam and flow out to the fluid 14. Any deposit surface openings larger than a specific diameter were counted as steam chimneys 20 and empirical relationships were derived to relate the number of steam chimneys 20 to the heat flux capacity of the deposit layer 10. A method of operating a nuclear reactor is provided. The method includes defining a layer increment of a deposit layer modeling a deposit on a heat transfer surface of the nuclear reactor; periodically updating a thickness of the deposit layer by adding the layer increment to the deposit layer; recalculating properties of the deposit layer after each layer increment is added to the deposit layer; determining a temperature related variable of the heat transfer surface as a function of the recalculated properties of the deposit layer; and altering operation of the nuclear reactor when the temperature related variable of the heat transfer surface reaches a predetermined value. A method of modeling a deposit layer on a heat transfer surface of a nuclear reactor is also provided. The method includes defining a geometry of layer increment of a deposit layer modeling a deposit on the heat transfer surface of the nuclear reactor; periodically updating a thickness of the deposit layer by adding the layer increment to the deposit layer; recalculating properties of each layer increment after each new layer increment is added to the deposit layer; and displaying at least one of the recalculated properties on a display device. Studies of inner and outer surfaces of deposit layers 10, i.e. surface of deposit in contact with a heat transfer surface 12, respectively with a coolant fluid 14, at using image processing software have illustrated that openings in deposits on heat transfer surfaces are more accurately represented as a distribution of different sized channels, rather than as two separate and distinct populations of capillaries and steam chimneys. One such study is Pop et al., PWR Fuel Deposit Analysis at a B&W Plant with a 24 Month Fuel Cycle, 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, Aug. 7 to 11, 2011. The physics and chemistry of the system determine which channels within the deposit layer 10 are large enough to allow the escape of steam and which pores are small enough to provide the capillary driving force to draw water into the region of the deposit layer where it is converted to steam. The chemical and heat transfer processes involved determine the evolution of the deposit layer porosity and structure as more water is drawn in to the deposit layer 10 and converted to steam. Modeling deposit layers 10 on heat transfer surfaces 12 according to embodiments of the present invention use a population of channels whose members are defined in terms of a quantity and an initial radius. The deposit layer 10 is defined in terms of an area of heat transfer surface 12, for example a square meter. For each new layer increment, the population membership is constructed using a minimum radius and a radius increment to determine the member radii and quantities such that the summed total area of the entire channel population does not exceed the defined area of the heat transfer surface times the thickness of the layer increment. This approach more closely represents the observed deposit structure and yields more realistic initial porosity and solid fraction values for each new layer increment. The deposit model considers the evolution of the deposit layer 10 to be an ongoing process. As particulates in the coolant fluid 14 in contact with the heat transfer surface 12 attach to the heat transfer surface 14, the thickness of the deposit layer 10 grows. As the deposit structure forms, the fluid 14 and contained species continue to be drawn into the deposit layer 10 and the fluid to be converted to steam. As the fluid 14 within a deposit channel is converted to steam, some of the contained species precipitate and the channel may reach a minimum radius at which the channel is not able to carry the same substances out of the deposit. This continued deposition within the deposit structure reduces the open volume in the channels where the fluid is converted to steam. Over time, the layer increments of the deposit that are closer to the heat transfer surface 12 continue to become less porous due to the continued deposition of particulate, colloidal and ionic substances from the fluid 14. From known measurements on the surface of the deposit in contact with the coolant fluid 14, there is a population of openings which can be characterized as a distribution. Since this population will be the most recently formed deposit structure, it is used to define a typical starting condition for each layer increment in the deposit model. The deposit model is based on an observed distribution of deposit openings on the coolant fluid boundary of the deposit layer 10, for example using a Scanning Electron Microscope (SEM) as described for example in U.S. Pat. No. 7,822,259. FIG. 2 shows a plan view of an example of a distribution of different size channels 22 in a deposit layer increment 24 as it may be observed on an actual flake of a deposit layer 10 removed from a heat transfer surface of a nuclear reactor. As shown in FIG. 2, the channels 22 vary in diameter throughout layer increment 24. The channels 22 in combination with the local heat flux and chemistry conditions determine the thermal hydraulic functioning of the different openings. FIG. 3 shows an exemplary plot of actual observed population data of a distribution for use in creating a model of a deposit layer increment. The plot shows the population of channels 22 for each sized radius in the distribution. As shown in this example, the majority of the channels 22 have a radius that is less than 2 microns. Different population distributions are applied if the observed deposit flake conditions are different. FIG. 4 shows an enlarged view of a portion of the exemplary plot of FIG. 3 around 1 to 2 microns with the corresponding calculated regression curve. In the deposit model, each new layer increment 24 is added on top of the previous layer increments 24, adjacent to the coolant fluid 14, and starts with the porosity and physical structure defined by the observed porosity of an outer layer of an actual deposit sample, such as for example structure represented by the distribution shown in FIG. 3. Flow of the coolant fluid 14 into the older layer increments 24 continues through the smaller channels 22 (those acting as capillaries). The fluid flashes to steam in the larger channels 22 (those acting as chimneys) leaving behind most of the substances, the insoluble species, that were in the fluid 14 and the initially soluble species transformed to insoluble species due to the localized conditions. The continued deposition within these larger channels 22 reduces the open volume. This process of consolidation results in the reduction of the porosity of the layer increment 24 from its initial condition. The radius of the larger channels 22 is reduced by deposition within the channels 22. Since the effect is cumulative, the porosity reduction will be more significant in the older layer increments 24 (nearest the heat transfer surface 12). The rate of the deposit buildup is a function of the coolant fluid chemistry and the heat flux. Due to the thermodynamics of the deposit layer 10, there is a limit on the smaller radius of a channel within the deposit. When the channel becomes too small to allow the coolant fluid to convert to steam, the rate of deposition slows or stops. Thus there is a minimum radius limit from which channels are considered to act as chimneys. FIG. 5 schematically shows a perspective view of a representation of deposit layer increment 24 of the deposit layer 10, which for example could have a thickness on the order of 1 micrometer depending on the model parameters. It should also be noted that the actual channel diameter and number of channels 22 in a layer increment 24 would be consistent with the channels 22 represented in FIG. 3. FIG. 5 illustrates the channels 22 as being completely cylindrical throughout their length; however, each channel 22 has some tortuosity on a macro scale as illustrated in FIG. 6 with diameter variations such that channels 22 are modeled in the deposit model as being irregular open volumes, not perfect cylinders. As a layer increment 24 of the deposit layer 10 is modified over time, the tortuosity effect may be increased. Adding material to an irregular open volume changes the direction of capillary flow or steam evacuation through those free spaces, resulting ultimately in a longer fluid or steam path until the exit point. FIG. 7 shows the evolution of the layer increments 24 by illustrating plan views of three different layer increments 24 of a deposit layer 10. The top illustration shows a top or newest layer increment 40, recently formed on the outer surface of the deposit layer 10 in contact with the fluid 14. The middle illustration shows a middle deposit layer increment 42, which is approximately located halfway between the newest layer 40 and the heat transfer surface 12. The bottom illustration shows a bottom or oldest deposit layer increment 44 formed directly on the heat transfer surface 12. These three layer illustrations show the channels 22 in the deposit layer 10 reduce in radius due to further deposition within the layer increments until each channel achieves a minimum radius. FIG. 8 shows a layer increment 46 in which each channel 22 has achieved the minimum radius because the high concentrations of deposited substances in the channels 22 of the layer increment 46. In this case, the layer increment 46 will be primarily a conduction layer and the heat transfer of this particular layer increment 46 is thus at a minimum. FIG. 9 shows a flow chart illustrating a method of operating a nuclear reactor in accordance with an embodiment of the present invention. In a step 102, an actual deposit sample is obtained from the heat transfer surface of a nuclear reactor, for instance a fuel element (i.e., a fuel rod or fuel pin) or a steam generator tube. In a step 104, the actual deposit sample is analyzed using a SEM to determine the porosity of the deposit sample. Specifically, the actual crud deposit sample is analyzed to determine the porosity of a layer segment of the deposit sample. In a preferred embodiment, the layer segment is an outer layer segment of the deposit sample, which formed a surface of the deposit interacting with coolant fluid in the reactor core. For example, the layer segment may be one micrometer thick. A model of an actual deposit forming on heat transfer surface is initiated by a computer. The computer builds the modeled deposit layer 10 in layer increments 24 which allow a quasi-static equilibrium analysis process that follows the formation of the actual deposit in the core of an actual operating nuclear reactor. The model may maintain the quasi static equilibrium condition by assuming that the heat available at the heat transfer surface 12 is transferred through the modeled deposit layer to the coolant fluid 14. While an exception may be granted during short transient periods, the system adapts to transfer the available heat. The adaptation is typically in the form of an increased temperature at the heat transfer surface 12 until the layer is again capable of transferring all of the heat away from the heat transfer surface 12. The combined conduction and convection of the modeled deposit layer 10 are equal to the input heat flux. The model in this embodiment used an iterative equilibration of thermal hydraulic, chemical and physical deposition balances over a defined operational period. By using small time and volume increments the modeled deposit layer 10 is maintained near a quasi-equilibrium state as the deposit is formed and as the density of the older layers, those closest to the heat transfer surface 12, increases. In a step 106, the computer defines the geometry of a layer increment 24 of the modeled deposit layer 10 by setting an initial channel population for the layer increment 24 based on the porosity of layer segment analyzed by SEM. The defined layer increment 24 is used to start each new layer increment 24 added to the modeled deposit layer 10. An exemplary embodiment of numerical values for the initial channel population in shown in FIG. 3, which defines the initial channel population in the number of channels 22 and the radii of the channels 22. Depending on the minimum radius and radius increment of the initial channel population, which may be arbitrarily set to reasonably represent the layer segment of the actual deposit, the computer may create layer increments 24 so each new layer increment 24 has an initial porosity, i.e., a fraction of volume of voids over the total volume, between 0.90 and 0.95, which is in agreement with observed SEM data. The example of FIG. 3 uses a minimum radius of 3.0 E-7 meters and a radius increment of 2.5 E-8 meters to yield a discrete population distribution of fifty three (53) different channel radii with the maximum radius of 1.625 E-6 meters and a porosity of 0.91. After the layer increment 24 is defined, the computer initiates an iterative process to model the growth of the actual deposit on the heat transfer surface of the fuel rod over time. Each layer increment 24 is defined as composed of a porous matrix consisting of water, solids and a distribution of open channels 22. The porous matrix will transfer heat by conduction through the combined solid and liquid matrix. The open channels 22 will transfer heat through evaporation of liquid oozing on their interior surfaces and the subsequent convective transfer through evacuation of the resulting steam. Thus each layer increment 24 consists of three phases—solid, liquid and vapor—of material in proportions determined by the model. The porous solid of the deposit layer 10 forms the physical matrix of the layer. The liquid material permeates the porous solid and the smaller channels 22 conduct liquid into the matrix. The larger channels 22 contain the steam component of the material. The volumetric proportions of each phase evolve over time as the model iteration progresses. This balance between solid, liquid and vapor is used to determine the heat transfer of the combined deposit mass. As the model iteration continues, new layer increments 24 are added and the older layer increments continue deposition and reduction of the channel radii (or diameters) until the minimum radius is reached. The model then iterates the process of waiting until the deposit layer 10 grows by an amount having the thickness of a layer increment 24, creating the layer increment 24 based on the observed open structure and porosity, calculating the temperature profile across the full deposit layer thickness, determining the deposition within the openings of the previous layer increments 24 and determining the new chemical equilibrium conditions for the full deposit layer 10. This iteration is continued for a specified duration and used to monitor at least one temperature related variable of the heat transfer surface 12 to ensure that the heat transfer surface is not heated to a dangerously high value. More specifically, a preferred embodiment of the modeling operates in accordance with the following steps 108 to 120. In a step 108, after a sufficient amount of time has elapsed that the computer estimates a deposit thickness equal to the layer increment thickness has been added to the heat transfer surface 12, the computer adds a layer increment 24 having the predefined geometry to the deposit layer 10. Initially, during a first iteration, a first layer increment 24 of the deposit layer 10 representing an initial layer segment formed directly on the heat transfer surface 12 of the nuclear reactor is formed. During a second iteration by the computer, a second layer increment 24 of the deposit layer 10 is formed on the outer surface of the first layer increment 24 in contact with the fluid 14. For each subsequent iteration of step 108, an additional layer increment 24 is added to the deposit layer 10. Next, in a step 110, the computer uses a composition of solid species in the coolant fluid 14 to define the composition of solid species in each of the layer increments 24 of the deposit layer 10. The solid species are elements in the form of particulate, colloidal and/or ionic substances in the coolant fluid 14 that are attracted to the heat transfer surface 12 of the nuclear reactor. An increase of temperature difference between the heat transfer surface 12 and the coolant fluid 14 increases the rate of attraction of the solid species to the heat transfer surface 12. The deposition of the species is a function of the heat flux of the heat transfer surface 12, which for instance is dependent on the heat generated by the heated surface, the heat transfer of the deposit layer 10, and the temperature of the coolant fluid 14. In a step 112, the computer uses the temperature and pressure of the layer increments 24 to define a temperature profile of the deposit layer 10 for a full thickness of the deposit layer 10, layer increment 24 by layer increment 24. Each layer of deposition has its own internal steam pressure. The pressure is higher on layers closer to the heated surface, which makes the steam to be evacuated towards water at the surface of the deposition. There, the pressure is approximately equal (slightly higher) to the pressure in the bulk cooling fluid. The temperature profile is calculated inside each layer increment 24 using a given distribution of channels 22 acting as steam chimneys (steam evacuation members) that evacuate heat out of the deposit layer 10 to the coolant fluid 14. The distribution of channels 22 changes layer increment 24 by layer increment 24. For example, the temperature profile may be based on the number of channels 22 in the deposit layer increment 24 having a radius greater than a predefined limit at which a channel 22 acts as a steam chimney, drawing in coolant fluid from capillaries, heating the fluid and outputting it into the coolant fluid 14 as steam. The temperature profile is dependent upon the amount and composition of the solid species forming each layer increment 24 and the heat flux of the heat transfer surface. In a step 114, the computer uses the temperature profile determined in step 112 for each layer increment 24 and a pressure calculated for each layer increment to define the chemical equilibrium conditions of the deposit layer, layer increment by layer increment. During step 114, the computer determines the solubility and diffusivity of the different species in deposit layer 10 and in the solute within the channels 22 of the deposit layer 10. In other words, soluble species concentrations (as hydroxides) in the deposit layer 10, i.e. in each layer increment 24 of the deposit layer 10, and their diffusivity are determined. The diffusivity depends on the soluble species in each layer increment 24 and the temperature profile in each layer increment 24, as determined in step 112. Defining the chemical equilibrium conditions involves calculating the hydroxide forms of the soluble species deposited as insoluble species into the layer increments 24 of deposit layer 10 and the diffusivity coefficients of each liquid hydroxide form. The diffusivity and the solubility of the species in the deposit layer 10 are used to define the chemical equilibrium by determining the conversion of soluble species into insoluble species and the deposition of the insoluble species into the layer increments 24. In a step 116, the computer uses the chemical equilibrium established in step 114 to redefine the distribution of the species in the deposit layer 10 to determine the actual deposition of the soluble species within the deposit layer 10. The amount and composition of the solid species in the layer increments 24 as determined in step 110 and the amount and composition of the insoluble species in the layer increments 24 as determined in step 114 are summed together. The computer calculates the mass balances of elements (e.g., Fe, Zn, Si and Cu) as hydroxides (e.g., in parts per million) in each layer increment 24, based on the capillary movement of the species towards evacuation through the steam and the diffusion to and from neighboring layer increments 24. The deposition of the soluble species transformed to insoluble species due to the localized conditions in the layer increments 24 occurs within the channels 22 acting as chimneys (but not within the channels acting as capillaries). Accordingly, the amount of deposition within each layer increment 24 depends on the number and volume of the channels 22 acting as chimneys. When the channel 22 acting as a chimney becomes too small to allow the coolant fluid 14 to convert to steam, the rate of deposition within the channel 22 slows or stops, causing the channel 22 to act as a capillary. If channel 22 acting as a chimney experiences a reduction in radius so as to have a smaller radius than an adjacent channel 22 acting as capillary, the adjacent channel 22 acting as a capillary may be forced to act as a chimney. Thus, chimneys may be converted to capillaries and capillaries may be converted to chimneys. Ultimately, the computer may calculate the deposition rate of the species, which may include Zn2SiO4, ZnO, CuO and SiO2, in grams/second, to determine the total deposition of species in each layer increment 24 during each iteration. In a step 118, the computer uses the deposition of the species in each layer increment 24 to recalculate the geometry of the deposit layer 10. The volumes of the insoluble species are distributed in the channels 22 of each layer increment 24 to redefine the geometry of the full deposit layer 10. The porosity and the radiuses of the channels 22 acting as chimneys are decreased by the computer. As coolant fluid is converted to steam within a layer increment 24, soluble species which are transformed to insoluble species due to local conditions within each layer increment 24 are left behind in the steam chimneys reducing the diameter of the chimneys and reducing the overall porosity of the layer increment 24. The volume of the combined species deposited in each layer increment 24 contributes to changing the free volume by decreasing the porosity of layer increments 24. In a step 120, based on the recalculated geometry of the deposit layer 10, at least one temperature related variable of the heat transfer surface 12 of the nuclear reactor is determined and compared to a corresponding limit. Increased buildup of the actual deposit on the heat transfer surface affects the ability of the coolant fluid to cool the heat transfer surface. If the temperature related variable of heat transfer surface 12 reaches the predetermined limit, operation of the actual nuclear reactor is altered. The modeling then returns to step 108 and proceeds through the iterative loops of steps 108 to 120 to add another layer increment 24 to the deposit layer 10 and recalculate the properties of the deposit layer 10 and the temperature of the heat transfer surface 12. As used herein, temperature related variable of the heat transfer surface 12 includes the temperature of the heat transfer surface 12 or any variable that is dependent on the temperature and thus may be considered an indirect measure of the temperature of the heat transfer surface 12. For example, the thermal expansion of the cladding material is an indirect measure of the temperature of the heat transfer surface 12 and is a temperature related variable. In order for the nuclear reactor to be safely operated, the temperature of the heat transfer surface 12 is kept below a predetermined value. Once the layer increments 24 of layer 10 deposited on the heat transfer surface 12 cause the temperature of the heat transfer surface 12 to reach the predetermined value (or another temperature related variable reaches the corresponding limit), at a step 122, operation of the nuclear reactor is altered, either automatically by a computer system programmed to monitor and control the nuclear reactor in accordance with the above steps or by an operator of the nuclear reactor. In some instances, altering the operation of the nuclear reactor may include stopping operation of the nuclear reactor and then replacing the heat transfer surface 12. For example, in a nuclear reactor, once the temperature on a hottest point of a hottest nuclear fuel element reaches the predetermined value, the nuclear reactor is stopped, the fuel elements are removed from the nuclear reactor and replacement fuel elements are inserted into the nuclear reactor. The nuclear reactor may be restarted with the replacement fuel elements. In other instances, altering the operation of the nuclear reactor may include operating the nuclear reactor at modified conditions. Once the temperature related variable of the heat transfer surface reaches the predetermined value, the operability of the reactor may be compromised and a region of the boundary of the heat transfer surface may be at a high risk of failure. In such situations, it may be possible to suppress operation of the region of the heat transfer surface having a high risk of failure. For example, in a nuclear reactor, once the temperature related variable of the heat transfer surface reaches the predetermined value, the boundary of the hottest nuclear fuel element may be at a high risk of failure (i.e., the cladding has a high risk of breaking) and the nuclear reactor may be operated in a high risk mode. The hottest nuclear fuel element may then be suppressed by limiting for instance the nuclear reactor power. FIG. 10 schematically shows a nuclear reactor 200 operated in accordance with an embodiment of the present invention. Nuclear reactor 200 includes a plurality of schematically shown fuel elements 202 in its core. A controller 204 is provided for operating nuclear reactor 200 in accordance with a non-transitory computer readable media programmed or structured to define modules having logic for performing the steps described with respect to the method of FIG. 9. The non-transitory computer readable media includes computer executable process steps operable to control controller 204 in accordance with the method described with respect to FIG. 9. Controller 204 may be in wired or wireless communication with a display device 206 and at least one user input device, for example a keyboard 208 and a mouse 210. Display device 206 may also be a touchscreen display that may be used as an additional or alternative user input device. Display device 206 may display graphic user interfaces illustrating the values used in the method to the user and allowing the user to alter the values. A first exemplary graphic user interface is shown in FIG. 11, which displays inputs that a chemistry module may use in steps 110 or 114 to define the chemical equilibrium of the deposit layer 10. As shown in FIG. 11, the inputs used in step 110 or 114 may include the total, dissolved and solid particles of substances in the coolant fluid 14 (e.g., for example iron, zinc, silicon and copper) and the content of the coolant fluid 14 (e.g., H202, H2, O2 and oxide content). The inputs may be automatically generated by the chemistry module or may be input or altered by the user via one or more of the user input devices. A second exemplary graphic user interface is shown in FIG. 12, which displays inputs that a geometry module may use in steps 106 or 118 to define a layer increment. In FIG. 12, chimney means channel 22, Crud means deposit layer 10, Crud increment means deposit layer increment 24. As shown in FIG. 12, the inputs used in step 106 or 118 may include the minimum channel radius for the defined layer increment 24, the radius increment for the defined layer increment 24, the maximum channel radius for the defined layer increment 24, the number of channel members for the defined layer increment 22, the initial porosity for the defined layer increment 22, the maximum deposit layer thickness (i.e., crud thickness) of the modeled deposit, the thickness of each defined layer increment 24 and the initial tortuosity for the defined layer increment 24. The maximum deposit layer thickness is the limit on the thickness of the entire deposit. The operator programs controller 204 to set this value to exceed a maximum estimated thickness of the entire deposit. If the calculated value of the thickness of the entire deposit exceeds the maximum deposit layer thickness, controller 204 will abort the modeling of the deposit layer. The inputs may be automatically generated by the geometry module or may be input or altered by the user via one or more of the user input devices. A third exemplary graphic user interface is shown in FIG. 13, which displays the values for the reactor core parameters (i.e., fuel rod heat flux, coolant fluid temperature, fuel rod temperature, coolant fluid pressure) and the modeling parameters (i.e., time step duration of each iteration and total time steps, which is the amount of days the nuclear fuel will be within the reactor) that may be used in steps 108 to 120. The inputs may be automatically generated or may be input or altered by the user via one or more of the user input devices. Steps 108 to 120 are performed using mathematical representations stored on the computer readable medium. Numerical values used in steps 108 to 120 are defined as constants, lookup tables or curve fits that can avoid the interpolations required for a lookup table. As directed by the computer readable media and/or one of more of the user input devices, display device 206 may display representations of a plurality of real time and past properties of the deposit layer 10, layer increment 24 by layer increment 24. FIGS. 14 to 24 show graphical interfaces that illustrate the evolution of exemplary properties of each layer increment 24 (X-axis) over time (Y-axis). The older layer increments 24 close to the heat transfer surface 12 are shown at the left side of the plots and the newer outer layer increments 24 close to the circulating coolant fluid 14 are shown at the right side of the plots. FIG. 14 shows a plot of the temperature profile of each layer increment 24 over time illustrating that the temperature is higher near the heat transfer surface 12 and increases with time within a layer increment 24; FIG. 15 shows a plot of the conductivity of each layer increment 24 over time, illustrating that because the conductivity is function of a layer increment composition and porosity, the conductivity of the newer (outer) layer increments 24 is lower than that of the older layer increments 24; FIG. 16 shows a plot of the porosity of each layer increment 24 over time, illustrating that the older layer increments 24 are less porous than the newer layer increments 24; FIG. 17 shows a plot of the fractional solid (i.e., the inverse of the porosity) of each layer increment 24 over time; FIG. 18 shows a plot of the evacuated (also called evaporative or convective) heat output of each layer increment 24, which is primarily due to the conversion of water within the larger channels 22 into steam, over time; FIG. 19 shows a plot of the total heat of each layer increment 24 over time (illustrating the quasi static equilibrium condition of the modeled deposit layer 10); FIG. 20 shows a plot of the FeOH2 (soluble iron) concentration of each layer increment 24 over time; FIG. 21 shows a plot of Fe2O3 (deposited iron oxide) concentration of each layer increment 24 over time; FIG. 22 shows a plot of deposited volume of each layer increment 24 over time; and FIGS. 23 and 24 show plots of channel radii over time, each for one different given family of channels 22. The channels 22 of the family in FIG. 24 each have a radius that is much greater than the channels 22 of the family in FIG. 23. The ability of the user to view these properties of the model analysis allows for both easier model validation and for a more complete understanding of the evolution of the deposit layer 10 and the associated heat transfer performance. In the preceding specification, the invention has been described with reference to specific exemplary embodiments and examples thereof. It will, however, be evident that various modifications and changes may be made thereto without departing from the broader spirit and scope of invention as set forth in the claims that follow. The specification and drawings are accordingly to be regarded in an illustrative manner rather than a restrictive sense. |
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description | The present invention relates to a radiation image acquisition system. Conventionally, as described in the following Patent Document 1, there is known a device which irradiates a tabular scintillator with X-rays emitted from an X-ray source and transmitted through an imaging object, detects visible light (scintillation light) generated in the scintillator by solid-state photodetectors laminated on both surfaces of the scintillator, and superimposes image signals output from the respective solid-state photodetectors on each other to acquire a radiation image. In this device, photodetecting elements are coupled to an X-ray incidence surface of the scintillator and its back surface, and the detection efficiency for visible light is enhanced by detecting visible light in each of the photodetecting element on the incidence surface side and the photodetecting element on the back surface side. Also, as described in the following Patent Document 2, there is known a device which, by use of two scintillators overlaid with each other and one detector, detects scintillation light emitted from the scintillator on an incidence surface side by one surface of the detector, and detects scintillation light emitted from the scintillator on the opposite side by the other surface of the detector. In this device, images are formed with two types of different wavelengths on the respective surfaces of the detector. Patent Document 1: Japanese Patent Application Laid-Open No. H07-27866 Patent Document 2: Japanese Translation of International Application No. 2000-510729 Meanwhile, because the X-ray source is a spot light source, it is necessary for the object to be disposed in at least a region that is irradiated with X-rays. For example, when the object to be imaged is large, and a full picture of the object is desired to be captured, it is necessary to dispose the object at a position closer to the scintillator. By bringing the object close to the scintillator, the projection magnification ratio with respect to the scintillator can be lowered, which allows having a full picture of the object within the range of the scintillator. The present inventors have diligently studied a radiation image acquisition system including a first imaging means that condenses and images scintillation light emitted from an X-ray incidence surface of a scintillator and a second imaging means that condenses and images scintillation light emitted from a surface opposite to the incidence surface. In such a radiation image acquisition system, the first imaging means, i.e., the imaging means on the incidence surface side is located on the same side as that of the object with reference to the scintillator. When the object is brought close to the scintillator in order to adjust the magnification ratio as described above, the object sometimes enters the field of view of the imaging means on the incidence surface side. If the object enters the field of view of the imaging means on the incidence surface side, for example, vignetting due to the object is produced in an image. Therefore, a radiation image acquisition system capable of preventing an object from entering the field of view of the imaging means on the incidence surface side while acquiring an image at a desired magnification ratio has been demanded. It is an object of the present invention to provide a radiation image acquisition system capable of preventing an object from entering the field of view of the imaging means on the incidence surface side while acquiring an image at a desired magnification ratio. A radiation image acquisition system of an aspect of the present invention is characterized by including a radiation source emitting radiation toward an object, a holding unit holding the object, a wavelength conversion member generating scintillation light in response to incidence of the radiation emitted from the radiation source and transmitted through the object, a first imaging means condensing and imaging scintillation light emitted from an incidence surface of the radiation of the wavelength conversion member, a second imaging means condensing and imaging scintillation light emitted from a surface opposite to the incidence surface of the wavelength conversion member, a holding unit position adjusting means adjusting the position of the holding unit between the radiation source and the wavelength conversion member, and an imaging position adjusting means adjusting the position of the first imaging means. According to this radiation image acquisition system, scintillation lights emitted from the radiation incidence surface of the wavelength conversion member and its opposite surface are respectively condensed and imaged by the first imaging means and the second imaging means. The first imaging means is an imaging means on the incidence surface side, and the second imaging means is an imaging means on the side opposite to the incidence surface. By adjusting the position of the holding unit between the radiation source and the wavelength conversion member by the holding unit position adjusting means, the object can be brought close to the wavelength conversion member or moved away from the wavelength conversion member. By bringing the object close to the wavelength conversion member, the magnification ratio can be lowered. By moving the object away from the wavelength conversion member and bringing the object close to the radiation source, the magnification ratio can be increased. Here, even when the object is brought close to the wavelength conversion member, by adjusting the position of the first imaging means by the imaging position adjusting means, entry of the object into the field of view of the first imaging means can be prevented. Thus, entry of the object into the field of view of the first imaging means being an imaging means on the incidence surface side is prevented, while an image can be acquired at a desired magnification ratio. The imaging position adjusting means rotates the first imaging means with a point where an optical axis of the first imaging means and the incidence surface of the wavelength conversion member cross each other set as a rotation center. According to this arrangement, even when the position of the first imaging means is adjusted, the optical path length from the wavelength conversion member to the first imaging means does not change. Accordingly, correction to an image becomes easy. The imaging position adjusting means keeps an angle created by the optical axis of the first imaging means and the incidence surface of the wavelength conversion member while rotating the first imaging means and the wavelength conversion member. According to this arrangement, even when the position of first imaging means is adjusted, the angle created by the optical axis of the first imaging means and the incidence surface of the wavelength conversion member is kept fixed, and thus correction to an image becomes even easier. Also, it is not necessary to frequently perform calibration in the first imaging means, so that the convenience is improved. The imaging position adjusting means keeps an angle created by an optical axis of the second imaging means and the opposite surface of the wavelength conversion member while rotating the first imaging means, the wavelength conversion member, and the second imaging means. According to this arrangement, the first imaging means, the wavelength conversion member, and the second imaging means integrally rotate with the point described above set as a rotation center. Accordingly, even when the position of the first imaging means and the second imaging means is adjusted, the relative positional relationship of the first imaging means, the wavelength conversion member, and the second imaging means does not change. Therefore, images for which an inter-image operation is easily performed can be captured. Also, it is not necessary to frequently perform calibration in the second imaging means, so that the convenience is improved. The above-described radiation image acquisition system includes a detecting means detecting whether the object is in a field of view of the first imaging means. According to this arrangement, because whether the object is in the field of view of the first imaging means is detected by the detecting means, the occurrence of “vignetting” in an image can be reliably prevented. The detecting means detects whether the object is in the field of view of the first imaging means based on a first image captured by the first imaging means and a second image captured by the second imaging means. According to this arrangement, whether the object is in the field of view of the first imaging means can be accurately detected. The detecting means detects whether the object is in the field of view of the first imaging means based on a difference in light intensity between the first image and the second image. According to this arrangement, whether the object is in the field of view of the first imaging means can be accurately detected. The detecting means detects whether the object is in the field of view of the first imaging means based on a difference image between the first image and the second image. According to this arrangement, whether the object is in the field of view of the first imaging means can be accurately detected. The detecting means detects whether the object is in the field of view of the first imaging means based on a ratio of brightness between the first image and the second image. According to this arrangement, whether the object is in the field of view of the first imaging means can be accurately detected. The detecting means detects whether the object is in the field of view of the first imaging means based on successive images successively captured by the first imaging means while the holding unit is moved by the holding unit position adjusting means. According to this arrangement, the point in time where the object has slipped out of the field of view of the first imaging means or the point in time where the object has entered the field of view of the first imaging means can be accurately detected. As a result, the inclination angle of the wavelength conversion member with respect to the radiation source can be minimized, so that an image with little perspective is easily acquired. The above-described radiation image acquisition system includes an image operating means performing an image operation of a first image captured by the first imaging means and a second image captured by the second imaging means based on a rotation angle of the first imaging means, the wavelength conversion member, and the second imaging means. According to this arrangement, a CT (Computed Tomography) image of the object can be acquired. According to an aspect of the present invention, entry of an object into the field of view of the first imaging means being an imaging means on the incidence surface side can be prevented, while an image can be acquired at a desired magnification ratio. Hereinafter, an embodiment of the present invention will be described with reference to the drawings. In addition, the same elements will be denoted by the same reference signs in the description of the drawings, and overlapping description will be omitted. Also, the respective drawings are prepared for the purpose of description, and are drawn so that the portions to be described are especially emphasized. Therefore, the dimensional ratios of respective members in the drawings are not always coincident with actual ratios. As shown in FIG. 1 to FIG. 3, a radiation image acquisition system 1 of a first embodiment is a system for acquiring a radiation image of an object A. The radiation image acquisition system 1 includes a radiation source 2 that emits radiation such as white X-rays toward the object A, a wavelength conversion plate (wavelength conversion member) 6 that generates scintillation light in response to incidence of the radiation emitted from the radiation source 2 and transmitted through the object A, a front observation photodetector (first imaging means) 3 that condenses and images scintillation light emitted from a radiation incidence surface 6a of the wavelength conversion plate 6, and a back observation photodetector (second imaging means) 4 that condenses and images scintillation light emitted from a back surface 6b (refer to FIG. 3) that is a surface opposite to the incidence surface 6a. The radiation source 2 emits cone beam X-rays from an X-ray emission spot 2a. The object A is an electronic component such as a semiconductor device, and is, for example, a semiconductor integrated circuit. The object A is not limited to a semiconductor device, and may be food or the like. The object A may even be a film or the like. The radiation image acquisition system 1 acquires a radiation image of the object A for the purpose of, for example, a non-destructive analysis of an industrial product. The wavelength conversion plate 6 is a tabular wavelength conversion member, and is, for example, a scintillator such as Gd2O2S:Tb, Gd2O2S:Pr, CsI:Tl, CdWO4, CaWO4, Gd2SiO5:Ce, Lu0.4Gd1.6SiO5, Bi4Ge3O12, Lu2SiO5:Ce, Y2SiO5, YAlO3:Ce, Y2O2S:Tb, or YTaO4:Tm. The thickness of the wavelength conversion plate 6 is, in a range of several micrometers to several millimeters, set to an appropriate value according to the energy band of detecting radiation. The wavelength conversion plate 6 converts X-rays transmitted through the object A to visible light. X-rays with relatively low energy are converted by the incidence surface 6a that is a front surface of the wavelength conversion plate 6, and is emitted from the incidence surface 6a. Also, X-rays with relatively high energy are converted by the back surface 6b of the wavelength conversion plate 6, and is emitted from the back surface 6b. The front observation photodetector 3 (hereinafter, referred to as a “front surface detector 3”) is an imaging means according to an indirect conversion method that captures a projection image (i.e., a radiation transmission image) of the object A projected on the wavelength conversion plate 6 from the side of the incidence surface 6a of the wavelength conversion plate 6. That is, the front surface detector 3 is an imaging means on the side of the incidence surface 6a. The front surface detector 3 has a condenser lens unit 3a that condenses scintillation light emitted from the incidence surface 6a of the wavelength conversion plate 6, and an imaging unit 3b that images scintillation light condensed by the condenser lens unit 3a. The front surface detector 3 is a lens coupling type detector. The condenser lens unit 3a condenses scintillation light in a field of view 23. As the imaging unit 3b, for example, an area sensor such as a CMOS sensor or a CCD sensor is used. The back observation photodetector 4 (hereinafter, referred to as a “back surface detector 4” is an imaging means according to an indirect conversion method that captures a projection image (i.e., a radiation transmission image) of the object A projected on the wavelength conversion plate 6 from the side of the back surface 6b of the wavelength conversion plate 6. That is, the back surface detector 4 is an imaging means on the side of the back surface 6b. The back surface detector 4 has a condenser lens unit 4a that condenses scintillation light emitted from the back surface 6b of the wavelength conversion plate 6, and an imaging unit 4b that images scintillation light condensed by the condenser lens unit 4a. The back surface detector 4 is a lens coupling type detector, and has the same configuration as that of the front surface detector 3 described above. The condenser lens unit 4a condenses scintillation light in a field of view 24 via a mirror 5. As the imaging unit 4b, for example, an area sensor such as a CMOS sensor or a CCD sensor is used. The mirror 5 reflects light emitted from the back surface 6b of the wavelength conversion plate 6, and directs the reflected light toward the back surface detector 4. Exposure to radiation of the back surface detector 4 can thereby be prevented. As shown in FIG. 3, the radiation image acquisition system 1 includes a timing control unit 27 that controls imaging timing in the front surface detector 3 and the back surface detector 4, an image processing device 28 that is input with image signals output from the front surface detector 3 and the back surface detector 4, and executes a predetermined processing such as an image processing based on the respective input image signals, and a display device 29 that is input with image signals output from the image processing device 28, and displays a radiation image. The timing control unit 27 and the image processing device 28 are constructed by a computer having a CPU (Central Processing Unit), a ROM (Read-Only Memory), a RAM (Random-Access Memory), an input/output interface, etc. As the display device 29, a publicly known display is used. In addition, the timing control unit 27 and the image processing device 28 may be constructed as programs to be executed by a single computer, or may be constructed as units that are separately provided. The image processing device 28 has an image acquisition unit 28a, a detection unit (detecting means) 28b, and an image processing unit (image operating means) 28c. The image acquisition unit 28a is input with image signals output from the front surface detector 3 and the back surface detector 4. The detection unit 28b detects whether the object A is within the field of view 23 of the front surface detector 3 based on a radiation image indicated in the image signals input by the image acquisition unit 28a. The image processing unit 28c executes a predetermined processing such as an inter-image operation including a difference operation and an addition operation based on the image signals input by the image acquisition unit 28a. The image processing unit 28c outputs image signals after the image processing to the display device 29. As shown in FIG. 1 to FIG. 3, the radiation source 2, the front surface detector 3, the back surface detector 4, and the wavelength conversion plate 6 described above are mounted on a plate-like base 10. On one end portion of the base 10, the radiation source 2 is placed, and fixed. The radiation source 2 has an optical axis X parallel to an extending direction of the base 10. On the optical axis X of the radiation source 2, the object A and the wavelength conversion plate 6 are disposed. That is, the object A is disposed between the X-ray emission spot 2a of the radiation source 2 and the wavelength conversion plate 6. The object A is held by a projection angle changing stage (holding unit) 11. The projection angle changing stage 11 is for holding the object A and rotating the object A. Rotating the object A by the projection angle changing stage 11 allows acquiring radiation images with various projection angles. The projection angle changing stage 11 has a drive mechanism (not shown), and rotates the object A about a rotation axis L1 by the drive mechanism. The rotation axis L1 is perpendicular to the extending direction of the base 10. The rotation axis L1 intersects the optical axis X of the radiation source 2, and also passes substantially the center of the object A. In addition, the rotation axis L1 is not limited to the case of passing substantially the center of the object A, and may be located at a position deviated from the object A. Further, the projection angle changing stage 11 is supported by a magnification ratio changing stage (holding unit position adjusting means) 12. The magnification ratio changing stage 12 is for moving the object A along the optical axis of the radiation source 2 between the radiation source 2 and the wavelength conversion plate 6. The magnification ratio changing stage 12 moves the object A to change the distance FOD (Focus-Object Distance) between the radiation source 2 (X-ray focus) and the object A, and thereby adjusts a ratio of FOD to the distance FID (Focus-Image Distance) between the radiation source 2 (X-ray focus) and the wavelength conversion plate 6. The magnification ratio of a radiation image can thereby be changed. The projection angle changing stage 12 is attached to the base 10, and extends parallel to the optical axis X of the radiation source 2. The magnification ratio changing stage 12 has a drive mechanism (not shown), and causes a sliding movement of the projection angle changing stage 11 between the radiation source 2 and the wavelength conversion plate 6 by the drive mechanism. The moving direction of the projection angle changing stage 11 is parallel to the optical axis X of the radiation source 2. To the other end portion of the base 10, a rotating body 20 that is rotatable with respect to the base 10 is attached. The rotating body 20 is supported by a shooting angle changing stage (imaging position adjusting means) 17. The shooting angle changing stage 17 has a drive mechanism 17a, and rotates the rotating body 20 about a rotation axis L2 by the drive mechanism 17a. The rotation axis L2 is parallel to the rotation axis L1. The rotation axis L2 is perpendicular to the extending direction of the base 10. The rotation axis L2 intersects the optical axis X of the radiation source 2, and also passes over the incidence surface 6a of the wavelength conversion plate 6. Also, the rotation axis L2 intersects an optical axis 3c of the front surface detector 3. That is, the shooting angle changing stage 17 rotates the rotating body 20 with a point where the optical axis 3c of the front surface detector 3 and the incidence surface 6a of the wavelength conversion plate 6 cross each other (i.e., a point α to be described later) set as a rotation center. The rotating body 20 has an X-ray protection box 14 that is supported by the shooting angle changing stage 17, a front surface camera mount 13 on which the front surface detector 3 is placed, and an interlocking arm 16 that interlocks the X-ray protection box 14 and the front surface camera mount 13. The X-ray protection box 14 is a casing made of, for example, an X-ray shielding material such as lead, and houses the back surface detector 4. The X-ray protection box 14, by shielding X-rays emitted from the radiation source 2, prevents the back surface detector 4 from being exposed thereto. In a surface of the X-ray protection box 14 opposed to the radiation source 2, a quadrangular opening is formed. The wavelength conversion plate 6 is fitted in the opening to be fixed to the X-ray protection box 14. To the interior of the X-ray protection box 14, the back surface detector 4 and the mirror 5 are fixed. The mirror 5 has a reflecting surface that is perpendicular to the extending direction of the base 10 and creates 45 degrees with respect to the back surface 6b of the wavelength conversion plate 6. The condenser lens unit 4a of the back surface detector 4 is opposed to the mirror 5. The back surface detector 4 has an optical axis 4c that is parallel to the extending direction of the base 10. The optical axis 4c of the back surface detector 4 is parallel to the back surface 6b of the wavelength conversion plate 6. That is, the optical axis 4c is perpendicular to the reflecting surface of the mirror 5. The mirror 5 reflects scintillation light emitted from the back surface 6b of the wavelength conversion plate 6, and directs this light toward the back surface detector 4. In addition, the angles of the mirror 5 and the optical axis 4c with respect to the back surface 6b of the wavelength conversion plate 6 are not limited to the angles described above, and can be appropriately set. It suffices with an arrangement which enables condensing scintillation light emitted from the back surface 6b of the wavelength conversion plate 6 by the back surface detector 4. The interlocking arm 16 extends from the other end portion to the one end portion of the base 10. That is, the interlocking arm 16 extends from near a side of the wavelength conversion plate 6 in the X-ray protection box 14 to a side of the radiation source 2. The interlocking arm 16 is disposed at a position so as not to interfere with the optical axis X of the radiation source 2. On the front surface camera mount 13, the front surface detector 3 is fixed. Accordingly, the front surface detector 3 is disposed lateral to the radiation source 2. In other words, the front surface detector 3 is disposed on the same side as that of the radiation source 2 with reference to a virtual plane that passes the position of the object A and is perpendicular to the optical axis X of the radiation source 2. The condenser lens unit 3a of the front surface detector 3 is opposed to the wavelength conversion plate 6. The optical axis 3c of the front surface detector 3 is parallel to the extending direction of the base 10, and is perpendicular to the incidence surface 6a of the wavelength conversion plate 6. In addition, the angle of the optical axis 3c with respect to the incidence surface 6a of the wavelength conversion plate 6 is not limited to the angle described above, and can be appropriately set. It suffices with an arrangement which enables condensing scintillation light emitted from the incidence surface 6a of the wavelength conversion plate 6 by the front surface detector 3. In addition, a light receiving surface of the imaging unit 3b may be substantially parallel to the incidence surface 6a. Due to the above configuration, the rotating body 20 including the wavelength conversion plate 6, the front surface detector 3, the mirror 5, and the back surface detector 4 is rotatable in an integrated manner centering on the rotation axis L1. That is, the shooting angle changing stage 17 keeps the angle created by the optical axis 3c of the front surface detector 3 and the incidence surface 6a of the wavelength conversion plate 6 at 90 degrees, while rotating the front surface detector 3 and the wavelength conversion plate 6. Further, the shooting angle changing stage 17 keeps the angle created by the optical axis 4c of the back surface detector 4 and the back surface 6b of the wavelength conversion plate 6 at 90 degrees, while rotating the front surface detector 3, the wavelength conversion plate 6, and the back surface detector 4. The shooting angle changing stage 17 changes the angles created by the optical axis 3c of the front surface detector 3 and the optical axis 4c of the back surface detector 4 with respect to the optical axis X of the radiation source 2. With the rotation of the rotating body 20 by the shooting angle changing stage 17, the field of view 23 of the front surface detector 3 and the field of view 24 of the back surface detector 4 also rotate. As above, because the front surface detector 3, the back surface detector 4, and the wavelength conversion plate 6 rotate in an integrated manner, the relative positional relationship of the front surface detector 3, the wavelength conversion plate 6, and the back surface detector 4 does not change. Therefore, images that are acquired by the front surface detector 3 and the back surface detector 4 are images for which an inter-image operation is easily performed in the image processing device 28. Also, because the angles of the front surface detector 3 and the back surface detector 4 with respect to the wavelength conversion plate 6 are also fixed, it is not necessary to frequently perform calibration in the front surface detector 3 and the back surface detector 4, so that the convenience is high. The optical axis X of the radiation source 2 fixed on the base 10 creates an angle θ with respect to a normal B to the incidence surface 6a of the wavelength conversion plate 6. That is, the radiation source 2 faces the object A and the incidence surface 6a, and is disposed at a position off the normal B to the incidence surface 6a. In other words, the optical X of the radiation source 2 creates an acute angle with respect to the incidence surface 6a. The angle θ changes with a rotation of the rotating body 20. Here, the optical axis X of radiation is a straight line connecting the X-ray emission spot 2a of the radiation source 2 and an arbitrary point γ on the incidence surface 6a of the wavelength conversion plate 6. In the present embodiment, the arbitrary point γ is set so as to correspond to a center point of the incidence surface 6a, and in this case, radiation is irradiated relatively evenly. Also, the normal B is a straight line extending from an arbitrary point α on the incidence surface 6a and normal to the incidence surface 6a. In the present embodiment, the arbitrary point α is set so as to correspond to a center point of the incidence surface 6a, and the optical axis X of radiation and the normal B cross each other at the arbitrary point γ (i.e., the arbitrary point α) of the incidence surface 6a. Of course, the arbitrary point γ and the arbitrary point α are not necessarily a center point of the incidence surface 6a, or not necessarily the same point. The optical axis 3c of the condenser lens unit 3a of the front surface detector 3 is coincident with the normal B to the incidence surface 6a. The front surface detector 3 is capable of imaging scintillation light emitted in the direction of normal B to the incidence surface 6a, and thus easily acquires an image with little perspective. The condenser lens unit 3a focuses on the incidence surface 6a, and condenses scintillation light emitted in the direction of normal B from the incidence surface 6a toward the imaging unit 3b. In addition, the optical axis 3c of the front surface detector 3 may not be coincident with the normal B to the incidence surface 6a. In this manner, the front surface detector 3 is disposed off the optical axis X of the radiation source 2. That is, the front surface detector 3 is disposed so as to separate from an emission region of radiation from the radiation source 2 (region where a radiation flux 22 exists). Exposure of the front surface detector 3 to radiation from the radiation source 2 is thereby prevented, which prevents a direct conversion signal of radiation from being generated in the interior of the front surface detector 3 to generate noise. Also, the front surface detector 3 is disposed such that a perpendicular line drawn from the center of the condenser lens unit 3a to the incidence surface 6a of the wavelength conversion plate 6 is within the range of the incidence surface 6a, and is disposed over the incidence surface 6a of the wavelength conversion plate 6. A relatively large amount of scintillation light can thereby be detected. The optical axis 4c of the condenser lens unit 4a of the back surface detector 4 is coincident with a normal C to the back surface 6b via the mirror 5. The back surface detector 4 is capable of imaging scintillation light emitted in the direction of normal C to the back surface 6b, and thus easily acquires an image with little perspective. Here, the normal C is a straight line extending from an arbitrary point β on the back surface 6b and normal to the back surface 6b. Particularly, in the present embodiment, the arbitrary point β is set as a center point of the back surface 6b, the arbitrary point α on the incidence surface 6a and the arbitrary point β on the back surface 6b are located on the same line, and this straight line is coincident with the normal B and the normal C. The condenser lens unit 4a focuses on the back surface 6b, and condenses scintillation light emitted in the direction of normal C from the back surface 6b toward the imaging unit 4b. In addition, the optical axis 4c of the back surface detector 4 may not be coincident with the normal C to the back surface 6b. In the radiation image acquisition system 1, the optical path length from the incidence surface 6a of the wavelength conversion plate 6 to the front surface detector 3 is equal to the optical path length from the back surface 6b of the wavelength conversion plate 6 to the back surface detector 4. In addition, the optical path length from the incidence surface 6a of the wavelength conversion plate 6 to the front surface detector 3 may be different from the optical path length from the back surface 6b of the wavelength conversion plate 6 to the back surface detector 4. In this case, it is necessary to match the image size etc., by an image processing or the like. As in the foregoing, because the front surface detector 3, the back surface detector 4, and the wavelength conversion plate 6 rotate in an integrated manner, each of the optical path length from the incidence surface 6a of the wavelength conversion plate 6 to the front surface detector 3 and the optical path length from the back surface 6b of the scintillator 6 to the back surface detector 4 does not change even by a rotation of the rotating body 20, and is fixed. Accordingly, correction to images acquired by each of the front surface detector 3 and the back surface detector 4 is easy. Subsequently, the operation of the radiation image acquisition system 1 having the configuration described above will then be described. First, control by the timing control unit 27 is performed such that imaging by the front surface detector 3 and imaging by the back surface detector 4 are simultaneously performed. Imaging timing control by the timing control unit 27 allows imaging radiation transmission images of the object A in different energy bands. In detail, a radiation transmission image in a relatively low energy band is imaged by the front surface detector 3, and a radiation transmission image in a relatively high energy band is imaged by the back surface detector 4. Dual-energy imaging is thereby realized. In addition, it is possible in the radiation image acquisition system 1 to control the imaging timings of the front surface detector 3 and the back surface detector 4 so as to be different from each other. Also, the front surface detector 3 and the back surface detector 4 may be controlled so as to be different from each other in the exposure time and number of shots. Regarding the function of the front surface detector 3 and the back surface detector 4, in other words, fluorescence (scintillation light) converted at the side relatively close to the incidence surface 6a is detected by the front surface detector 3. Detection of fluorescence converted at the incidence surface 6a-side has features that the fluorescence has little blur and the brightness of fluorescence is high. This is because, in front observation, the influence of diffusion and self-absorption in the interior of the wavelength conversion plate 6 can be reduced. On the other hand, in the back surface detector 4, fluorescence converted at the side relatively close to the back surface 6b of the wavelength conversion plate 6 is detected. Also in this case, the influence of diffusion and self-absorption in the interior of the wavelength conversion plate 6 can be reduced. Next, image signals corresponding to radiation images of both front and back surfaces are output to the image processing device 28 by each of the front surface detector 3 and the back surface detector 4. When the image signals output from each of the front surface detector 3 and the back surface detector 4 are input to the image acquisition unit 28a of the image processing device 28, a predetermined processing such as an inter-image operation including a difference operation and an addition operation is executed based on the input image signals and image signals after the image processing are output to the display device 29 by the image processing unit 28c of the image processing device 28. Then, when the image signals after the image processing output from the image processing device 28 are input to the display device 29, a radiation image according to the input image signals after the image processing is displayed by the display device 29. Particularly, in the image processing device 28, a three-dimensional image of the object A can also be prepared by rotating the object A by the projection angle changing stage 11. Here, according to the radiation image acquisition system 1 of the present embodiment, an image of the object A can be acquired at a desired magnification ratio, and further, entry of the object A into the field of view 23 of the front surface detector 3 can be prevented. Hereinafter, imaging of the object A by the radiation image acquisition system 1 will be described in greater detail with reference to FIG. 3 to FIGS. 6A-FIG. 6C. As shown in FIG. 3, in a normal shooting state, the object A is disposed within the range of cone beam-shaped X-rays emitted from the radiation source 2 (i.e., within the range of the radiation flux 22). At this time, the front surface detector 3 is disposed such that the field of view 23 of the front surface detector 3 does not include the object A. In this case, as shown in FIG. 6A, the shot image Pa has a projection image P2a reflected in the luminescent part P1 of the wavelength conversion plate 6. As above, when the object A is shot at a certain level of magnification ratio, vignetting due to the object A is not produced. On the other hand, as shown in FIG. 4, when it is desired to change the magnification ratio or the object A cannot be within a cone beam (i.e., within the radiation flux 22) because the sample is large, the object A is moved in a direction to approach the wavelength conversion plate 6 by use of the magnification ratio changing stage 12. At this time, the object A sometimes enters the field of view 23 of the front surface detector 3. In this case, the object A may block light from the wavelength conversion plate 6. Accordingly, as shown in FIG. 6B, the shot image Pb has not only the projection image P2b but also vignetting P3 due to the object A reflected in the luminescent part P1 of the wavelength conversion plate 6. As above, the object A enters the field of view 23 of the front surface detector 3 when the magnification ratio is lowered, so that vignetting is produced. Therefore, as shown in FIG. 5, the X-ray protection box 14 is rotated, by use of the shooting angle changing stage 17, centering on the point α where the optical axis 3c of the front surface detector 3 and the incidence surface 6a of the wavelength conversion plate 6 cross each other (i.e., the rotation axis L2). At this time, with the rotation of the X-ray protection box 14, the front surface detector 3 and the front surface camera mount 13 also rotate with the same rotation center by the same angle through the interlocking arm 16. That is, the rotating body 20 rotates. At this time, because the positional relationship of the front surface detector 3 with the wavelength conversion plate 6 is maintained, it is not necessary to change calibration conditions. As a result of thus moving the front surface detector 3 by rotation, as shown in FIG. 6C, the shot image Pc has the projection image P2c free from vignetting due to the object A reflected in the luminescent part P1 of the wavelength conversion plate 6. As above, by deepening the camera shooting angle of the front surface detector 3, vignetting due to the object A can be eliminated. As above, by rotating the front surface detector 3 centering on the rotation axis L2 by the shooting angle changing stage 17, entry of the object A into the field of view 23 of the front surface detector 3 can be prevented. In the example shown in FIG. 7, the object A is removed from the field of view 23 of the front surface detector 3 by further rotating the front surface detector 3 by only an angle Δθ from an angle θ. In the radiation image acquisition system 1, whether the object A is in the field of view 23 of the front surface detector 3 can be detected by the detection unit 28b of the image processing device 28. The detection unit 28b detects whether the object A is in the field of view 23 of the front surface detector 3 by performing various types of processing to be mentioned below. Specifically, the detection unit 28b can detect whether the object A is in the field of view 23 of the front surface detector 3 based on an incidence surface image captured by the front surface detector 3 and a back surface image captured by the back surface detector 4. The detection unit 28b can also detect whether the object A is in the field of view 23 of the front surface detector 3 based on a difference in light intensity between the incidence surface image and the back surface image. The detection unit 28b can also detect whether the object A is in the field of view 23 of the front surface detector 3 based on a difference image between the incidence surface image and the back surface image. The detection unit 28b can also detect whether the object A is in the field of view 23 of the front surface detector 3 based on a ratio of brightness between the incidence surface image and the back surface image. The detection unit 28b can also detect whether the object A is in the field of view 23 of the front surface detector 3 based on successive images of the incidence surface successively captured by the front surface detector 3 while the projection angle changing stage 11 is moved by the magnification ratio changing stage 12. According to the radiation image acquisition system 1 of the present embodiment described above, scintillation lights emitted from the incidence surface 6a and the back surface 6b of the wavelength conversion plate 6 are respectively condensed and imaged by the front surface detector 3 and the back surface detector 4. By adjusting the position of the projection angle changing stage 11 between the radiation source 2 and the wavelength conversion plate 6 by the magnification ratio changing stage 12, the object A can be brought close to the wavelength conversion plate 6 or moved away from the wavelength conversion plate 6. By bringing the object A close to the wavelength conversion plate 6, the magnification ratio can be lowered. By moving the object A away from the wavelength conversion plate 6 and bringing the object A close to the radiation source 2, the magnification ratio can be increased. Here, even when the object A is brought close to the wavelength conversion plate 6, by adjusting the position of the front surface detector 3 by the shooting angle changing stage 17, entry of the object A into the field of view 23 of the front surface detector 3 can be prevented. Thus, entry of the object A into the field of view 23 of the front surface detector 3 being an imaging means on the incidence surface side can be prevented, while an image can be acquired at a desired magnification ratio. Also, the occurrence of vignetting due to the object A can be prevented. Because the shooting angle changing stage 17 rotates the front surface detector 3 with the point α where the optical axis 3c of the front surface detector 3 and the incidence surface 6a of the wavelength conversion plate 6 cross each other set as a rotation center, even when the position of the front surface detector 3 is adjusted, the optical path length from the wavelength conversion plate 6 to the front surface detector 3 does not change. Accordingly, correction to an image is easy. Even when the position of the front surface detector 3 is adjusted, the angle created by the optical axis 3c of the front surface detector 3 and the incidence surface 6a of the wavelength conversion plate 6 is kept fixed, and thus correction to an image becomes even easier. Also, it is not necessary to frequently perform calibration in the front surface detector 3, so that the convenience is improved. The front surface detector 3, the wavelength conversion plate 6, and the back surface detector 4 integrally rotate with the point α described above set as a rotation center. Accordingly, even when the position of the front surface detector 3 and the back surface detector 4 is adjusted, the relative positional relationship of the front surface detector 3, the wavelength conversion plate 6, and the back surface detector 4 does not change. Therefore, images for which an inter-image operation is easily performed can be captured. Also, it is not necessary to frequently perform calibration in the back surface detector 4, so that the convenience is improved. Conventionally, when the object A is large-sized or has a low magnification ratio (i.e., the object A is close to the wavelength conversion plate 6), the object A overlaps the field of view 23 of the front surface detector 3, and the shootable area has consequently been limited. According to the radiation image acquisition system 1, the shootable area can be widened by widening the angle range in which the optical axis 3c can be moved. By making the angle created by the optical axis X of the radiation source 2 and the optical axis 3c of the front surface detector 3 to a minimum when the object A is small, the influence of “vignetting” due to an inclination of the wavelength conversion plate 6 can be reduced, and a loss or decline in resolution can be reduced as much as possible. Because whether the object A is in the field of view 23 of the front surface detector 3 is detected by the detection unit 28b, the occurrence of vignetting in an image can be reliably prevented. The detection unit 28b detects whether the object A is in the field of view 23 of the front surface detector 3 based on an incidence surface image captured by the front surface detector 3 and a back surface image captured by the back surface detector 4. This allows accurately detecting whether the object A is in the field of view 23. The detection unit 28b can also detect whether the object A is in the field of view 23 of the front surface detector 3 based on a difference in light intensity between the incidence surface image and the back surface image. This allows accurately detecting whether the object A is in the field of view 23. The detection unit 28b can also detect whether the object A is in the field of view 23 of the front surface detector 3 based on a difference image between the incidence surface image and the back surface image. This allows accurately detecting whether the object A is in the field of view 23. The detection unit 28b can also detect whether the object A is in the field of view 23 of the front surface detector 3 based on successive images of the incidence surface successively captured by the front surface detector 3 while the projection angle changing stage 11 is moved by the magnification ratio changing stage 12. This allows accurately detecting the point in time where the object A has slipped out of the field of view 23 of the front surface detector 3 or the point in time where the object A has entered the field of view 23 of the front surface detector 3. As a result, the inclination angle of the wavelength conversion plate 6 with respect to the radiation source 2 can be minimized, so that an image with little perspective is easily acquired. Meanwhile, when the radiation image acquisition system 1 is an X-ray CT system, information on the angles with respect to the optical axis X of the radiation source 2 and the incidence surface 6a of the wavelength conversion plate 6 becomes necessary. In the radiation image acquisition system 1, because the angle of the incidence surface 6a of the wavelength conversion plate 6 and the optical axis 3c of the front surface detector 3 is kept fixed, by determining the angle of the optical axis X of the radiation source 2 and the optical axis 3c of the front surface detector 3, a CT image can be acquired. Specifically, as shown in FIG. 8, the angle can be changed by driving the drive mechanism 17a that is a rotation actuator. In this case, a shot image is checked (visual or algorithmic detection is performed), and the angle of the front surface detector 3 is changed to reach a position where the object A is not reflected in the image. Then, the angle at that time is detected. Changing the angle of the wavelength conversion plate 6 and the front surface detector 3 by the drive mechanism 17a allows obtaining the changed angle by, for example, a PC 30 connected to the drive mechanism 17a. The angle between the optical axis X and the optical axis 3c can thereby be obtained. Also, as shown in FIG. 9, the angle can also be manually changed. In this case, by disposing the wavelength conversion plate 6 and the front surface detector 3 on a graduated rotating stage (imaging position adjusting means) 17A and reading the graduation when the angle was manually changed, the angle between the optical axis X and the optical axis 3c can be obtained. In these cases, the image processing unit 28c of the image processing device 28 can perform an image operation of the incidence surface image and the back surface image based on a rotation angle of the front surface detector 3, the wavelength conversion plate 6, and the back surface detector 4. According to the image processing device 28 including the image processing unit 28c, a CT image of the object A can be acquired. Next, a radiation image acquisition system 1A of a second embodiment will be described with reference to FIG. 10 to FIG. 12. The difference in the radiation image acquisition system 1A shown in FIG. 10 to FIG. 12 from the radiation image acquisition system 1 of the first embodiment is the point of adopting a rotating body 20A having a vertical X-ray protection box 14A in place of the rotating body 20 having the horizontal X-ray protection box 14. In the vertical X-ray protection box 14A, the disposition itself of the wavelength conversion plate 6 has not been changed from that of the radiation image acquisition system 1, but the disposition of a mirror 5A and a back surface detector 4A has been changed. That is, the optical axis 4c of the back surface detector 4A is perpendicular to the extending direction of the base 10. The mirror 5A has a reflecting surface that is inclined at 45 degrees with respect to the extending direction of the base 10. In addition, similar to the radiation image acquisition system 1, the radiation image acquisition system 1A also includes a timing control unit 27, an image processing device 28, and a display device 29. Also according to such a radiation image acquisition system 1A, the position of the front surface detector 3 can be adjusted by rotating the rotating body 20A centering on the rotation axis L2. Thus, the same advantageous effects as those of the radiation image acquisition system 1 can be provided. As above, the embodiment of the present invention has been described, but the present invention is not limited to the above-described embodiment. In the above-described embodiment, a description has been given of the case where the front surface detector 3 rotates centering on the rotation axis L2, but the front surface detector 3 may rotate centering on another rotation axis. The other rotation axis may pass an intersection of the optical axis 3c of the front surface detector 3 and the incidence surface 6a of the wavelength conversion plate 6, but may not pass the intersection. The movement of the front surface detector 3 is not limited to a rotational movement, and may be a sliding movement. Whether the object A is in the field of view 23 of the front surface detector 3 may be detected by other means. For example, a dedicated detector may be used separately. According to an aspect of the present invention, entry of an object into the field of view of the first imaging means being an imaging means on the incidence surface side can be prevented, while an image can be acquired at a desired magnification ratio. 1, 1A . . . radiation image acquisition system, 2 . . . radiation source (radiation source), 3 . . . front observation photodetector (first imaging means), 3c . . . optical axis, 4 . . . back observation photodetector (second imaging means), 4c . . . optical axis, 6 . . . wavelength conversion plate (wavelength conversion member), 6a . . . incidence surface, 6b . . . back surface (surface on the opposite side), 11 . . . projection angle changing stage (holding unit), 12 . . . magnification ratio changing stage (holding unit position adjusting means), 17 . . . shooting angle changing stage (imaging position adjusting means), 23 . . . field of view (field of view of first imaging means), 28b . . . detection unit (detecting means), 28c . . . image processing unit (image operating means), A . . . object, α . . . point. |
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claims | 1. An X-ray optical element for collimating an X-ray beam, the element comprising:a Soller slit having an axis defined by a plurality of lamellas, said lamellas collimating the X-ray beam with respect to a direction of said axis; anda collimator for delimiting the X-ray beam, said collimator being rigidly connected to said Soller slit during operation of the optical element, wherein the X-ray beam delimited by said collimator intersects said axis of said Soller slit within said Soller slit, a direction of the X-ray beam thereby subtending an angle α≧10° with respect to said axis of said Soller slit. 2. The X-ray optical element of claim 1, wherein said Soller slit is a linear Soller slit. 3. The X-ray optical element of claim 1, wherein said Soller slit is a radial Soller slit. 4. The X-ray optical element of claim 2, wherein said lamellas of said linear Soller slit are disposed parallel to a direction of the X-ray beam delimited by said collimator. 5. The X-ray optical element of claim 1, wherein said Soller slit has a recess perpendicular to said Soller slit axis. 6. The X-ray optical element of claim 1, wherein said Soller slit comprises two partial slits, wherein said collimator is at least partially disposed between said two partial slits. 7. The X-ray optical element of claim 1, wherein said collimator has at least two collimator jaws, said collimator jaws being disposed on different sides of said Soller slit. 8. The X-ray optical element of claim 7, wherein said collimator jaws subtend an angle with respect to said axis of said Soller slit which differs from 90° or an angle of 45°. 9. The X-ray optical element of claim 1, wherein said collimator is disposed on one side of said Soller slit. 10. The X-ray optical element of claim 9, wherein said collimator is made in one piece. 11. The X-ray optical element of claim 1, wherein said collimator is made from tantalum. 12. The X-ray optical element of claim 1, wherein a geometry of said collimator or of a collimator opening in said collimator can be adjusted in a non-operating state. 13. The X-ray optical element of claim 1, wherein said collimator is a further linear Soller slit. 14. The X-ray optical element of claim 13, wherein said Soller slit is a linear Soller slit, said linear Soller slit and said further linear Soller slid having different divergence angles. 15. The X-ray optical element of claim 1, wherein said collimator is a further radial Soller slit. 16. The X-ray optical element of claim 15, wherein said Soller slit is a radial Soller slit, said radial Soller slit and said further radial Soller slit having different opening angles and/or different divergence angles. 17. A diffractometer having a source for generating a primary beam, a sample holder for arranging a sample, a detector for detecting a secondary beam emitted by the sample, and the X-ray optical element of claim 1. 18. The diffractometer of claim 17, wherein the X-ray optical element is installed in the diffractometer in such a fashion that it can be rotated about an axis of rotation which is perpendicular to said axis of said Soller slit. 19. The diffractometer of claim 18, further comprising a motor for rotating the X-ray optical element. 20. The diffractometer of claim 18, further comprising automatic control or computer control of rotation of the X-ray optical element. 21. The diffractometer of claim 17, wherein the X-ray optical element is disposed on a side of the primary beam. 22. The diffractometer of claim 17, wherein the X-ray optical element is disposed on a side of the secondary beam. 23. The diffractometer of claim 22, wherein said detector is disposed in a point of intersection of the lamella directions of at least one radial Soller slit. 24. The diffractometer of claim 21, wherein said sample holder is disposed in a point of intersection of lamella directions of at least one radial Soller slit. 25. The diffractometer of claim 21, wherein said source is disposed in a center of at least one radial Soller slit. |
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046631186 | claims | 1. In a nuclear fuel assembly including a plurality of fuel rods supported in spaced array upon a lower tie plate, a nozzle adjacent said lower tie plate for receiving a flow of coolant, a tubular flow channel surrounding said array for directing said coolant through said array of said fuel rods, the lower end of said flow channel surrounding the sides of said nozzle, said nozzle being formed of a material having a greater thermal coefficient of expansion than the material of said channel, attachment means for affixing said lower end of said channel to said sides of said nozzle comprising: a plurality of grooves having tapered sides formed in said sides of said nozzle around the periphery thereof; a like plurality of attachment bars formed of a material having a thermal coefficient of expansion similar to that of said channel, each of said bars being fitted in a respective one of said grooves of said nozzle and being formed with tapered sides mated with the tapered sides of the respective groove; and means for securing said bars to the inside surfaces of the lower end of said channel around the periphery thereof, the angle of the mating taper between said attachment bars and said grooves being selected such that the fit between said bars and said grooves is maintained with changes in temperature without substantial stress of said lower end of said channel. 2. The channel-to-nozzle attachment of claim 1 including clearance between the faces of said bars and the bottoms of said grooves at a predetermined temperature for allowing said bars to move further into said grooves as temperature increases. 3. The channel-to-nozzle attachment of claim 1 wherein said angle of the mating taper A is determined according to the relationship: ##EQU5## where: W is half the width of said grooves D is half the width of said nozzle. 4. The channel-to-nozzle attachment of claim 1 wherein said nozzle has opposite sides and wherein the upper tapered surface of the groove on one side is in the same plane as the lower tapered surface of the groove on the opposite side. 5. The combination of claim 1 wherein said nozzle is formed of stainless steel and said channel and said attachment bars are formed of an alloy of zirconium. 6. In a nuclear fuel assembly including a plurality of fuel rods supported in spaced array upon a lower tie plate, a nozzle adjacent said lower tie plate for receiving a flow of coolant, a tubular flow channel surrounding said array for directing said coolant through said array of said fuel rods, the lower end of said flow channel surrounding the sides of said nozzle, said nozzle being formed of a material having a different thermal coefficient of expansion than the material of said channel, attachment means for affixing said lower end of said channel to said sides of said nozzle comprising: a plurality of grooves having tapered sides formed in said sides of said nozzle around the periphery thereof; a like plurality of attachment bars formed of a material having a thermal coefficient of expansion similar to that of said channel, each of said bars being fitted in a respective one of said grooves of said nozzle and being formed with tapered sides mated with the tapered sides of the respective groove; and means for securing said bars to the inside surfaces of the lower end of said channel around the periphery thereof, the angle of the mating taper between said attachment bars and said grooves being selected such that the fit between said bars and said grooves is maintained with changes in temperature without substantial stress of said lower end of said channel. 7. In a nuclear fuel assembly including a plurality of fuel rods supported in spaced array between a lower tie plate and an upper tie plate, a nozzle adjacent said lower tie plate for receiving a flow of coolant, an open-ended tubular flow channel surrounding said array, said channel being formed of a material having a different thermal coefficient of expansion than the material of said nozzle, and attachment means for affixing the lower end of said channel to said nozzle comprising: a plurality of tapered grooves formed in sides of said nozzle around the periphery thereof adjacent the lower end of said channel; a like plurality of similarly tapered attachment bars positioned in said grooves, said attachment bars being formed of a material having a thermal coefficient of expansion similar to that of the material of said channel; and means for securing said bars to the lower, inner end of said channel around the periphery thereof, the angle of taper of said grooves and said bars being selected such as to allow said bars to move more or less deeply into said grooves with changes in temperature without bending of the lower end of said channel. |
description | The present invention relates to a technology for suppressing the deposition of a radioactive substance onto a metallic material surface with which a coolant containing a radioactive substance comes in contact, and specifically, relates to a method of and system for suppressing the deposition of the radioactive substance after a chemical decontamination operation. In a light water reactor (LWR) using water as a coolant, measures for the reduction of a radiation exposure dose for workers in periodic inspection works (periodic inspection), preventive maintenance works and the like have become important. As a part of such measures, a chemical decontamination operation has been frequently applied to structural materials, pipes, pumps and the like of the reactor. The chemical decontamination can remove an oxide film on the metallic material surface (hereafter referred to as structural material surface) with which a coolant containing a radioactive substance comes in contact, such as a structural material, and as a result, can remove radioactive substances such as cobalt-60 and cobalt-58 in a crud or in the oxide film on the metallic material surface, by combining reductive dissolution, oxidative dissolution and the like with the use of chemicals. However, when the reactor is restarted after the decontamination operation, the radioactive substance will again deposit on the structural material surface. The deposition of the radioactive substance occurs together with the formation of the oxide film. Because the film grows at a high speed on a bared surface of metal after the decontamination operation, a radioactive substance is quickly deposited on that surface, and as a result, a dosage rate of the structural material surface rises again in a short time after the decontamination operation. In order to solve this problem, there are proposed a method of forming an iron oxide film by bringing high-temperature water, steam, oxygen or ozone into contact with the surface of the material after the decontamination operation (Patent Documents 1 and 2), and a method of forming an iron oxide film by bringing a chemical agent containing an iron ion into contact with the surface of the material after the decontamination operation (Patent Documents 2 and 3). In addition, as a general way to suppress the deposition of radioactivity, there is proposed a method of forming a film that resists capturing the radioactivity by injecting zinc or the like during the operation of the reactor (Patent Document 4). Patent Document 1: Japanese Patent Laid-Open Publication No. 2004-294393 Patent Document 2: Japanese Patent Laid-Open Publication No. 2002-236191 Patent Document 3: Japanese Patent Laid-Open Publication No. 2006-38483 Patent Document 4: Japanese Patent Laid-Open Publication No. 8-292290 By the way, a conventional technology of previously forming the iron oxide film on the structural material surface can suppress an amount of the radioactive substance captured during the initial corrosion of the material. However, this conventional technology does not suppress the capturing action itself. Accordingly, the amount of the captured radioactive substance increases with time. In addition, in the technique of forming the film during the operation of the reactor, it is necessary to always control the operating state during the operation, resulting in a high work load for workers. In addition, in order to form a film other than the iron oxide film on the structural material surface, it is necessary to inject a chemical liquid at a high temperature of 100° C. or higher, or a high-concentration solution which contains approximately 10% of the chemical agent for electroless plating or the like. It is also difficult to form the film under an atmospheric pressure condition of the reactor when the reactor is out of service, and there are various problems including a concern that a residual reagent affects the material of the reactor due to the use of the high-concentration solution. The present invention has been achieved for solving the defects in the conventional technology mentioned above, and an object is to provide a method of and system for efficiently suppressing the deposition of a radioactive substance onto the metallic material surface after the start-up of the reactor, by applying film-forming treatment with the use of a low-concentration chemical agent to the metallic material surface with which the coolant containing the radioactive substance comes in contact, such as a structural material of which the amount of the radioactivity has been reduced by a chemical decontamination operation, at low temperature and the atmospheric pressure while the operation of the reactor is stopped for a periodic inspection or the like. In order to solve the above described problems, the present invention provides a method of suppressing deposition of a radioactive substance comprising the steps of: removing an oxide film on a metallic material surface with which a coolant containing the radioactive substance comes in contact; and depositing a titanium oxide on the metallic material surface after the removal of the oxide film. According to the present invention, there is provided a method of effectively suppressing the deposition of a radioactive substance onto a metallic material surface with which a coolant containing the radioactive substance comes in contact, by conducting recontamination-suppressing treatment after a chemical decontamination operation. Hereunder, embodiments of a method of suppressing the deposition of a radioactive substance according to the present invention will be described with reference to the accompanying drawings. A first embodiment according to the present invention will be described below with reference to FIGS. 1 to 4. FIG. 1 is a flow chart illustrating a process of suppressing the redeposition of a radioactive substance, the process including a chemical decontamination step for removing an oxide film on a metallic material surface (hereinafter referred to as structural material surface of reactor) with which a coolant containing a radioactive substance comes in contact, such as on a structural material surface of a reactor, and including a recontamination-suppressing treatment step. In the chemical decontamination steps (S1 to S5) performed during a period in which the operation of the reactor is stopped before the restart after a periodic inspection, a stop or the like operation, an iron oxide is reduced and dissolved so as to consequently remove the contamination by reduction, by injecting a reducing agent such as oxalic acid into a pipe in order to reduce and dissolve the iron oxide in the upper layer of a portion that may become contaminated such as a pipe surface (S1); and subsequently, the reducing agent is decomposed (S2). Next, a chromium oxide is dissolved in an oxidation step with the use of an oxidizing agent (S3); and a reducing agent is injected into the pipe (S4). According to these steps, the oxidizing agent is decomposed by the excessive reducing agent. Next, the reducing agent is decomposed after the reduction treatment (S5). The oxidation step, the reduction step and the reducing-agent decomposition step (S3 to S5) are further repeated, and the oxide film on the inner surface of the pipe is removed. After the above described chemical decontamination step, a titanium oxide is deposited onto the portion that may become contaminated, which has been chemically decontaminated, during the period in which the operation of the reactor is stopped before a rated operation (S6), in order to suppress recontamination; and thereafter, a titanium-oxide waste liquid containing a residual reducing agent is purified in the final purification step (S7). FIG. 2 illustrates the state of the surface of a reactor pipe which has been subjected to the operations such as mentioned above. The surface of the pipe 1 is in such a state that the whole amount of or a part of the oxide film 2 is removed by the decontamination operation and a titanium oxide 3 deposits thereon. After the titanium oxide 3 has been deposited, the pipe 1 is subjected to the operation. When the operation (rated operation) is restarted, cobalt 60, which is a radioactive substance 4, is included in a cooling water 5 of the reactor, and when the high-temperature cooling water 5 of the reactor comes in contact with the pipe 1, the oxide film is again progressively formed on the pipe 1. However, by making the titanium oxide 3 deposit on the pipe surface, the formation of the oxide film 2 is suppressed, and as a result, the deposition of cobalt 60 to the pipe surface can be suppressed. (Confirmation Test) FIG. 3 and FIG. 4 are graphs representing results of a radioactivity deposition test of making titanium dioxide, which is a titanium oxide, deposited on the metal after the chemical decontamination operation. Two types of test pieces were prepared. A first was a test piece from which approximately half of an oxide film was removed. A second was a test piece from which almost all of the oxide film was removed (stainless steel: SUS316L based on Japanese Industrial Standard (JIS)). In the test pieces an oxide film was formed under the condition of the reactor water, and by repeatedly oxidizing the test pieces by ozone and reducing the resultant test pieces by oxalic acid, and titanium dioxide was deposited on these test pieces. Thereafter, these test pieces were immersed in a water of 280° C. for 500 hours, which included cobalt 60. FIG. 3 shows an amount of deposited radioactivity of each test piece. It is understood from FIG. 3 that the amount of deposited radioactivity of the test piece having the titanium dioxide deposited thereon decreases to ⅔ to ½ of the test piece having no titanium dioxide deposited thereon. FIG. 4 also shows a removed amount of the oxide film formed on the metallic test piece which has been tested. It is understood from the results shown in FIG. 4 that the amount of the produced oxide film of the test piece having the titanium dioxide deposited thereon is small in comparison with that of the test piece having no titanium dioxide deposited thereon, regardless of a residual rate of the oxide film after the decontamination operation. It is also understood that the production of the oxide film on the metallic surface is suppressed by the deposition of titanium dioxide, and as a result, the deposition of radioactivity is suppressed, with the result that an effect of reducing the amount of the produced oxide film substantially coincides with the effect of reducing the deposition of radioactivity. In addition, a period necessary for a final purification step among the whole steps (S1 to S7) of decontamination is relatively long, and approximately one to two days are required in the case in which the primary system of a boiling-water reactor is decontaminated. Accordingly, if the titanium dioxide treatment is conducted after the final purification step, it becomes necessary to purify the system again, which results in extremely extending a time period for the whole process. However, in the present invention, the period necessary for the recontamination-suppressing treatment step after the chemical decontamination operation is shortened to the minimum by conducting titanium-dioxide deposition treatment before the final purification step. According to the described first embodiment, there can be provided a method for suppressing the deposition of the radioactive substance, which can efficiently suppress the deposition of the radioactive substance onto the structural material of a nuclear power plant by conducting a recontamination-suppressing treatment at a low temperature and under the atmospheric pressure after the chemical decontamination operation, while the operation of the reactor is stopped at the time of a periodic inspection or the like. Hereunder, a second embodiment according to the present invention will be described with reference to FIG. 5 to FIG. 7. FIG. 5 is a flow chart representing a process of suppressing the redeposition of a radioactive substance, which shows a chemical decontamination step for removing an oxide film formed on a structural material surface of a reactor and a recontamination-suppressing treatment step. In FIG. 5, in steps S1 to S5, the chemical decontamination operation of repeating the oxidation step and the reduction step is conducted in a manner similar to that in the first embodiment. After the reducing-agent decomposition step (S5), an aggregation and deposition step (S7) of a titanium oxide is conducted as recontamination-suppressing treatment. Thereafter, the titanium-oxide waste liquid containing the residual reducing agent is purified in a final purification step (S8). A titanium oxide solution contains titanium oxide particles and a dispersing agent. The dispersing agent has a function of dispersing the titanium oxide particles in a solution. The process according to the second embodiment realizes enhancement of the adhesion performance of the titanium oxide particles to a surface of a metal by converting the titanium oxide particles in the solution into an aggregated state by a heat treatment, an addition of an electrolyte or the like. However, when the aggregation of the titanium oxide particles excessively proceeds, the titanium oxide sediments and the injection itself may become difficult, and it is accordingly desirable to conduct the deposition treatment by using the solution in the state of keeping some extent of dispersibility even though aggregation has been caused therein. FIG. 6 is a flow chart representing an aggregation treatment and deposition step of a titanium oxide (S7). The titanium oxide particles in the titanium oxide solution having the titanium oxide particles dispersed and stabilized therein (S7-1) are made to start to aggregate by heating and/or addition of an electrolytic chemical agent (S7-2). The electrolytic chemical agent added to the solution at this time may be oxalic acid which is used as a reductive decontamination agent. In addition, in order to suppress excessive aggregation, the temperature of the chemical liquid may be lowered after the start of the aggregation treatment. Next, the titanium oxide is deposited onto a surface of the metal by supplying a solution in which the aggregation has started to a reactor site to be subjected to the recontamination-suppressing treatment through the decontamination treatment system illustrated in FIG. 9, and by bringing the aggregation solution into contact with the surface of the metal (S7-3). The state of the surface of a reactor pipe after having been subjected to such treatment becomes the state illustrated in FIG. 2 as in the first embodiment. The pipe 1 is subjected to the operation after the titanium oxide 3 has been deposited thereon. When the operation is restarted, cobalt 60 which is a radioactive substance 4 is contained in cooling water 5 of the reactor, and when the high-temperature cooling water 5 of the reactor comes in contact with the pipe 1, the oxide film is progressively formed on the pipe 1 again. However, by making the titanium oxide 3 deposit onto the pipe surface, the formation of the oxide film 2 is suppressed, and as a result, the deposition of cobalt 60 onto the pipe surface can be suppressed. (Confirmation Test) FIG. 7 represents the result of cohesively depositing the titanium dioxide of a titanium oxide on the metal after the chemical decontamination operation and conducting a deposition test of radioactivity to the resultant metal. FIG. 7 represents a graph showing the amount of the titanium dioxide deposited onto the test piece after a surface of the test piece SUS316L has been brought into contact with a titanium dioxide solution which has been subjected to the aggregation treatment by heating to 90° C., at a fixed flow rate for a fixed period of time. It is understood from FIG. 7 that the adhesivity of the titanium dioxide in the case in which the aggregation treatment has been conducted is enhanced in comparison with the case in which no aggregation treatment has been conducted, regardless of whether the concentration of the titanium dioxide is high or low. As described above, according to the present second embodiment, by conducting recontamination-suppressing treatment with the use of the titanium oxide solution subjected to the aggregation treatment after the chemical decontamination operation, it becomes possible to provide a method of suppressing the redeposition of a radioactive substance, whereby the deposition of the radioactive substance onto the structural material surface of the reactor can be further efficiently suppressed. Hereunder, a third embodiment according to the present invention is described with reference to FIG. 8. The third embodiment is characterized by controlling the residual concentration of the reducing agent after the reducing-agent decomposition step (S5) and the concentration of a dispersing agent contained in the titanium oxide solution, in the titanium-oxide deposition step (S6) in the above described first embodiment. As described above, the titanium oxide solution contains the dispersing agent for dispersing the titanium oxide particles. The dispersing agent absorbs to the particle surface of the titanium oxide, electrifies the particles, and makes the particles in a state of being dispersed by the electrical repulsion. When an electrolyte such as oxalic acid which is used as a reductive decontamination agent is added to the solution, an electric charge of the dispersing agent is neutralized, and an electrified layer is lost. Then, it becomes impossible for the dispersing agent to keep a dispersion state, and particles aggregate. Thus, the oxalic acid which is the reductive decontamination agent has an effect of neutralizing the electric charge of the dispersing agent because the oxalic acid is an electrolyte, and can provides this aggregation effect by being left in the solution at the end of the decomposition step. However, if the organic ions are present in excess, the positive charge of the dispersant adhering to the titanium oxide particle surface is neutralized. Further, since the organic ions itself are attached to the titanium oxide particle surfaces, the titanium oxide particles are dispersed by being negatively charged. For this reason, it is desirable to control a concentration of the electrolyte to the preferred concentration condition for the aggregation of the particles. In consideration of the above fact, the inventors of the subject application have newly found that the optimum aggregation state can be obtained by controlling a ratio of the concentration by normality (N) of a residual reducing agent such as oxalic acid with respect to that of the dispersing agent contained in the titanium oxide solution, to about 1 or less. Accordingly, in the present third embodiment, the ratio of the concentration by normality (N) of the residual reducing agent after a reducing-agent decomposition step to that of the dispersing agent contained in the titanium oxide solution is controlled to about 1 or less. Titanium-oxide deposition treatment (S6) is conducted, and the solution is subjected to a final purification step (S8) which includes the purification of the residual reducing agent. (Confirmation Test) FIG. 8 is a view representing an amount of titanium dioxide deposited onto the test piece when the titanium dioxide solution has been brought into contact with the test piece SUS316L, at a fixed flow-rate for a fixed period of time, while a ratio of the normality of the dispersing agent contained in the titanium dioxide solution to the normality of the oxalic acid to be added has been varied. As is illustrated in FIG. 8, it is understood that the deposition amount increases when the normality ratio of an oxalic acid/dispersing agent is about 1. The case in which the ratio of the normality of the oxalic acid to the normality of the dispersing agent is 1 specifically means a neutral point at which electric charges are neutralized. Accordingly, the adhesivity of the titanium dioxide to the metal is enhanced by controlling a ratio of the concentration by the normality of the residual reducing agent to that of the dispersing agent so as to be about 1 in a reducing-agent decomposition step after the decontamination operation. However, when the concentration of the dispersing agent in an undiluted titanium dioxide solution is originally low and the aggregation easily proceeds only by heating, an adsorption effect of an oxalate ion more remarkably appears than the charge neutralization effect, and thus, the concentration condition of oxalic acid is preferably as low as 1 or less by the normality ratio. As described above, according to the third embodiment, it becomes possible to provide a method of suppressing the deposition of a radioactive substance, which can effectively suppress the deposition of cobalt onto a pipe surface of a reactor now in operating, because a titanium oxide can be more efficiently deposited on a metal even by using a chemical liquid with low concentration, even at low temperature and even under the atmospheric pressure, by appropriately controlling the concentration of a residual reducing agent after a reducing-agent decomposition step with respect to the concentration of the dispersing agent in the titanium oxide solution. FIG. 9 is a schematic diagram illustrating a whole structure of a recontamination-suppressing treatment system according to the fourth embodiment of the present invention. The system includes a redeposition-suppressing apparatus for implementing the above described method for suppressing the deposition of a radioactive substance according to the present invention, the system being applied, for example, to the recirculation system of the reactor coolant in a boiling-water type reactor, in which the recontamination-suppressing treatment by the deposition of a titanium oxide is conducted after the decontamination of the recirculation system of a reactor coolant. With reference to FIG. 9, a recirculation system of a reactor coolant includes a recirculation pump 25, and a riser tube nozzle 26 of a jet pump. On the other hand, a decontamination treatment system includes a main circulation line 7 for decontamination for decontaminating an area 6 to be decontaminated in the recirculation system of the reactor coolant, a circulation pump 8 for decontamination, a chemical liquid tank 9 for preparing a chemical liquid therein, a heater 10, an ozone generator 11 to be used as an oxidation decontamination agent, a mixer 12 for dissolving an ozone gas into the liquid, an ion-exchange resin column 13, and a aggregation tank 14 for accommodating a titanium oxide solution therein. The aggregation tank 14 is provided with a circulating pump 15, a stirrer 16, a heater 17 for heating, and a turbidity sensor 18 for monitoring the state of aggregation of titanium dioxide. Incidentally, the stirrer 16 is not necessarily indispensable and may also be appropriately omitted. In addition, because the chemical liquid tank, the heater and the circulation line are included in the decontamination treatment system, these devices may be diverted as an aggregation tank. The aggregation is started by heating and stirring the titanium oxide solution in the aggregation tank, or by adding and stirring an electrolyte agent. When the aggregation starts, the titanium oxide solution becomes cloudy, and accordingly, it is possible to monitor the start of the aggregation with a turbidity sensor. The state that turbidity becomes approximately 300 degrees or higher is preferable as a deposition condition. When the aggregation has started, the titanium oxide solution in a aggregated state is supplied to the main circulation line for decontamination, and is circulated in the recirculation system. The titanium oxide deposits on a pipe surface by the contact of this titanium oxide solution. According to the fourth embodiment of the present invention, there can be provided an apparatus for suppressing the deposition of a radioactive substance, which can efficiently deposit the titanium oxide on a surface of a metal, by providing the aggregation tank in the decontamination system and supplying the titanium oxide solution in the optimum state to the area to be decontaminated. A fifth embodiment according to the present invention will be described hereunder with reference to FIG. 9 and FIG. 10. The fifth embodiment relates to waste liquid treatment of the titanium oxide solution which has finished the deposition treatment. The waste liquid after the deposition treatment onto a pipe surface contains mainly a titanium oxide, and a small amount of a residual reductive-decontamination agent and a dispersing agent, as its component. Thus, in the waste liquid, several hundreds ppm levels of the titanium oxide is contained, and it is therefore necessary to remove this titanium oxide from the waste liquid, to purify the waste liquid, and to discharge the waste liquid into an existing waste-water treatment system. However, because a titanium-oxide waste liquid is in a state including a certain extent of remaining dispersibility, and accordingly, a period of time is needed for the sedimentation of the titanium oxide. Then, a titanium-oxide waste liquid is collected into the aggregation tank 14, and is heated and stirred to progress the aggregation to the state in which the titanium oxide can be easily sedimented. When the aggregation has progressed, stirring is stopped, and the sludge, which has sedimented at the bottom of the aggregation tank 14, is collected into a waste liquid treatment tank 21 through a discharging line 19. As illustrated in FIG. 10, the sludge is subjected to physical dispersion treatment such as crushing, for instance, by an ultrasonic wave in a waste liquid treatment tank 21. Thereafter, the collected titanium oxide is regenerated and reused by the addition of the dispersing agent 22 and stirring with a stirrer 23. On the other hand, a supernatant liquid in the aggregation tank 14 is purified in the ion-exchange resin column of the decontamination apparatus and is discharged to an existing discharging system. According to the fifth embodiment, a load to the ion-exchange resin can be greatly reduced by sedimenting the titanium oxide which occupies the most parts of the waste liquid component, and the amount of a secondary radioactive waste to be produced can be greatly reduced by regenerating and reusing the titanium oxide. 1 . . . pipe, 2 . . . oxide film, 3 . . . titanium oxide, 4 . . . cobalt, 5 . . . cooling water, 6 . . . reactor coolant recirculation system, 7 . . . main circulation line for decontamination, 8 . . . circulation pump for decontamination, 9 . . . chemical liquid tank, 10 . . . heater, 11 . . . ozone generator, 12 . . . mixer, 13 . . . ion-exchange resin column, 14 . . . aggregation tank, 15 . . . circulation pump, 16 . . . stirrer, 17 . . . heater, 18 . . . turbidity sensor, 19 . . . sludge-discharging line, 20 . . . collected titanium oxide, 21 . . . waste-liquid treatment tank, 22 . . . dispersing agent, 23 . . . stirrer, 25 . . . recirculation pump, 26 . . . riser tube nozzle of jet pump. |
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053226445 | summary | BACKGROUND OF THE INVENTION Environmental contamination with radioactive materials is a common problem. The problem may occur as a result of mining operations, such as for uranium, or contamination due to operation of nuclear facilities with inadequate environmental controls, or from the disposal of radioactive wastes. Alternatively, contamination may occur as a result of dispersion of uranium billets which have been used as a high density material in military or civil applications as a result of warfare or civil accident. Mining operations have established practical and economic methods for the economic recovery of some radioactive elements from contaminated materials. The objective of mining, however, is usually the economic recovery of materials and secondary waste is rarely the major issue. In environmental clean-up, the economic objective is to complete effective clean-up with minimum secondary waste at minimum cost, and the value of recovered radioactive substances is of secondary importance. Techniques and chemicals which would not be economical or appropriate for mining applications may become practical for environmental clean-up. It is well established that radioactive elements can be recovered from environmental materials by mechanically washing with water with or without surface active additives. However, such procedures are generally limited to the mechanical separation of solids, and will not remove contaminants that are chemically bound to the solid phase. There are established chemical methods for dissolving insoluble radioactive contaminants in concentrated solvents, such as strong acids, in a process known as acid leaching. Such procedures are effective, but are disadvantageous if the spent concentrated solution ultimately becomes waste. In many cases, the concentrated solvents themselves are hazardous in addition to containing the radioactive contaminant that the process is designed to concentrate. The acid leaching and other processes using concentrated solvents to dissolve the radioactive contaminant have the further disadvantage of also dissolving other contaminants that the process was not designed to remove, such as nonradioactive metals. In the decontamination of internal surfaces of nuclear reactor circuits, early processes involved washing with concentrated chemical solutions to dissolve contaminants to yield a concentrated solution containing the contamination. The processing of these waste solutions was found to be difficult and inconvenient and resulted in them becoming waste and requiring disposal. The technology has now progressed to allow the recovery of radioactivity, typically by ion exchange, in a dilute acidic recirculating system. These solutions, being dilute and acidic, do not contain carbonate and are not particularly useful or appropriate for dissolving actinide elements because they do not form soluble complexes with the actinide elements. In reactor decontamination processes, it has been established that certain organic reagents can be used to dissolve contamination and yield it to an ion exchange resin in a recirculating process in such a way that the organic reagent is continuously re-used. Examples of solutions used in acidic reactor decontamination processes are vanadous formate, picolinic acid and sodium hydroxide. Other processes typically use mixtures of citric acid and oxalic acid. These reactor decontaminating solutions have the disadvantage of not being capable of being used in a single one time application to dissolve actinides, radium, and certain fission products, such as technetium. Previous reactor decontaminating solutions do not contain carbonate and are acidic, dissolving the iron oxides of the radioactive elements commonly found in contaminated reactor circuits. This nonselective metal dissolving capacity is a disadvantage of the acidic solutions and makes them unsuitable for decontamination of material such as soil that contains iron and other metals that are not intended to be recovered. Another disadvantage of acidic solutions is that materials such as concrete or limestone are subject to damage or dissolution in an acidic medium. Also, in dealing with previously known washing solutions for treating soil, these solutions contain too many nonselectively dissolved contaminants preventing subjection of the solution to recovery of contaminants and recirculation of the solution to accomplish further decontamination. It has been established that uranium and transuranic radioactive elements can be dissolved in concentrated acidic (pH<1) chemical systems. The acidity poses difficulties as discussed above. Uranium and sometimes thorium are recovered in mining operations in a concentrated basic medium containing carbonate. The use of concentrated solutions is motivated by the need to dissolve materials at a rate economic for mining operations, and such solutions are not particularly suitable where avoidance of secondary waste is of primary concern. There are also references that suggest that uranium and plutonium can be dissolved in a dilute basic solution containing carbonate, citrate (as a chelating agent) and an oxidizing or reducing agent. Such solutions are not, however, suitable for the recovery of radium/barium sulfate because they do not form soluble complexes from barium sulfate. SUMMARY OF THE INVENTION This invention relates to the recovery of radioactive elements, especially technetium, radium, and actinides such as thorium, uranium and transuranic elements, from certain types of contaminated materials. These materials could be natural, such as soil, or man-made materials, such as concrete or steel, which have become subject on a large scale to contamination. The process of the present invention provides that contaminated material is contacted with a dilute, basic, carbonate recirculating dissolving composition that dissolves radioactive contaminants. Contaminated material can be fed in to the process and cleaned material removed continuously therefrom. The contaminants are recovered from the solution by ion exchange, selective adsorption, reagent destruction, filtration or a combination of these techniques. The recovery steps concentrate the contaminants for recovery in such a way that non-residual reagent constituents do not build up in the system. The recirculating dissolving composition can be applied to small particulate materials such as soil in a contained vessel, or to large standing objects such as concrete walls, or steel structures. It is an object of the invention to provide a method to dissolve and concentrate radioactive contaminants from materials. Another feature of the invention is that the concentrated contamination can be further processed for recovery or disposal. It is a further object of the invention to provide a method for the decontamination of soil and the recovery of radioactive contaminants, which uses a dilute basic carbonate solution to achieve dissolution, thereby minimizing risks of environmental or safety hazards, or structural damage. It is an object of the invention to use chemical systems that dissolve the contaminants in a material as selectively as possible and avoid the dissolution of metals, such as iron and lead. It is another object of the present invention to use a recirculating dissolving system wherein secondary chemical waste is avoided, and reagents do not build up in concentration during the application of the process. |
claims | 1. A steam generator, comprising:a heat transfer tube assembly having a plurality of spiral heat transfer tubes arranged on a concentric circle; anda header formed to connect the heat transfer tube assembly and a nozzle,wherein one end of each of the plurality of spiral heat transfer tubes is extended in a vertical direction to form one end of the heat transfer tube assembly, andthe header is disposed on an upper or lower side of the heat transfer tube assembly to be connected to one end of the heat transfer tube assembly along a vertical direction, andthe header is connected along a horizontal direction to the nozzle disposed at one side of the header, andthe header has an L-shape to switch the flow direction of fluid flowing in a vertical direction through one end of the heat transfer tube assembly to a horizontal direction so as to flow to the nozzle,wherein the header comprises:a tube plate member coupled to one end of the heat transfer tube assembly;a cover member fixed to an opposite side to a side to which one end of the heat transfer tube assembly is coupled in the tube plate member, and connected to an edge of the tube plate member to form a fluid flow space between the tube plate member and the cover member; anda nozzle connection member coupled to a side portion of the tube plate member and a side portion of the cover member to communicate with the flow space formed by the tube plate member and the cover member, and connected to the nozzle. 2. The steam generator of claim 1, wherein one end of the heat transfer tube assembly is inserted into the header along a vertical direction and connected to the header. 3. The steam generator of claim 1, wherein the tube plate member comprises:a tube portion formed in a cylindrical shape to surround one end of the heat transfer tube assembly;a plate portion having a plurality of insertion holes formed to block an upper portion of the tube portion, and formed to be forcibly inserted and coupled to one end of each of the plurality of spiral heat transfer tubes; anda flange portion protruded upward from an edge of the plate portion and coupled to the cover member to form the flow space above the plate portion. 4. The steam generator of claim 3, wherein at least one of the tube plate member, the cover member, and the nozzle connection member has a fitting protrusion protruded from one surface thereof, anda counterpart coupled to at least one of the tube plate member, the cover member, and the nozzle connection member has a fitting groove portion formed to accommodate the fitting protrusion, andthe fitting protrusion is inserted into the fitting groove portion to fix a relative position between the at least one and the counterpart. 5. The steam generator of claim 3, wherein the cover member is fixed to an upper end of the flange portion, andone side of the cover member and one side of the flange portion are open to form a fluid access port for allowing the flow space to communicate with an inner space of the nozzle connection member. 6. The steam generator of claim 3, wherein the plurality of insertion holes are formed at positions spaced apart from each other along a plurality of concentric circles around the center of the tube portion. 7. The steam generator of claim 6, wherein a plurality of spiral heat transfer tubes being forcibly inserted and coupled to a plurality of insertion holes formed along any one of the plurality of concentric circles are formed in the same shape as each other. 8. The steam generator of claim 1, wherein the nozzle connection member is welded and coupled to a circumference of a fluid access port formed by coupling between the tube plate member and the cover member, andthe nozzle connection member comprises:a first portion coupled to a circumference of the access port; anda second portion connected to the nozzle, anda diameter of the first portion is greater than that of the second portion. 9. The steam generator of claim 1, wherein:one side of the tube plate member and one side of the cover member form the nozzle connection portion connected to the nozzle. 10. The steam generator of claim 9, wherein the tube plate member comprises:a tube portion formed in a cylindrical shape to surround one end of the heat transfer tube assembly;a plate portion having a plurality of insertion holes formed to block an upper portion of the tube portion, and formed to be forcibly inserted and coupled to one end of each of the plurality of spiral heat transfer tubes;a flange portion protruded upward from an edge of the plate portion and coupled to the cover member to form the flow space above the plate portion; anda first extension portion protruded from one side of the plate portion and one side of the flange portion to form a lower portion of the nozzle connection portion. 11. The steam generator of claim 10, wherein the cover member comprises:a dome portion fixed to an upper end of the flange portion to form the flow space; anda second extension portion protruded from one side of the dome portion to form an upper portion of the nozzle connection portion, andthe first extension portion and the second extension portion are coupled to each other to form an inner space communicating with the flow space. 12. The steam generator of claim 1, wherein the header comprises:the tube plate member having an open upper portion; andthe cover member formed to cover the open upper portion of the tube plate member, andone end of each of the plurality of spiral heat transfer tubes is inserted and coupled to a lower portion of the tube plate member, andthe fluid flow space is formed inside the tube plate member, andone side of the tube plate member is connected to the nozzle. 13. The steam generator of claim 12, wherein the tube plate member comprises:a tube portion formed in a cylindrical shape to surround one end of the heat transfer tube assembly;a plate portion having a plurality of insertion holes formed to block an upper portion of the tube portion, and formed to be forcibly inserted and coupled to one end of each of the plurality of spiral heat transfer tubes;a flange portion protruded in a dome shape from an edge of the plate portion to form the flow space above the plate portion, and form an opening portion at an upper portion thereof; andthe nozzle connection portion protruded in a pipe shape from one side of the plate portion and one side of the flange portion and connected to the nozzle. |
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description | This application is a divisional under 35 U.S.C. § 121 of U.S. application Ser. No. 11/777,377, filed on Jul. 13, 2007, the contents of which is incorporated by reference in its entirety. The present invention was made with government support under contract number DE-FC07-07ID 14778, which was awarded by the U.S. Department of Energy. The government has certain rights in the present invention. The present teachings relate to systems and methods for controlling the temperature of recirculation coolant in natural recirculation boiling water nuclear reactors to thereby control the power output by the reactor. The statements in this section merely provide background information related to the present disclosure and may not constitute prior art. Typically, in nuclear boiling water reactors (BWRs), the amount of reactivity of the core is controlled by positioning of control rods and adjusting the amount of liquid coolant recirculation flow through the reactor core. For example, the power can be adjusted 30%-40% simply by changing the amount of recirculation flow. Generally, typical BWRs are set to operate at a relatively low power level, i.e., where the fuel bundles are operating at fairly low powers, by moving control rods and then the recirculation flow can be increased to obtain 100% power without moving any control rods. By utilizing the recirculation flow to control power output, the fuel rods and fuel pellets change power levels at a relatively slow, uniform rate, thereby avoiding pellet clad interaction (PCI) and damage to the fuel rod cladding. However, natural circulation boiling water reactors (NCBWRs) do not include recirculation pumps, but instead employ natural circulation. Therefore, recirculation flow cannot be utilized to control the power output of NCBWRs. At least some known NCBWRs manipulate the control rods to control the power level of fuel bundles, which causes the power level to change at fairly rapid, non-uniform rates. Increasing the power too rapidly as the control rods are withdrawn can cause significant damage to the fuel cladding. According to one aspect, a system for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) is provided. In various embodiments, the system includes a heating subsystem for heating feedwater flowing into a reactor vessel of a NCBWR to increase the temperature of recirculation water flowing through the core above a predetermined recirculation water operating temperature. Additionally the system includes a temperature sensor operable to sense the temperature of the feedwater flowing into the reactor vessel. The temperature sensor is communicatively coupled with a temperature controller operable to, based on temperature readings from the temperature sensor, command the heating subsystem to increase the temperature of the feedwater flowing into the reactor vessel to a requested temperature above a predetermined operating temperature of the feedwater flowing into the reactor vessel. By increasing the temperature of the feedwater flowing into the reactor vessel, the temperature of the recirculation water is increased above the predetermined recirculation water operating temperature flowing into the core causing a reduction in the power level generated by the NCBWR core. According to another aspect, a method for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) is provided. In various embodiments the method includes sensing the temperature of the feedwater flowing into the reactor vessel, utilizing a temperature sensor, and increasing the temperature of the feedwater flowing into the reactor vessel, utilizing a heating subsystem of the NCBWR. The heating subsystem increases the temperature of the feedwater flowing into the reactor vessel to a requested temperature above a predetermined operating temperature of the feedwater flowing into the reactor vessel. Thus, the temperature of the recirculation water flowing into the core is increased above the predetermined recirculation water operating temperature, resulting in a reduction of the power level generated by the NCBWR core. Further areas of applicability of the present teachings will become apparent from the description provided herein. It should be understood that the description and specific examples are intended for purposes of illustration only and are not intended to limit the scope of the present teachings. The following description is merely exemplary in nature and is in no way intended to limit the present teachings, application, or uses. Throughout this specification, like reference numerals will be used to refer to like elements. Referring to FIG. 1, a general schematic of a natural circulation boiling water nuclear reactor (NCBWR) 10 that includes a power level control system 14, is provided. The NCBWR 10 generally includes at least one principal feedwater heater 18, a reactor vessel 22, steam line header 26, a high-pressure steam turbine 30, a moisture separator reheater 34 and low-pressure steam turbine 38. In various embodiments, as illustrated in FIG. 1 and described herein, the NCBWR 10 includes a plurality of principal feedwater heaters 18. The principal feedwater heaters 18 each include a feedwater heater core (not shown) that receives feedwater supplied via a feedwater input line 42 and a feed water pump 46, or via a feedwater link line 50 and another principal feedwater heater 18. Each principal feedwater heater 18 additionally includes a steam shell (not shown) that surrounds the respective heater cores and is structured to receive high-pressure, high-temperature steam from the moisture separator reheater 34 and/or the high-pressure steam turbine 30, via steam extraction lines 52 and/or 53. Generally, steam from the high-pressure steam turbine 30 and/or the moisture separator reheater 34 is diverted to the principal feedwater heaters 18 and circulated through the respective shells to heat the feedwater flowing through the respective heater cores to a predetermined temperature, such as a normal operating temperature, e.g., less than 420° F. The heated feedwater is then fed into an annulus 25 of the reactor vessel 22, via a reactor vessel feedwater inlet line 56, where the heated feedwater is mixed with and becomes part of a recirculation flow circulating into a reactor core 54. The recirculation flow cools the fuel rods within the reactor core 54 which, in turn, converts the water coolant to high temperature steam, e.g., 540° F., that exits the reactor vessel 22 through a plurality of reactor vessel output steam lines 58. The output steam lines 58 terminate into the steam line header 26 that regulates and controls high-temperature, high-pressure steam output to the high-pressure steam turbine 30, via a high-pressure turbine feed steam line 62. The high-temperature, high-pressure steam fed into the high-pressure steam turbine 30 is utilized to spin the turbine 30. Subsequently, at least a portion of the high-pressure, high-temperature steam exits the high-pressure steam turbine 30, via a high-pressure turbine outlet line 66. The steam output from the high-pressure turbine 30 is fed into the moisture separator reheater 34 where liquid water is removed from the steam to dry the steam. The dried steam is then fed into the low-pressure steam turbine 38. The high-temperature, low-pressure steam fed into the low-pressure steam turbine 38 is utilized to spin the turbine 38. The power level control system 14 is structured and operable to increase the temperature of the feedwater exiting the principal feedwater heaters 18, prior to the feedwater flow entering the annulus 25 via the reactor vessel feedwater inlet line 56. More particularly, the power level control system 14 is structured and operable to controllably heat the feedwater flowing into the annulus 25 to thereby controllably increase the temperature of the recirculation water flow to a temperature above a predetermined normal recirculation water operating temperature. Increasing the temperature of the recirculation water flow to a temperature above a predetermined normal recirculation water operating temperature controllably reduces the power level output by the reactor core 54. Generally, the power level control system 14 includes a heating subsystem 70 and a temperature sensor 74 communicatively connected to a temperature controller 75 that is communicatively connected to the heating subsystem 70. The temperature sensor 74 senses the temperature of the feedwater flowing into the reactor annulus 25 and communicates with the temperature controller 75 to monitor and control the temperature of the feedwater flowing into the reactor annulus 25. Based on temperature commands received from a control center 78, the temperature controller 75 sends command signals to the heating subsystem 70 to increase the temperature of the feedwater flowing into the reactor annulus 25 to a commanded, or requested, temperature above a normal operating temperature of the feedwater flowing into the annulus 25. The heated feedwater is fed into the annulus 25 to mix with the recirculation flow to controllably increase the temperature of the recirculation flow within the reactor core 54 to a temperature above a normal operating temperature of the recirculation flow. Accordingly, the controlled increase in temperature of the recirculation flow controllably reduces the power output level of the reactor core 54. Referring now to FIG. 2, in various embodiments, the heating subsystem 70 includes the steam header 26, at least one high temperature feedwater heater (HTFH) and a steam bypass valve 82. In various implementations, each HTFH is a supplemental feedwater heater 86 structured to receive an input flow of feedwater from one or more of the principal feedwater heaters 18 into a heater core (not shown). Additionally, each supplemental feedwater heater 86 is structured to receive high temperature steam from the steam header 26, via a header steam diversion line 84, into a shell (not shown) surrounding the respective heater core. Each supplemental feedwater heater 86 is further structured to output the flow of feedwater to the reactor annulus 25. When the temperature controller 75 receives commands to increase the temperature of the feedwater flowing into the reactor annulus 25, the temperature controller 75, in turn, communicates command signals to the steam bypass valve 82. The steam bypass valve 82 receives the commands from the temperature controller 75 and, based on the received commands, controls an amount of flow of the high temperature steam from the steam header 26 into the shell(s) of the supplemental feedwater heater(s) 86. As the high temperature steam circulates through the respective shell(s), the feedwater from the principal feedwater heater(s) 18 flowing through the respective supplemental feedwater heater core(s) is heated. Based on temperature reading sensed by the temperature sensor 74, the temperature controller 75 communicates with the steam bypass valve 82 to control the amount of high temperature steam flowing into the supplemental feedwater heater(s) 86. Thus, based on the requested temperature received from the control center 78, the temperature sensor 74 and controller 75 establish a feedback loop to control the increase in temperature of the feedwater flow output to the reactor annulus 25. More particularly, the sensor/controller 74/75 feedback loop adjusts temperature of the feedwater flow to obtain the requested temperature above the normal operating temperature of the feedwater flowing into the reactor annulus 25. In various embodiments, an orifice 88 is provided in a bypass line 89 around the steam bypass valve 82 to provide for a constant steam flow through the header steam diversion line 84 in order to keep the header steam diversion line 84 and the shell side of the supplemental/high temperature feedwater heater(s) 86 warm and ready for use on demand. With further reference to FIG. 2, in various embodiments, the heating subsystem 70 further includes a feedwater bypass valve 90 operable to also receive commands from the temperature controller 75. In response to the received commands, the feedwater bypass valve 90 controls an amount of flow of the feedwater through a heater bypass line 94 that directs a flow of feedwater from the feedwater pump 46 into the reactor annulus 25. The temperature of the feedwater flowing from the feedwater pump 46 is significantly lower than the feedwater flowing from the principal and supplemental feedwater heaters 18 and 86. Therefore, an increase in the flow of feedwater through the heater bypass line 94, as commanded by the temperature controller 75, will decrease the temperature of the feedwater flowing into the reactor annulus 25. For example, in response to a request from the control center 78 to increase the power level output by the reactor core 54, i.e., a request to lower the temperature of the feedwater flowing into the reactor annulus 25 to a specific temperature, the temperature sensor and controller combination 74/75 will increase the flow of feedwater flowing through the heater bypass line 94. The increase of lower temperature water flowing through the heater bypass line 94 will reduce the temperature of the feedwater flowing into the reactor annulus 25. More particularly, the temperature sensor 74 and temperature controller 75 control the flow of feedwater through the heater bypass line 94 to reduce the temperature of the feedwater flowing into the reactor annulus 25 to reduce the temperature of the recirculation water within the reactor core 54 to a requested temperature below the normal operating temperature, thereby increasing the power level generated by the reactor core 54. Therefore, in various embodiments, the power level control system 14 controls the power level output by the reactor core 54 by increasing the temperature of the feedwater flow to decrease the power output, and decreasing the temperature of the feedwater flow to increase the power output. Referring now to FIGS. 3 and 4, in various other embodiments, the NCBWR includes a plurality of the principal feedwater heaters 18, wherein at least one of the principal feedwater heaters 18 also functions as HTFHs. In some embodiments, each of the principal feedwater heaters 18 is structured to also function as an HTFH, as shown in FIG. 3. In such embodiments, each HTFH/principal feedwater heater 18 is structured to receive high-pressure, high-temperature steam from the high-pressure steam turbine 30 and/or the moisture separator reheater 34, via steam extraction lines 53 and/or 52, to heat the flow of feedwater output to the reactor core 54 to the normal operating temperature. Additionally, in such embodiments, each HTFH/principal feedwater heater 18 is structured to receive high temperature steam from the steam header 26 to increase the temperature of the feedwater output to the reactor annulus 25 to a requested temperature above the normal operating temperature, as described above. Alternatively, in some embodiments, the principle feedwater heaters 18 nearest the feedwater pump 46 (the initial principle feedwater heaters 18) would receive extraction steam only from the high pressure steam turbine 30 via steam extraction lines 53 and not from either of the steam header 26 or the moisture separater/reheater 34. Accordingly, the temperature of feedwater flowing from the initial principle feedwater heaters 18 to the subsequent principle feedwater heaters 18 will increase above the normal operation temperature. In other embodiments, only a portion of the plurality of principal feedwater heaters 18 are structured to also function as a HTFH, as shown in FIG. 4. In such embodiments, each non-HTFH principal feedwater 18 is structured to receive high-pressure, high-temperature steam from the high-pressure steam turbine 30 and/or the moisture separator reheater 34, via steam extraction lines 53 and/or 52, to heat the flow of feedwater output to the reactor annulus 25 to the normal operating temperature, as described above. However, each HTFH/principal feedwater heater 18 is structured to receive high temperature steam from the steam header 26, via the header steam diversion line 84. Each HTFH/principal feedwater heater 18 is structured such that, when normal operation temperature of the feedwater is requested, steam from the steam header 26 is utilized by the HTFH/principal feedwater heater(s) 18 to heat the flow of feedwater output to the reactor annulus 25 to the normal operating temperature. Furthermore, when a feedwater temperature above the normal operation temperature is requested, each HTFH/principal feedwater heater 18 utilizes additional high temperature steam from the steam header 26 to increase the temperature of the feedwater output to the reactor core 54 to the requested temperature. Thus, with further reference to FIGS. 3 and 4, when the temperature controller 75 receives commands to increase the temperature of the feedwater flowing into the reactor annulus 25, the temperature controller 75 commands the steam bypass valve 82 to increase the amount of high temperature steam allowed to flow from the steam header 26 into each HTFH/principal feedwater heater 18. The increase of high temperature steam flowing through the shell of each HTFH/principal feedwater heater 18 increases the temperature of the feedwater flowing through the respective heater core(s) and output to the reactor annulus 25 to the requested/commanded temperature. Accordingly, the increase in temperature of the feedwater flowing into the reactor annulus 25 increases the temperature of the recirculation water within the reactor core 54, thereby reducing the power output level of the reactor vessel 22, without manipulation of control rods within the reactor vessel 22. Referring now to FIG. 5, in various embodiments, the heating subsystem 70 includes a heating device 98 structured and operable to heat the feedwater flowing from the principal feedwater heater(s) 18 into the reactor annulus 25. More particularly, the heating device 98 is responsive to commands from the temperature controller 75 to increase the temperature of the feedwater flowing into the reactor annulus 25 to a requested temperature above the normal operating temperature, as sensed by the temperature sensor 74. For example, in response to a request from the control center 78 to decrease the power level output by the reactor core 54 the temperature controller 75 will command the heating device 98 to increase the temperature of the feedwater flowing into the reactor annulus 25 to the requested temperature. Accordingly, the temperature of the recirculation flow within the reactor core 54 will be increased and the power level output by the reactor vessel 22 will decrease to a requested level that corresponds to the temperature reduction of the recirculation flow. The heating device 98 may be any heating device suitable for heating the feedwater flowing into the reactor annulus 25 to the requested temperature above the normal operating temperature produced by the principal feedwater heaters 18. For example, the heating device 98 may be a suitable gas or electric power water heating device. Referring now to FIG. 6, in various embodiments, each of the principal feedwater heaters 18 is structured to function as a HTFH. In such embodiments, each HTFH/principal feedwater heater 18 is structured to receive high-pressure, high-temperature steam from the high-pressure steam turbine 30 and/or the moisture separator reheater 34, via steam extraction lines 53 and/or 52, to heat the flow of feedwater output to the reactor annulus 25 to a temperature above the normal operating temperature. Additionally, in such embodiments, the heating subsystem 70 includes a feedwater bypass valve 102 operable to receive commands from the temperature controller 75. In response to the received commands, the feedwater bypass valve 102 controls an amount of flow of the feedwater through a heater bypass line 106 that directs a flow of feedwater from the feedwater pump 46 into the reactor annulus 25. The temperature of the feedwater flowing from the feedwater pump 46 is significantly lower than the feedwater flowing from the HTFH/principal feedwater heaters 18. Therefore, an increase in the flow of feedwater through the heater bypass line 106, as sensed by the temperature sensor 74 and commanded by the temperature controller 75, will decrease the temperature of the feedwater flowing into the reactor annulus 25. Thus, in such embodiments, the temperature sensor and controller combination 74/75 controls the operation of the feedwater bypass valve to reduce the temperature of the feedwater flowing into the reactor annulus 25 to any requested temperature. For example, in response to a request from the control center 78 to achieve a given power level output by the reactor core 54, e.g., 90% maximum power output, the temperature controller 75 will allow a flow of feedwater through the heater bypass line 106 sufficient to cool the feedwater flowing out of the HTFH/principal feedwater heaters 18 to a temperature sufficient to achieve the requested power level output. Accordingly, to increase the power level output, the temperature controller 75 will command an increase in the flow of feedwater through the heater bypass line 106 to reduce the temperature of the feedwater and recirculation flows, as sensed by the temperature sensor 74. Conversely, to decrease the power level output, the temperature controller 75 will command a decrease in the flow of feedwater through the heater bypass line 106 to increase the temperature of the feedwater and recirculation flows. Referring now to FIG. 7, in accordance with various embodiments, two independent mechanisms of power output level change of the reactor core 54, i.e., control rod movement and feedwater temperature change, may be combined to construct a Core Power-Feedwater Temperature Map, such as the exemplary map illustrated in FIG. 7. The line B-A represents the power ascension line, e.g., from approximately 25% to 100% power, by the conventional control rod withdrawal only. The temperature of the feedwater flowing into the reactor annulus 25 will increase automatically since more high-pressure, high-temperature steam from the high-pressure steam turbine 30 and/or the moisture separator reheater 34 is available to heat the feedwater as the reactor power level increases. The slope of line B-A depends on the steam extraction points and the design of the balance of the plant system including the principal feedwater heaters 18. The line A-C represents the reduction of reactor power as the temperature of the feedwater flowing into the reactor annulus 25 is increased above the normal operating temperature, e.g., 420° F. using the various power level control system 14 embodiments described above. The A-C line may also be referred to as the 100% load line to indicate that no control rod movement is employed along the A-C line. A ‘Rod Block’ line above and parallel to the line A-C may be developed and implemented to ensure that feedwater temperature is not changing in a manner that would allow the fuel pins to get too hot. The D-C line represents a path where both feedwater temperature and control rod positions are changed in small steps. The D-C line may be followed during startup of the reactor vessel 22 to bring the reactor core to point C, e.g., 85% power with a recirculation flow temperature above the normal operating temperature. Necessary control rod adjustments may be done and the feedwater temperature may be reduced to traverse the path from C to A in order to bring the reactor vessel 22 to 100% power output with the temperature of the feedwater flowing into the reactor annulus 25 at normal operating temperature, e.g., approximately 420° F. The reversed path A-C-D-B may be traversed during power suppression testing or other reactor power maneuvering. The hatched area of FIG. 7 is an exemplary representation of a possible operating domain of the NCBWR 10 using temperature control of the feedwater flowing into the reactor annulus 25 in conjunction with different control rod positioning to control the power level output by the reactor vessel 22, as described above. Thus, operators at the control center 78 may utilize Power-Feedwater Temperature Map as a tool and/or guide to provide a conceptual understanding of the effects on power level output that will result from any particular commanded change in feedwater temperature. Additionally, the control center 78 may store the Power-Feedwater Temperature Map in a memory device 108, or other database, and execute an control algorithm stored on memory device 108 to automatically communicate with the temperature controller 75 to automatically control the temperature of the feedwater flowing into the reactor annulus 25, as described above. That is, an operator may select a desired power level output, i.e., input a desired power level output into computer-based systems of the control center 78. In response thereto, the computer-based systems of the control center 78 will execute the control algorithm and access the stored Power-Feedwater Temperature Map to obtain the temperature to produce the desired power level output. Subsequently, the control center 78 will communicate a feedwater temperature command, or a sequence of feedwater temperature commands, to the temperature controller 75. These commands provide, at least in part, the temperature obtained from the Power-Feedwater Temperature Map for producing the desired power level output. Thus, execution of the control algorithm generates one or more temperature command signals set to the temperature controller 75. In response to the commanded temperature, the temperature controller 75 establishes a feed-back control loop, based on output from the temperature sensor 74, to automatically adjust the temperature of the feedwater flowing into the reactor annulus 25 to the commanded temperature. More specifically, in response to the temperature command signal(s), the temperature controller 75 iteratively communicates with the temperature sensor 74 to control the operation of the applicable bypass valve, e.g., bypass valve 82, 90 or 102, to adjust the temperature of the feedwater flowing into the reactor annulus 25. Moreover, automatically adjusting the temperature of the feedwater flowing into the reactor annulus 25, automatically adjusts the temperature of the recirculation flow circulating into a reactor core 54, resulting in adjustments to the power level output to the desired level. The control center 78 may further include other computer based components (not shown), such as a processor, a display, a user interface, e.g., a keyboard, mouse, stylus, touch screen etc. and other interfaces and/or memory devices suitable for executing and performing the automated control described above. Thus, the power level control system 14, described herein is structured and operable to independently change the temperature of the feedwater flowing into the reactor annulus 25 to thereby uniformly change the core power output level without the need for control rod movement. The description herein is merely exemplary in nature and, thus, variations that do not depart from the gist of that which is described are intended to be within the scope of the teachings. Such variations are not to be regarded as a departure from the spirit and scope of the teachings. |
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description | This application claims the priority benefit of French patent application number 10/54304, filed on Jun. 2, 2010, entitled “METHOD FOR DESIGNING MASKS USED TO FORM ELECTRONIC COMPONENTS,” which is hereby incorporated by reference to the maximum extent allowable by law. 1. Field of the Invention The present invention relates to the designing of masks to be used to manufacture integrated circuits. More specifically, the present invention relates to a method for improving the design of such masks. 2. Discussion of the Related Art To manufacture integrated electronic circuits, a set of several masks comprising openings defining work areas on the circuit is used. For example, different masks may be successively used to define locations of dopant implantation, of etching, etc. The manufacturing of the different masks necessary to obtain an integrated circuit implies a relatively high cost. Further, modern integrated circuits may require, for their production, several tens of masks. It is thus essential to properly test the mask generation files before launching the production of the masks, or even the production of the integrated circuits. Especially, the compatibility of the masks, for example, for their superposition, should be optimal. Having to redesign a set of masks, at the last minute before the launching of the production of integrated circuits or after the launching of the production, may imply very large manufacturing delays and costs. For each electronic component technology, integrated circuit designers should comply with a number of design rules put together in a “Design Rules Manual”, or DRM. Such a manual gathers, among other things, sets of rules relative to the superposition or to the juxtaposition of the layers necessary for the forming of the electronic components. FIG. 1 illustrates a few examples of rules that may be imposed on designers for the forming of integrated circuits. This drawing shows, in hatchings, different areas at the surface of a substrate intended to receive electronic components. Among the rules to be respected for electronic components, the following can be mentioned: respecting a minimum width W of some elements of the components, for example, the width of MOS transistor gates, the length of a transistor channel; respecting a minimum space S between different elements, to avoid interferences between these elements, for example, between two metal tracks, or again leakage currents; respecting a minimum surface area A for some elements. FIG. 2 illustrates rules that may be imposed by DRMs when several layers are used to form electronic components, at close or superposed locations of a circuit. This drawing shows the design levels from which are formed masks which will subsequently be necessary to the manufacturing of electronic components, LAYER1 for which the contour of the openings is shown in full lines, and LAYER2 having its openings shown in hatched portions. The examples of rules to be respected may be: respecting a minimum enclosure E between the edge of the openings of the second mask LAYER2 and the edge of the openings of the first mask LAYER1. This, for example, corresponds to the case of electronic components formed in a well of a specific conductivity type. In this case, for their proper operation, the components should not be formed too close to the edge of the well. This may also correspond to elements which should be formed in superposed fashion: for example, a transistor gate above a well of a specific conductivity type. respecting a minimum distance D between elements formed by means of second mask LAYER2 with respect to elements formed by means of first mask LAYER1. This, for example, corresponds to the case where the first mask defines a well of a given conductivity type and where the components formed at the level of the openings of the second mask should not be formed too close to this well due to a risk of interactions. It should be noted that the rules imposed by DRMs may also integrate an alignment error margin to take into account inaccuracies in the mask alignment on manufacturing of the circuits. The rules imposed by DRMs thus eliminate a number of situations with critical sizings, which could not operate properly and which are thus not accessible to designers. Once the integrated circuits have been designed by the designers, the obtained CAD files should be turned into image files of the masks which will be necessary to form the integrated circuits. To achieve this, logical operations are defined by technologists to be applied to the integrated circuit files. The logical operations also define all the elements missing for the proper operation of the circuit, which are not available to designers. Indeed, for an easy design of integrated circuits, designers only define some of the elements necessary to the forming of the integrated circuit. For example, in the case of a MOS transistor, designers may define the location of a well of a given conductivity type, and a second well necessary to the proper operation of the transistor may be automatically generated by the logical operations. The logical operations finally define an optimized shape of the different masks. They may in particular provide to slightly widen the openings formed in the masks to compensate for a possible subsequent narrowing when the mask is being used. FIG. 3 shows a conventional flowchart of the steps carried out to design integrated circuits, until the manufacturing of the masks used for manufacturing the integrated circuits. As described hereabove, a first step 10 comprises forming a computer file which is an image of the desired integrated circuit (DESIGN FILE). This file is formed by designers 12 (DESIGNER), in compliance with the rules imposed by integrated circuit DRMs 14 associated with the technology used. The integrated circuit file is then transformed, by a computer system, at a step 16 (LOP, Logical Operation Processing) and by means of a set of logical operations 18 (LO), to obtain an image file of the masks necessary to manufacture integrated circuits 20 (MASK FILES). The logical operations are especially provided to gather, within a same mask, the regions of the electronic components of the integrated circuits requiring a same processing. As an example, low-voltage MOS transistors, high-voltage MOS transistors, dual-gate transistors, etc. may be provided on a same integrated circuit. Each of these transistors requires, to be formed, a specific processing, often obtained by a mask differentiation, for example, to form the wells of these different transistors. The logical operations of step 18 are used to generate the right masks according to the different steps to be carried out. Step 16 of transformation of integrated circuit file 10 into a mask file may return errors, for example, in the case where the density of electronic components on the circuit would be too high, or in the case where there would be an incompatibility with the integrated circuit design rule manual. In this case, it is necessary to revise the transformation formulas 18 (LO) applied in transformation step 16 to validate or invalidate certain configurations provided by the designers. Once step 18 has been carried out, all masks 20 are visually verified by a technologist (to spot evident errors, for example, of superposition of elements which should not be superposed), then is tested statistically again, at a step 22 (PLC) before the mask production. This last test, performed by a computer system, is a dimensional verification of the generated masks, for example, in comparison with dimensional criteria imposed by mask manufacturers (MRC, Manufacturing Rule Check), or with criteria imposed by integrated circuit manufacturers (PLC, Post Logical Check). If test step 22 generates errors, a step 24 (ERROR) is provided, to modify logical operations 18 of transformation of step 20 so that the masks fulfill the conditions imposed by the mask manufacturers. This modification step is carried out manually by technologists and may be relatively long. Indeed, among all logical operations, the one having caused the incompatibility with the dimensional criteria imposed by the mask manufacturers should be targeted, and the required operation(s) should then be eliminated. Once logical operations 18 have been modified, the operation of transformation of the integrated circuit file into a mask file is applied again to the integrated circuit file provided by the designers. If an error still occurs after test step 22, logical operations 18 are modified again and the transformation operation of step 16 is repeated as many times as necessary. When test step 22 is validated, the masks are sent to production at a step 26 (MASK FAB) and the integrated circuit production may start. A problem may arise in specific cases where the designers desire to integrate new components in the integrated circuits. “New component” here means an entirely new component or a new adaptation of a known component, for example, the adding of a doped region at a new location of a transistor, the modification of the dimensions of an insulated gate, etc. When a new component is designed, the integrated circuit in which this component is provided may be transformed according to the method described in relation with FIG. 3 to obtain the set of masks corresponding to this circuit, if this set of masks can be generated with no error. It is generally provided, before performing this transformation, to form a test file in which many configurations of the new component, in interaction with other components, are gathered. This test file is then tested to see if it complies with the rules imposed by the DRMs, after which it is transformed by means of the logical operations. This enables to verify that this new component poses no problem, related to the DRMs, of integration into the desired integrated circuit, but also into other future configurations that may be given thereto. However, it is possible to have a test circuit comprising new components complying with the conditions imposed by the DRMs, where the transformations of the logical operations pose no problem, with a good post-transformation test regarding the criteria of mask manufacturers, but with finally produced masks which do not provide high-quality components. This may be due to the fact that the logical operations transforming the integrated circuit file into a mask file may incorrectly process the design of the new component, or may introduce errors during the transformation. If an erroneous set of masks is used to produce integrated circuits, this may have significant consequences in terms of time and cost, especially if an entire new set of masks has to be designed and manufactured. It thus cannot be envisaged to detect errors at the end of the mask manufacturing process. A method for limiting as much as possible the need to redesign integrated circuit masks is thus needed. An embodiment provides a method for designing integrated circuit manufacturing masks overcoming all or part of the disadvantages of usual methods. More specifically, an embodiment is a method for designing integrated circuit manufacturing masks implementing particularly efficient test steps. Thus, an embodiment provides a method for designing masks adapted to the forming of integrated circuits in a considered technology, comprising the steps of: (a) forming a first test file comprising a set of randomly-generated configurations of integrated circuit elements arranged according to layouts that may exceed the cases authorized by design rule manuals; (b) forming a second test file comprising all the elements of the first test file, less the elements corresponding to configurations forbidden by design rule manuals; (c) transforming the second test file by means of a set of logical operations implemented by computing means to obtain a mask file comprising the configuration of the set of masks necessary to obtain the integrated circuit associated with the second test file; (d) testing the mask file and, if the test is negative, modifying and adapting the design rule manuals according to the test result; and (e) reiterating steps (a) to (d) as many times as necessary until the test of step (d) is positive. An embodiment provides a method for designing masks adapted to the forming of integrated circuits in a considered technology, comprising the steps of: (a) forming a first test file comprising a set of configurations of integrated circuit elements arranged according to layouts that may exceed the cases authorized by design rule manuals, generated by using mathematical models of interaction of segments or of polygons; (b) forming a second test file comprising all the elements of the first test file, less the elements corresponding to configurations forbidden by design rule manuals; (c) transforming the second test file by means of a set of logical operations implemented by computing means to obtain a mask file comprising the configuration of the set of masks necessary to obtain the integrated circuit associated with the second test file; (d) testing the mask file and, if the test is negative, modifying and adapting the design rule manuals according to the test result; and (e) reiterating steps (a) to (d) as many times as necessary until the test of step (d) is positive. According to an embodiment, step (d) further comprises the step of modifying and adapting the logical operations of transformation of the second test file. According to an embodiment, step (d) further comprises the step of modifying and adapting the test of the mask file. According to an embodiment, step (e) is followed by the steps of: (f) forming an integrated circuit file respecting the rules imposed by the modified and adapted design rule manuals obtained by the last reiteration of step (d); and (g) transforming the integrated circuit file into a mask file by applying the logical transformation operations modified and adapted at step (d). According to an embodiment, step (g) is followed by a step of manufacturing of a set of integrated circuit masks based on the mask file obtained at step (g). According to an embodiment, the first test file comprises a set of randomly generated configurations of integrated circuit elements. According to an embodiment, the first test file comprises a set of configurations of integrated circuit elements generated by using mathematical models of interaction of segments or of polygons. The foregoing and other objects, features, and advantages will be discussed in detail in the following non-limiting description of specific embodiments in connection with the accompanying drawings. FIG. 4 is a flowchart of a method provided to improve the design of integrated circuit masks, and especially when new components are provided by integrated circuit designers. More specifically, this drawing illustrates steps of a method for performing efficient and robust integrated circuit file tests, to avoid that erroneous masks are formed based on new components which comply with conventional tests. The method described in relation with the flowchart of FIG. 4 will preferably be followed for each new electronic component technology. This method enables defining optimized integrated circuit design rules which may, later on, be adapted to any new component in the considered technology. The method of FIG. 4 starts with a step 30 (ALL CONF) of forming of a complete test file. This complete test file comprises all the possible configurations of the various elements of a given technology (doped wells, metal tracks, etc.), be they known and existing configurations or configurations which do not appear as achievable. It should be noted that “all the possible configurations of the various available elements” means that the complete test file for example comprises a set of random configurations of all the elements available to the designer, or also a set of configurations generated by using mathematical models of interactions of segments or of polygons. Any other method of formation of such a complete test file, comprising many configurations of the elements available to the designers, may also be used. The complete test file is then tested (step 32) by means of the design rules imposed in the DRMs associated with this technology (step 34) to form an improved test file (step 36, IMPROVED TEST CASE) comprising all the configurations of complete test file 30 less all the configurations which do not comply with the rules imposed by DRMs 34. Thus, the improved test file comprises all the configurations authorized by the DRMs of the considered technology, be they conventional configurations (currently used by designers) or unusual configurations which are not filtered by the DRMs. The obtained improved test file 36 is then transformed, at a step 38 (LOP), by a set of logical operations 40 (LO) implemented by computing means, for example, a computer, and defined by technologists, to form a mask set (MASK SET) at a step 42. The mask set obtained at step 42 is then submitted to a first visual test performed by a technologist, then to a mask post-generation test at a step 44 (PLC). The mask post-generation test 44 enables verifying that the obtained masks comply not only with the dimensional requirements of mask manufacturers, but also with the requirements of integrated circuit manufacturers. If an error is detected during one of the tests of step 44 (FAIL), and this error corresponds to a case which should not be reproduced afterwards in the considered technology, it is provided to report the specific case having caused the error directly at the level of design rules 34, at a step 46 (MOD DRM). Thus, such a situation can not be allowed afterwards by the test performed by means of the DRMs, for new integrated circuits. If the detected error does not correspond to a case to be forbidden afterwards for new circuits, in the considered technology, or corresponds to a case to be refined and/or verified differently, it may be provided, at a step 48 (MOD LO), to modify the logical operations 40 performed at step 38. The set of steps 32, 36, 38, 42, and 44 is then repeated as often as necessary until a mask post-generation test step 44 providing no error is obtained (PASS). In this case, it is proceeded to a step 50 (ROBUST DRM/LO) where it can be said that the subsequent integrated circuit configurations authorized by the modified DRMs will provide high-quality and error-free mask files. The mask files thus obtained will be adapted to the manufacturing of high-quality and error-free integrated circuits. It should be noted that the tests of mask post-generation test step 44 may also be modified and adapted when steps 32 to 42 are repeated. For example, some test steps may be modified and made less demanding if such a requirement appears not to be necessary. Indeed, if a new integrated circuit is provided by the designers, this circuit will follow the conventional methods of transformation of the integrated circuit into a corresponding mask file for the integrated circuit manufacturing. FIG. 5 illustrates the timing diagram of such a method. A new integrated circuit design file (NEW DESIGN FILE) is created at a step 60 by the integrated circuit designers. This integrated circuit design file complies with the integrated circuit design rules imposed by the improved design rule manuals formed by the method of FIG. 4 (IMPROVED DRM, step 62). The integrated circuit file is then transformed, in a step 64 (IMPROVED LOP), by the improved logical operations formed by the method of FIG. 4 (step 66, IMPROVED LO). The mask file thus obtained may be tested again with the mask post-generation steps, at a step 70 (PLC). If such a test is provided, the modified test will preferably be used on repeating of the method of FIG. 4, if it has been modified. A mask set is thus obtained at a step 72 (MASK FAB), the masks being of good quality. Thus, once the method described in relation with FIG. 4 has been carried out, any new integrated circuit file can be transformed into a mask file without fearing errors in the masks since the method described in relation with FIG. 4 has eliminated all forbidden design configurations. Specific embodiments of the present invention have been described. Various alterations, modifications and improvements will occur to those skilled in the art. In particular, any other test step generally provided in addition to the steps disclosed herein for the manufacturing of integrated circuit masks may be carried out in combination with the method steps disclosed herein. It should be noted that, conventionally, the complete test file, the improved test file, the mask files, and the integrated circuit design files may be formed in the GDSII file format, currently used for such files. It should however be noted that any other integrated circuit file format (for example, OASIS) may be used to form these files. Such alterations, modifications, and improvements are intended to be part of this disclosure, and are intended to be within the spirit and the scope of the present invention. Accordingly, the foregoing description is by way of example only and is not intended to be limiting. The present invention is limited only as defined in the following claims and the equivalents thereto. |
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abstract | An insulated solution injector may include an outer tube and an inner tube arranged within the outer tube. The outer tube and the inner tube may define an annular space therebetween, and the inner tube may define a solution space within. The annular space may be configured so as to insulate the solution within the solution space. As a result, the solution may be kept to a temperature below its decomposition temperature prior to injection. Accordingly, the decomposition of the solution and the resulting deposition of its constituents within the solution space may be reduced or prevented, thereby decreasing or precluding the occurrence of a blockage. |
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abstract | A lid frame for a nuclear fuel assembly shipping container and a shipping container for nuclear fuel assemblies are provided. The shipping container can include a lower container in which a cradle is installed, an upper container detachably coupled to the lower container, and a base frame coupled to the cradle with at least one nuclear fuel assembly placed thereon. The lid frame can include a plurality of supports installed apart from each other so as to surround the nuclear fuel assembly placed on the base frame, and a plurality of clamps separated from each other, coupled to the plurality of supports perpendicular to the plurality of supports, rotatably hinged to the base frame, and configured to clamp the nuclear fuel assembly. The lid frame safely protects the nuclear fuel assembly that is being transported. |
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claims | 1. A two-dimensional X-ray scanning system comprising:an X-ray scanner comprising:a beam focuser;a beam steerer for scanning an electron beam on a path along an X-ray production target as a function of time; andan aperture adapted for travel in an aperture travel path relative to the X-ray production target, wherein the X-ray scanner remains stationary with respect to the object of inspection; anda detector configured to detect X-rays passing through an object or scattered by the object of inspection and generate two-dimensional data indicative of the detected X-rays. 2. A The two-dimensional X-ray scanning system of claim 1, wherein the aperture is an intersection of a fixed slit and a moving slit. 3. The two-dimensional X-ray scanning system of claim 1, wherein the X-ray production target is a planar target block. 4. The two-dimensional X-ray scanning system of claim 1, wherein the X-ray production target is convex. 5. The two-dimensional X-ray scanning system of claim 4, wherein the X ray scanner is configured to have a predefined take-off angle and wherein, during operation, the electron beam is steered across the convex X-ray production target to maintain the pre-defined take-off angle with the travelling aperture. 6. A method for sweeping an X-ray beam across an object of inspection in two dimensions using a two-dimensional X-ray scanner wherein the X-ray scanner is configured to remain stationary with respect to the object of inspection, the method comprising:varying a direction of a beam of electrons relative to a target upon which the beam of electrons impinges; andcoupling X-rays generated at the target via an aperture that moves along a prescribed path as a function of time. 7. The method in accordance with claim 6, wherein coupling X-rays generated at the target may include coupling the X-rays via an intersection of a fixed slit and a moving slit. 8. The method in accordance with claim 6, wherein the target is a planar target block. 9. The method in accordance with claim 6, wherein the target is convex. 10. The method in accordance with claim 9, wherein the two-dimensional X ray scanner is configured to have a predefined take-off angle and wherein, during operation, the electron beam is steered across the convex X-ray production target to maintain the pre-defined take-off angle with the travelling aperture. 11. A two-dimensional X-ray scanner comprising:a beam steerer for steering an electron beam to impinge upon a target and thereby emit an X-ray beam; anda collimator comprising an aperture adapted for travel in an aperture travel path in order to rotate the electron beam impinging upon the target while maintaining an angular alignment with the target, wherein the two-dimensional X-ray scanner is configured to remain stationary relative to an object under inspection. 12. The two-dimensional X-ray scanner in accordance with claim 11, wherein the target is a planar target block. 13. The two-dimensional X-ray scanner in accordance with claim 11, wherein the target is convex. 14. The two-dimensional X-ray scanner in accordance with claim 13, wherein the two-dimensional X ray scanner is configured to have a predefined take-off angle and wherein, during operation, the electron beam is steered across the convex X-ray production target to maintain the pre-defined take-off angle with the travelling aperture. |
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claims | 1. A phase controller comprising:a hollow vacuum vessel to be a route for a soft X-ray;a reflection surface formed on an inside of the vacuum vessel and constituted by a transition metal having a core level absorption edge in the vicinity of a wavelength of the soft X-ray;a magnet for generating a magnetic field in a perpendicular direction to a longitudinal direction of the vacuum vessel in a position where the soft X-ray is to be reflected by the reflection surface; anda vacuum pump for bringing a vacuum state in the vacuum vessel,wherein the soft X-ray to be linearly polarized light is incident in the vacuum vessel set into the vacuum state by means of the vacuum pump and is reflected at least once over the reflection surface in a position where the magnetic field is applied so that the soft X-ray having a phase controlled is emitted from the vacuum vessel. 2. The phase controller according to claim 1, wherein the soft X-ray to be the linearly polarized light is incident in the vacuum vessel set into the vacuum state by means of the vacuum pump and is reflected at plural times over the reflection surface in the position where the magnetic field is applied so that the soft X-ray converted into circularly polarized light is emitted from the vacuum vessel. 3. The phase controller according to claim 2, further comprising a second reflection surface which is formed in a perpendicular direction to the reflection surface in a subsequent part to the reflection surface at an inside of the vacuum vessel, and is constituted by the same transition metal as the reflection surface,the soft X-ray to be the linearly polarized light being incident in the vacuum vessel set into the vacuum state by means of the vacuum pump and being reflected at plural times over the reflection surface in the position where the magnetic field is applied, and the soft X-ray being then reflected at plural times over the second reflection surface so that the soft X-ray converted into the circularly polarized light is emitted from the vacuum vessel. 4. The phase controller according to claim 3, wherein the reflection surface and the second reflection surface are formed to have an equal length, and a number of times of reflection over the reflection surface is equal to a number of times of reflection over the second reflection surface. |
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claims | 1. An X-ray diagnosis apparatus comprising:an X-ray tube configured to generate X-rays to an examination target having a particular directionality;an X-ray detecting unit configured to detect X-rays penetrated through the examination target;an X-ray collimating unit including a plurality of aperture blades for setting an irradiation region of the X-rays generated from the X-ray tube;a driving unit configured to rotationally move the X-ray tube and the X-ray detecting unit;an image data generating unit configured to generate image data by performing a reconstruction process based on projection data detected in correspondence with the rotational movements along a plurality of different imaging directions by the X-ray detecting unit;a region of interest setting unit configured to set up a region of interest on the examination target portion; andan X-ray aperture controlling unit configured to control the X-ray collimating unit so as to slide and turn the aperture blades in correspondence with the rotational movements, based on the set up data of the region of interest and the imaging direction,wherein in the X-ray collimating unit each aperture blade has a shape in which its thickness increases moving away from its inner edge closest to the region of interest, and each aperture blade includes a plurality of shield plates stacked on top of each other in a thickness direction, and each shield plate is movable in a planar direction perpendicular to the thickness direction. 2. The X-ray diagnosis apparatus according to claim 1, wherein the region of interest setting unit sets up a 3D region of interest on the examination target portion based on 3D image data or a plurality of 2D image data preliminarily acquired from the object. 3. The X-ray diagnosis apparatus according to claim 2, wherein the region of interest setting unit extracts the examination target portion or a medical treating device placed in the examination target portion based on the 3D image data or the plurality of 2D image data, and sets up the 3D region of interest based on a result of the extraction. 4. The X-ray diagnosis apparatus according to claim 2, further comprising an interest point designating unit configured to designate an interest point in the examination target portion indicated by the 3D image data or the plurality of 2D image data, or the medical treating device placed in the examination target portion; andwherein the region of interest setting unit sets up the 3D region of interest based on the interest point. 5. The X-ray diagnosis apparatus according to claim 1, wherein the X-ray aperture controlling unit slides each of the aperture blades in an approaching direction or a seceding direction to or from the center axis of X-ray beams based on the set up data of the region of interest, and turns them all about the center axis of the X-ray beams. 6. The X-ray diagnosis apparatus according to claim 2, wherein the X-ray movable control unit sets up X-ray imaging directions by rotating or moving the imaging system including the X-ray tube, the X-ray collimating unit and the X-ray detecting unit around the periphery of the examination target region, and slides and rotates the aperture blades based on the projected figure of the 3D region of interest in the set up imaging direction. 7. The X-ray diagnosis apparatus according to claim 1, wherein the image data generating unit generates difference projection data by performing a subtraction process between the mask projection data before administrating a contrast agent and the contrast projection data after administrating the contrast agent that are generated by the projection data generating unit in the X-ray imaging executed with successively renewing the imaging direction at the periphery of the examination target region, and generates at least one of 3D image data, MIP (maximum intensity projection) image data and MPR (multi-planar reconstruction) image data by performing a reconstruction process of the difference projection data acquired along each of the imaging directions. |
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description | The present application hereby claims priority under 35 U.S.C. §119 on German patent application numbers DE 10 2006 004 976.4 filed Feb. 1, 2006, and DE 10 2006 017 291.4 filed Apr. 12, 2006, the entire contents of each of which is hereby incorporated herein by reference. Embodiments of the invention generally relate to a focus/detector system of an X-ray apparatus for generating projective and tomographic phase contrast recordings. For example, they may relate to one including a beam source having a focus, a detector arrangement for detecting the X-radiation and a set of X-ray optical gratings, for determining the phase shift when the X-radiation passes through a subject. In computer tomography, tomographic recordings of a subject, in particular a patient, are generally made with the aid of absorption measurements of X-rays which pass through the subject, a radiation source generally being moved circularly or spirally around the subject and a detector on the opposite side from the radiation source, usually a multiline detector with a multiplicity of detector elements, measuring the absorption of the radiation when it passes through the subject. For tomographic image compilation, tomographic section images or volume data are reconstructed from the measured absorption data of all measured geometrical rays. Absorption differences in objects can be represented very well by these computer tomographic recordings, but regions with similar chemical composition, which naturally also have a similar absorptivity, can be represented only with insufficient detail. It is furthermore known that the effect of the phase shift when a ray passes through a subject is substantially stronger than the absorption effect of the matter through which the radiation has passed. Such phase shifts are measured in a known way by using two interferometric gratings. With respect to these interferometric measurement methods, reference is made for example to “X-ray phase imaging with a grating interferometer, T. Weitkamp et al., Aug. 8, 2005/Vol. 12, No. 16/OPTICS EXPRESS”. In this method, coherent X-radiation passes through a subject, the X-radiation having passed through is guided through a grating pair and the radiation intensity is measured immediately after the second grating. The first grating generates an interference pattern, which forms an image of a Moiré pattern with the aid of the second grating on the detector lying behind. If the second grating is displaced slightly, then this likewise causes a displacement of the Moiré pattern, i.e. a change of the local intensity in the detector lying behind, which can be determined relative to the displacement of the second grating. If the intensity change is plotted for each detector element of this grating, i.e. for each ray, as a function of the displacement distance of the second grating, then the phase shift of the respective ray can be determined. A problem, making it unsuitable for carrying out computer tomography of sizeable objects, is that this method requires a very small radiation source since coherent radiation is needed for forming the interference pattern. The method presented in the document cited above requires either a radiation source with an extremely small focus, so that there is a sufficient degree of spatial coherence in the radiation used. When using such a small focus, however, then a sufficient dose power for examining a sizeable object is in turn not available. It is nevertheless also possible to use monochromatically coherent radiation, for example synchrotron radiation as the radiation source, but this makes the CT system very expensive to construct so that widespread application is not possible. This problem can be circumvented by arranging a first absorption grating inside the focus/detector combination in the beam path, immediately after the focus. The alignment of the grating lines is in this case parallel to the grating lines of the interference grating which follows after the subject. The slits of the first grating generate a field of individually coherent rays with a particular energy, which is sufficient for generating the interference pattern known per se with the aid of the phase grating arranged behind the object in the beam direction. In this way, it is possible to use radiation sources which have extents that correspond to normal X-ray tubes in CT systems or transmitted-light X-ray systems so that, for example, even well-differentiated soft tissue tomographs can now be made in the field of general medical diagnosis with the aid of X-ray devices. A problem with this type of focus/detector combination is that on the one hand the analysis grating constitutes an additional sensitive component, which is cost-intensive to install and adjust. In a development according to a further aspect of at least one embodiment of the invention, a better dose utilization is intended to be achieved than is possible when using an absorption spectrum, in which half the applied dose is always lost. Furthermore, at least three measurements need to be carried out respectively with a slightly displaced analysis grating for each ray in space, so that it is possible to determine the phase shift of the X-radiation on the respective ray path through the subject. This entails an increased time and adjustment outlay for the measurements, which is intended to be reduced. In at least one embodiment of the invention, a focus/detector system is provided which allows simpler construction. The effect intended to be achieved according to a further aspect is to reduce the number of measurements required for determining the phase shift, or even to carry out just one measurement process on each ray in order to be able to generate projective or tomographic phase contrast recordings of a subject. According to a further aspect, better dose utilization is also intended to be achieved. The Inventors, in at least one embodiment of the invention, have discovered that instead of the previously used analysis grating, it is possible to use detector elements which comprise a multiplicity of scintillation strips that subdivide the individual detector element in the direction of the grating lines of an upstream phase grating, so that the previously required analysis grating can be obviated. The individual scintillation strips may furthermore be configured so that they alternately emit different frequency light, which is selectively measured. This entails a simply configured grouping of different scintillation strips inside a detector element, summation being carried out over the individual groups without great circuit technology outlay. The frequency-selective measurement and emission of all the scintillation light in a common space thus achieves selective summation of all light events at the different scintillator materials arranged in a strip shape. Depending on the number of groups which are formed and depending on the period with which the scintillation strips are arranged, i.e. depending on the fineness of the individual scintillation strips, it is therefore now possible to resolve an individual X-ray so that either the number of measurements with which a particular X-ray is sampled can be greatly reduced or, with a correspondingly high division of the scintillation strips, the average phase of the respectively considered X-ray can be determined directly with a single measurement of the scintillation strips dimensioned groupwise. In at least one embodiment, improved or even optimal dose utilization is also achieved by such a strip-shaped structure of the detector elements without “dead regions”, in which no measurement takes place. The full amount of the used dose to which the subject, in particular a patient, is exposed, is thus now used for the measurement and, unlike when an analysis grating is used, a part of the dose to which the patient has been exposed is not superfluously absorbed in the analysis grating. According to the basic concept of at least one embodiment of this invention, the Inventors propose a focus/detector system of an X-ray apparatus for generating projective or tomographic phase contrast recordings, comprising: a beam source having a focus and a focus-side source grating, which is arranged in the beam path and generates a field of ray-wise coherent X-rays, a grating/detector arrangement having a phase grating and grating lines arranged parallel to the source grating for generating an interference pattern, and a detector having a multiplicity of detector elements arranged flat for measuring the position-dependent radiation intensity behind the phase grating, wherein the detector elements are formed by a multiplicity of elongate scintillation strips, which are aligned parallel to the grating lines of the phase grating and have a small period, whose integer multiple corresponds to the average large period of the interference pattern which is formed by the phase grating. With respect to the grating/detector arrangement, it is proposed that this should be designed and arranged so that it satisfies the following geometrical conditions: p 2 = k × p s p 0 = p 2 × l d , p 1 = 2 × p 0 × p 2 p 0 + p 2 d = l × d ≡ l - d ≡ with d ≡ = 1 2 × ( p 1 2 4 λ ) , h 1 = λ 2 ( n - 1 ) . Here: p0=grating period of the source grating G0, p1=grating period of the phase grating G1, p2=large period of the scintillation strips SSi, average spacing of the interference lines after the phase grating, ps=small period of the scintillation strips SSi, distance from midline to midline of neighboring scintillation strips, d=distance from the phase grating G1 to the detector in fan beam geometry, d≡=distance from the phase grating G1 to the detector with parallel geometry, k =1, 2, 3, 4, 5, . . . , l=distance from the source grating G0 to the phase grating G1, λ=selected wavelength of the radiation, h1=bar height of the phase grating G1 in the beam direction, n=refractive index of the grating material of the phase grating. In a first simple alternative embodiment, the Inventors propose that the focus/detector system be configured so that precisely one scintillation strip, which alternates with a detector grating structure made of non-scintillating material, is arranged inside each large period. In terms of metrology, while having a simple structure, this essentially achieves the same effect as when using an analysis grating with the need for multiple measurements of the same ray in order to determine the existing phase shift. In order to achieve good stability and a large absorption difference between grating gaps and grating lines, it may be favorable for the detector grating structure to be made of metal. In another variant of the focus/detector system according to at least one embodiment of the invention, the Inventors propose to arrange precisely two scintillation strips made of different scintillation material, which generate light of different frequency f or different wavelength λi according to the relationship λi=c/fi, inside each large period, their sequence remaining the same over the entire detector element. This now makes it possible to utilize the detector surface optimally, since there are no longer any regions with radiation masking. In principle, this variant of the embodiment corresponds to a combination of two detector elements offset by half the large period, a detector material with different light emission properties respectively being used instead of the grating. Owing to the different frequencies of the light emissions, they can easily be measured separately from one another. Although in this variant of a focus/detector system the number of measurement processes required is reduced from at least three to at least two since two measurement values per measurement are obtained for two sample points to determine the phase profile, an offset between the measurements is nevertheless required. To this end, for example, the Inventors propose, in at least one embodiment, that a device/method be provided for offsetting the scintillation strips perpendicularly to the longitudinal direction of the scintillation strips in the detector, which can generate a defined offset of the order of the small period of the scintillation strips. As an alternative, a device/method may be provided for offsetting the detector elements or for offsetting the entire detector perpendicularly to the longitudinal direction of the scintillation strips in the detector. What is important for this offset and the devices selected therefor, is that it should be carried out in a defined way in the size range of the small period. Piezo elements, for example, are particularly suitable for this. In a further development of at least one embodiment of the inventive concept, in which a spatial offset is no longer categorically necessary, at least three scintillation strips made of different scintillation material, which generate light of different frequency, may be arranged inside each large period, here again their sequence remaining constant over the detector element. If this embodiment is used, then it is now possible for the spatial offset during the detection of the radiation intensity to be replaced by at least three frequency-selective measurements. Although a spatial offset of the scintillation strips is not categorically necessary here, for error reduction it may nevertheless be more favorable to increase the number of sample points and to this end provide a device/method for offsetting the scintillation strips perpendicularly to the longitudinal direction of the scintillation strips in the detector, which can generate a defined offset of the order of the small period of the scintillation strips. As an alternative, the detector elements or the detector may also be offset. As mentioned above, piezo elements in particular are suitable for this. The Inventors furthermore propose, in at least one embodiment, that a device/method be provided in the detector element which detect the light emissions of the scintillation strips of a detector element with different frequency separately according to frequency but summed over the entire detector element. Through such a configuration, it is possible to replace elaborate circuits for controlled groupwise combination of the scintillation strips of a detector element. The detector elements may furthermore be configured so that the scintillation strips emit their light with different frequencies at least partially into a mirrored space which adjoins frequency-selective light sinks, and each light sink includes a device/method for detecting the selected light. In a first alternative embodiment, the light sinks respectively consist of a filter with a downstream photodiode, the filters respectively being selective for precisely one of the emitted frequencies of the scintillation strips. Another alternative embodiment may reside in the light sinks being arranged in cascade fashion and respectively including a filter on the scintillator side with a photodiode, which limits the frequencies on one side so that a reduced number of frequencies is measured in the subsequent filter/photodiode set. With this variant the photodiodes are thus arranged in cascade behind filters, the filters increasingly cutting off the frequencies beginning on one side of the frequency spectrum. In this way, a photodiode on the side facing toward the input side of the light can detect the entire frequency spectrum and a respectively further restricted spectrum can be measured at each further photodiode, so that the intensity of individual spectral ranges can be determined. According to the basic concept of at least one embodiment of the invention, the Inventors also propose an X-ray system for generating projective phase contrast recordings, which is equipped with at least one of the focus/detector systems described above. Such focus/detector systems may also be used in conjunction with an X-ray C-arc system for generating projective and tomographic phase contrast recordings or an X-ray CT system for generating tomographic phase contrast recordings. Such X-ray systems may furthermore include a computation unit for controlling the detector and calculating the phase shift from a plurality of intensity measurements of the same ray. A computation and control unit is also proposed, which contains program code that carries out the method described below during operation. A storage medium of an X-ray system or for an X-ray system is likewise proposed, which contains program code that carries out this method during operation of an X-ray system. According to the basic concept of at least one embodiment of the invention, the Inventors furthermore propose a method for generating projective X-ray recordings of a subject, preferably of a patient, the method comprising: the subject is irradiated by a beam of rays, each ray in space being defined with respect to direction and extent by the focus-detector element connecting line and the extent of the detector element, the average phase shift of each ray is measured in that, for this ray, the intensity of the radiation is measured with the aid of the fine structured scintillation strips at scintillation strips arranged groupwise and offset with respect to one another or positioned offset from one another, phase contrast recordings, the pixel values of which represent the average phase shift per ray, are compiled from the measured average phase shifts. According to an example variant of an embodiment of the method, it is proposed that the various scintillation strips of a detector element emit light groupwise with different light frequencies during exposure and that this light be measured selectively with respect to the frequency but summed over the entire detector element. A spatial offset of the scintillation strips perpendicularly to the grating line direction may furthermore be induced between two measurements of the same ray. In this case, the spatial offset of the scintillation strips should be induced by an amount less than the small period of the scintillation strips. According to another variant of an embodiment, there are at least three different types of scintillation strips in a detector element, these are arranged uniformly alternating and one measurement for all emitted light frequencies is carried out per detector element and position, and the average phase shift of the measured X-ray is determined directly therefrom. It will be understood that if an element or layer is referred to as being “on”, “against”, “connected to”, or “coupled to” another element or layer, then it can be directly on, against, connected or coupled to the other element or layer, or intervening elements or layers may be present. In contrast, if an element is referred to as being “directly on”, “directly connected to”, or “directly coupled to” another element or layer, then there are no intervening elements or layers present. Like numbers refer to like elements throughout. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. Spatially relative terms, such as “beneath”, “below”, “lower”, “above”, “upper”, and the like, may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It will be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, term such as “below” can encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein are interpreted accordingly. Although the terms first, second, etc. may be used herein to describe various elements, components, regions, layers and/or sections, it should be understood that these elements, components, regions, layers and/or sections should not be limited by these terms. These terms are used only to distinguish one element, component, region, layer, or section from another region, layer, or section. Thus, a first element, component, region, layer, or section discussed below could be termed a second element, component, region, layer, or section without departing from the teachings of the present invention. The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of the present invention. As used herein, the singular forms “a”, “an”, and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “includes” and/or “including”, when used in this specification, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. In describing example embodiments illustrated in the drawings, specific terminology is employed for the sake of clarity. However, the disclosure of this patent specification is not intended to be limited to the specific terminology so selected and it is to be understood that each specific element includes all technical equivalents that operate in a similar manner. Referencing the drawings, wherein like reference numerals designate identical or corresponding parts throughout the several views, example embodiments of the present patent application are hereafter described. For better understanding of phase contrast measurement, a focus/detector system with a grating set G0 to G2 is shown in FIG. 1. Before the first grating G0, there is a focus F1 whose greatest extent is denoted by w. The first grating G0 has a grating line period p0 and a grating bar height h0. The gratings G1 and G2 are correspondingly also provided with a height hi and h2, respectively, and a period p1 and p2, respectively. In order for the phase measurement to function, it is necessary that the distance l between the gratings G0 and G1 and the distance d between the gratings G1 and G2 should be in a particular mutual ratio. Here, p 0 = p 2 × l d The distance of the detector D1 with its detector elements E1 to En from the last grating G2 is not essential. The height hi of the bars of the phase grating should be selected so that the following formula is satisfied according to the wavelengths in question, i.e. the relevant energy of the X-radiation, and in relation to the respective grating material: h 1 = λ 2 ( n - 1 ) Here, n denotes the refractive index of the grating material and λ denotes the wavelengths of the X-radiation, at which the phase shift is intended to be measured. This grating may advantageously be adjusted to an energy which corresponds to a characteristic line in the X-ray spectrum of the anode being used and at least a sufficient photon number should be available in this energy range. With the nowadays customary tungsten anodes, for example, the Kα line may be used. It is nevertheless also possible to use the Kβ line lying next to it. When other anode materials are selected, different energies and therefore different dimensioning of the phase grating will correspondingly be necessary. The height h2 of the analysis grating must be sufficient in order to generate effective absorption differences between the bars through which the X-radiation passes and the substantially free positions of the grating, in order to obtain a corresponding Moiré pattern on the rear side. The line orientation of the gratings G0 to G2 is regularly configured so that the grating lines of all three gratings extend mutually parallel. It is furthermore advantageous, but not necessary, that the grating lines should be oriented parallel or perpendicularly to the system axis S, in which case the gratings G0 to G2 will usually be designed to be flat and aligned perpendicularly to the midline between the focus and detector midpoints. In principle, it is nevertheless also possible to adapt the surface of the gratings to the ray profile of the ray cone so that, at any position, the gratings are intersected perpendicularly by the ray connection between the focus and the respective detector element, which entails a corresponding curvature of the gratings. FIG. 2 again shows the individually coherent radiation coming from the grating G0, which passes through the patient P, phase shift phenomena taking place after it passes through the patient P. When passing through the grating G1, an interference pattern is thereby generated, as represented by the gray shading, which with the aid of the grating G2 leads to different radiation intensities per detector element on the downstream detector D1 and its detector elements, a so-called Moiré pattern being formed there. If the detector element Ei, for example, is considered as a function of an offset xG of the analysis grating G2 and the intensity I(Ei(xG)) is plotted as a function of the offset xG against the intensity I, then a sinusoidal rise and fall of the intensity I at this detector element Ei is obtained. If these measured radiation intensities I are plotted for each detector element Ei or Ej as a function of the offset xG, then the function I(Ei(xG)) or I(Ej(xG)) can be approximated for the various detector elements, which in the end form the geometrical X-ray between the focus and the respective detector element. The phase shift φ relative to one another can be determined for each detector element from the functions. The following applies: φ = 2 π n v λ ,where v corresponds to the size of a voxel or pixel in the object examined, n is its refractive index and λ represents the wavelength of the X-radiation. For each ray in space, the phase shift per ray can therefore be determined by at least three measurements with a respectively offset analysis grating, from which either the pixel values of a projective recording can be calculated directly in the case of projective X-ray recordings, or projections whose pixel values correspond to the phase shift are compiled in the case of a CT examination, so that with the aid of reconstruction methods known per se it is possible to calculate therefrom which volume element in the subject is to be ascribed to which component of the measured phase shift. Either section images or volume data are thus calculated therefrom, which reflect the local effect of the examined object in respect of the phase shift of X-radiation. Since even minor differences exert a strong effect on the phase shift in this context, very detailed and high-contrast volume data can be obtained from materials which are relatively similar per se, in particular soft tissue. The previously described variant of detecting phase shifts of the X-rays which pass through a subject, with the aid of a multiply offset analysis grating and measuring the radiation intensity on a detector element behind the analysis grating, has the disadvantage that at least three measurements of each X-ray have to be carried out with a respectively displaced analysis grating. This makes the scanning of the subject relatively slow, the dosage also being increased. There is also the problem that a part of the radiation is lost from the detection owing to the analysis grating being used, since it is absorbed in the grating. At least one embodiment of the invention therefore proposes to obviate such an analysis grating and instead to structure the detector elements, which are arranged following the phase grating, so that at least no dose loss occurs in the measurement, and advantageously to select a division such that the phase shift in the relevant ray can be determined by a single measurement. Such an arrangement is schematically shown in a 3D representation of a focus/detector system of a computer tomograph in FIG. 3. This shows a focus F1 in whose beam path a source grating G0 is arranged and on the detector side there is a phase grating which generates the interference phenomena described above, which are measured by the subsequent detector so that each individual detector element can measure the phase shift, or more precisely the average phase shift, of the radiation over this detector element. In the representation shown, a detector D1 which is designed as a multiline detector is represented on the detector side, each line containing a multiplicity of detector elements and each detector element being preceded by a grating structure of the phase grating G1. This combination of a grating and detector element is shown on an enlarged scale in FIG. 4. Here, the detector element is represented as being structured, consisting of a multiplicity of scintillation strips SS1 to SS18 which are oriented in terms of their alignment parallel to the grating lines of the phase grating G1. It should be pointed out that the division as shown here is merely a schematic representation, which is intended to show the basic principle of the division, the dimensions in practice being fundamentally different therefrom. In practice, the size of such a detection element is in the range of from 100 to 1000 μm. The period p2, of the order of which the extent of the scintillation strips must be, is generally about 2 μm so that the individual scintillation strips, if they are divided into two divisions, are approximately one μm. FIG. 5 further illustrates the basic principle of measuring the phase shift with the aid of an analysis grating G2. This representation schematically shows the flux of the X-ray photons Φph over the x axis behind the phase grating at a spacing of one Talbot distance, the profile of the photon flux Φph(x) being plotted against the x axis. The x axis in this case extends perpendicularly to the grating lines. The analysis grating G2 is subsequently shown, which comprises a period p2 and absorbs the photons in its bars so that only at the free positions can the photons pass through downward and finally strike the detector element Ei lying behind, where their intensity is measured. If the grating G2 is now displaced slightly in the direction of the x axis, then a strong intensity variation of the measured radiation intensity Iph occurs on the detector element lying behind, which may be plotted against the length of the displacement of the grating. The phase φ can be determined for the respective detector element from the curve of the radiation intensity as a function of the offset xG of the analysis grating G2. According to at least one embodiment of the invention, the analysis grating can now be replaced by imparting a grating-like structure to the detector element, so that for the detection of radiation there are periodically arranged strip-shaped regions which groupwise provide information about the radiation incident there. In the simplest variant, this involves a single group of strips SSi which alternate with grating strips without detection GSi. The period of the replaced analysis grating, tuned to the respective energy of the grating arrangement, is selected as the period p2 with which these strips are arranged. Here, the width of the scintillation strips may advantageously be selected so that it is equal to half the period of the corresponding analysis grating. Such a situation of a detector element Ei is represented in FIG. 6. The photon flux Φph due to the interference phenomenon, which is caused by the phase grating, is here again firstly represented at the top against the x axis. This position-dependent photon flux Φph(X), also represented using the arrows denoted by γ, strikes the detector element with differing intensity and is periodically converted into light with the wavelength λ1 by the multiplicity of scintillation strips SS1 to SS6. This light shines into a space 17, which as far as possible is mirrored on all sides, where it is measured as a whole by a photodiode 12. Since this embodiment with respect to the division of detecting and non-detecting regions essentially constitutes no difference from the analysis grating described above, here again it is necessary to measure changes in the radiation intensity relative to the spatial offset of the scintillation strips. In this example, this is done by offsetting the detection element Ei as a whole relative to the detector housing 14 with the aid of two piezo elements 13.1 and 13.2, and measuring the radiation intensity on the scintillation strips at each offset. These intensity measurements Iph(xG=0), Iph(xG=¼p2), Iph(xG= 2/4p2) and Iph(xG=¾p2), with a relative offset xG of p2/4, are plotted at the bottom in FIG. 6. From this, it is possible to approximate a sine curve and calculate the phase shift. It should also be noted that measurement of three sample points is sufficient in principle, although more sample points can be advantageous for reducing the noise and compensating for other measurement errors. It should furthermore be pointed out that a detection element as represented in FIG. 6 may, for example, be produced by filling the free positions of a grating generated by etching technology, which produces the grating strips GSi, with a polymer that contains nanoparticles of scintillation material. FIG. 7 shows an embodiment of such a detector element Ei which is somewhat more realistic in respect of the dimensioning. The structure of the grating having the gaps filled with scintillation material is also represented in a detail enlargement. An improved embodiment of a detector element Ei according to an embodiment of the invention is schematically presented in FIG. 8. This differs from FIG. 6 in that the scintillation strips are not supported between one another by a grating structure insensitive to radiation, rather it consists exclusively of scintillation strips constructed in layer fashion. The small period pss of the scintillation strips corresponds here to half the period p2 of a corresponding analysis grating, also referred to as the large period. For example, this involves a polymer material which is alternately filled with different nanoparticles that emit different light with the wavelengths λ1 and λ2 in a space 17 when exposed, the layered structure being obtained for example by successive laser irradiation of different liquid polymer-nanoparticle mixtures—similarly as prototype construction. The light emitted during the X-ray exposure is radiated into a mirrored space which through two frequency-selective filters 16.1, 16.2 of two photodiodes 12.1 and 12.2, which respectively present a measurement path A or B. In this way, light of the wavelength λ1 is received only via the measurement path A and light of the wavelength λ2 is received only via the measurement path B. Correspondingly, the radiation intensities at the even-numbered or odd-numbered scintillation strips may be determined simultaneously by a measurement. In this way, by considering the dose measured respectively via the measurement path A or the measurement path B, it is possible to measure the intensity change which would result if an analysis grating—corresponding to FIG. 5—would be displaced by one half period. If two further measurements A′, B′ are now carried out with an offset of p2/4, then four measurement values A, B, A′ and B′ are available at four sample points. The average phase φ of the X-ray of this detector element can be calculated directly therefrom. This is represented at the bottom of FIG. 8. In this special embodiment, the offset of the scintillation strips is generated by electrically controlling a piezo element 13 on one side of the layered scintillation strips and compensating by a spring element 15 on the other side. A substantial advantage of this embodiment variant is that no radiation dose which has passed through a patient is lost, since the entire surface of the detection elements is used for determining the phase shift. While it is still necessary to carry out at least two measurements with respectively offset scintillation strips in the alternative embodiment of FIG. 8, this is not necessary in a further improved embodiment of the inventive detection system in FIG. 9. Here, the phase of the X-radiation detected by a detection element can be determined via a single measurement. Similarly as FIG. 8, FIG. 9 shows a detection arrangement with the radiation arriving from above on a detection element Ei, the radiation generating small-space interference phenomena owing to the upstream energy-specific grating arrangement, which leads to the periodically varying photon flux as described above. Realistically, this variation is not strictly periodic but, because of the phase shift occurring with differing strength, is subject to spatial fluctuations from which the phase shift of a ray can reciprocally be determined. In the present example, the subdivision of the detection element into scintillation strips is configured so that the individual scintillation strips merely have a width or small period pss of ¼ of the period p2 of a corresponding analysis grating. Here, four differently doped scintillation materials are used, which respectively generate light at different frequencies and wavelengths because of the different doping. The four different scintillation strips are repeatedly sequenced with the same period and in the same order. Each group of scintillation strips with the same doping emits light of the same wavelength, and the four groups emit light of four different wavelengths λ1, λ2, λ3, and λ4. Selected according to wavelength with the aid of the filters 16.1 to 16.4, this light is measured at four different photodiodes 12.1 to 12.4, corresponding to the measurement paths A to D, and therefore represents a measure of the dose arriving on the various scintillation strips. In this way, the scintillation strips are thus interconnected so that every fourth strip uses the same measurement path. If a measurement is now carried out with such a detector arrangement at a particular position, i.e. for a particular X-ray, then the phase-corresponding intensity can respectively be read from the intensities measured via the measurement paths A, B, C and D and the phase of the X-radiation which strikes this detector element can be determined directly from these four measurements. The evaluation of these four measurement values A, B, C, D is represented at the bottom in this figure. It should furthermore be noted here that this measurement does not correspond for instance to a phase determination of the X-radiation in the region of an individual scintillation strip, rather it represents averaging over the entire surface of the detection element. In this alternative embodiment as well, it is particularly advantageous that the entire dose used for the measurement which irradiates the subject, in particular a patient, is employed for the evaluation and scarcely no dose losses occur. The essence of the two alternative embodiments presented last is thus that a detector element is divided into a multiplicity of scintillation strips, which are read groupwise in respect of the measured X-ray intensity, the division needing to be carried out so that on the one hand it matches the period p2 of a corresponding analysis grating, but at the same time it comprises at least two and preferably at least three scintillation strips per period so that each of the groups of scintillation strips is represented once per period. This type of division thus makes it possible to fit two, three, four, five or more scintillation strips within one period and sequence this division repeatedly in a direction perpendicular to the alignment of the scintillation strips, so that the number of measurement groups corresponds to the number of scintillation strips per period p2. FIG. 10 represents a complete computer CT system for carrying out the method according to an embodiment of the invention. It shows the CT system 1 which comprises a first focus/detector system with an X-ray tube 2 and a detector 3 lying opposite, which is arranged on a gantry (not represented in detail) in a gantry housing 6. An X-ray optical grating system according to the invention is arranged in the beam path of the first focus/detector system 2, 3 so that the patient 7, who lies on a patient support 8 displaceable along the system axis 9, can be displaced into the beam path of the first focus/detector system and scanned there. The CT system is controlled by a computation and control unit 10 in which programs Prg1 to Prgn are stored in a memory 11, which carry out the method according to an embodiment of the invention as described above and reconstruct corresponding tomographic images from the measured ray-dependent phase shifts. Optionally, instead of the single focus/detector system, a second focus/detector system may be arranged in the gantry housing. This is indicated in FIG. 10 by the X-ray tube 4 shown in dashes and the detector 5 represented in dashes. Moreover, it should also be pointed out that the focus/detector systems as presented are not only capable of measuring phase shifts of the X-radiation, rather they are furthermore suitable for conventional measurement of the radiation absorption and reconstruction of corresponding absorption recordings. Optionally, combined absorption and phase contrast recordings may even be generated. It should furthermore be pointed out that in practical embodiment, the gaps between the grating lines in the source gratings used may be filled with a highly absorbent material to improve the contrast. For example, gold may be used for this. In principle, the source gratings should be configured so that they achieve a contrast factor of at least e−1. It is to be understood that the features of the invention as mentioned above may be used not only in the combination respectively indicated, but also in other combinations or in isolation, without departing from the scope of the present invention. Example embodiments being thus described, it will be obvious that the same may be varied in many ways. Such variations are not to be regarded as a departure from the spirit and scope of the present invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims. |
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description | The present invention provides a process for fabricating a robust x-ray mask tool. In particular, the present invention provides a process for fabricating an x-ray mask tool capable of replicating features having lateral dimension of less then 1 micron. Such a mask has great utility for production of metal and ceramic microparts by the well-known LIGA processes since current technology is limited to providing part molds having lateral features sizes greater then 1 micron. Lithographic masks are available with features less then 1 micron and are fabricated typically in 1000 xc3x85 chromium supported on glass slips. However, in order for a mask to effectively stop the high energy, high fluence synchrotron radiation used to prepare molds for LIGA microparts, the image-stop layer, opaque to x-rays, would need to be several microns thick. Typically, LIGA x-ray masks have a gold image-stop layer at least 8 microns thick. Masks this thick are very difficult to produce by conventional means if this layer includes features of less than 1 micron across. The instant invention employs a combination of processes to produce a patterned mask which overcomes this limitation. A mask pattern is replicated in a thin layer of photoresist applied to a silicon substrate, developed to expose the pattern on the surface of the silicon, the exposed areas deeply etched by a reactive plasma technique to provide a series of trenches on the silicon surface, the trenches filled by plating a metal opaque to high energy x-rays, the plated surface planarized, and the substrate thinned to provide the desired mask. General Description This invention describes a lithographic mask having x-ray attenuating structures embedded in an essentially x-ray transparent support media. Furthermore, the invention describes a lithographic mask having features which can have lateral dimensions much smaller than 1 micron across, whether those features are the x-ray blocking structures themselves or the separation spaces between such structures. The process begins with a standard silicon substrate. A layer of a polymer photoresist is placed onto a top surface of substrate such that the layer is no more than 1-2 microns thick. Any technique for applying such layers may be used, including dipping, spraying, spinning or vapor depositing, and either organic or inorganic resists may be used. The method of application and composition of the resist is not critical except for the need for providing a coating layer of less than 2 microns. The resist layer is baked, or otherwise cured, and the desired image pattern rendered onto the layer top surface by using any of a number of conventional lithographic processes, such as by a direct contact transmission mask. It is also possible to create the desired pattern by imaging the reflection of a non-contact mask through camera optics onto the resist surface, or by directly xe2x80x9cwritingxe2x80x9d the image by using a programmable e-beam writer. Important to the proper operation of the invention is the ability of the exposing xe2x80x9clightxe2x80x9d used to penetrate the full depth the resist since it is known that as xe2x80x9clightxe2x80x9d wavelengths decrease toward the hard UV ( less than 190 nm) their penetrating power is significantly reduced necessitating thinner resist layers. Being able to fully penetrate the resist layer will allow the user to achieve the very small lateral dimensions desired. Use of a thin resist layer and a broadband light source helps to satisfy this requirement. Since the resist coating will act as an etchant barrier during subsequent processing, the amount of protection needed will be determined by the processing necessary to provide the desired structure. Different combinations of resist compounds provide additional options. In the present case a thin polymer resist is placed directly onto a silicon substrate, cured, masked and exposed to broadband light. Such a structure can provide about a 50-to-1 processing-protection ratio; a sufficiently robust etchant barrier to allow etching deep, narrow, channel structures in the silicon substrate. A composite resist comprising a thin layer of conventional polymer resist may be applied over a thin silicon dioxide layer grown onto the silicon substrate, where UV light is used to create the image pattern. Such resists permit direct transfer of the image into a silicon dioxide (glass) xe2x80x9chardxe2x80x9d resist which provides a processing protection ratio of 200-to-1 which is about equivalent to the former resist barrier since the glass resist layer is much thinner, typically about 1000 xc3x85. After rendering the image of the mask into the resist, the resist layer is chemically xe2x80x9cdevelopedxe2x80x9d and the exposed areas of the resist either removed or retained, depending upon the specific resist chemistry used. Following the development of the resist, the patterned substrate is exposed to a series of anisotropic reactive etching steps such as those set forth in the so-called BOSCH process described in U.S. patent Ser. No. 5,501,893, herein incorporated by reference in its entirety. In this, or similar anisotropic processes, the top surface of the silicon substrate is protected by the retained resist layer. This first etching step is followed by a first polymerization step which coats the walls, edges and bases of the etched recesses in the silicon substrate. Plasma reactor parameters and etching times are adjusted and limited to avoid excessive damage to the resist layer and the process proceeds in this manner, alternating between etching and coating steps, until a etch depth of between 10 to 30 microns is achieved. In particular, in order for an mask to effectively stop the high energy synchrotron radiation used to prepare molds for LIGA microparts, a thickness of at least 8 microns of gold is necessary. Etch depths of at least this dimension are therefore critical to the success of this invention. After etching the silicon substrate, the remaining resist is stripped away and the substrate cleaned, after which a xe2x80x9cseedxe2x80x9d layer of 0.025 microns of chromium followed by 0.08 microns of gold is vapor deposited onto the entire surface. Alternately, this layer may be omitted if the substrate used is a doped, highly conductive, form of silicon. Where the more conventional undoped silicon substrate is used, a second, thicker gold layer is deposited over the xe2x80x9cseedxe2x80x9d layer so as to completely fill and cover the etched recesses. Coating is typically done by electroplating or by electroless deposition onto the xe2x80x9cseedxe2x80x9d layer but may be done by any method providing the applied layer is uniform in composition and structure and provides a continuous, condensed layer. The thick x-ray blocking layer may be laid down, for instance, by continuing the vapor deposition of the xe2x80x9cseedxe2x80x9d layer, by plasma spraying, or by epitaxy deposition. Time and cost, however, favor a plating process. Once plated, the incipient mask is planarized by lapping the top surface of the substrate in order to remove the metal layers from this surface leaving the surface flat, and essentially free of the plated metal. What remains is a silicon substrate with a fine metal structure embedded into the thickness of the substrate forming an imaging pattern comprising a gold (or other similar x-ray opaque material) xe2x80x9cribbonxe2x80x9d structures extending to a depth of 10 microns or more wherein the widths of the structures may be less than 1 micron is provided In a final step, the back surface of the silicon substrate, the side opposite the planarized surface, is etched away in a region underneath the plated gold pattern to a depth sufficient to reduce the total thickness of the silicon in this region to below about 100 microns. This is done because it is known that x-ray radiation at energy levels of about 10 KeV is not significantly attenuated by passing through silicon of these thicknesses. Specific Description An embodiment of the steps of the invention are described with reference to FIGS. 1 through 5. As required, detailed embodiments of the present invention are disclosed herein. However, it is to be understood that the disclosed embodiments are merely exemplary of the present invention which may be embodied in various systems. Therefore, specific details disclosed herein are not to be interpreted as limiting, but rather as a basis for the claims and as a representative basis for teaching one skilled in the art to variously practice the present invention. Referring to FIG. 1A, the process begins with a silicon substrate or wafer 10. This substrate can, generally, having any useful shape and thickness but should of necessity be a thin wafer having parallel top and bottom surfaces 11 and 12. In particular the present invention is most easily implemented by using an industry standard 100 mmØxc3x970.67 mm thick wafer. In FIG. 1B a liquid photoresist film 20 (herein SRP 3612 Novolak) is applied by spin coating to a thickness of about 1.8 microns or less, and then baked at a temperature of 95xc2x0 C. for about 90 seconds in order to at least partially cure the resist layer. The particular resist thickness is chosen so as to balance the need for providing a thick enough layer to protect the unexposed portions of the silicon substrate from the effects of the later ion etch phase against the desire to fully expose the full thickness of the resist during the light exposure phase. In a next step, shown in FIG. 1C, a standard direct-contact lithographic mask 13, herein embodying a negative trace image of the desired pattern, is placed directly on the surface of the of resist layer 20 (FIG. 1C intentionally shows mask 13 above this surface for clarity sake only). In FIG. 1D the exposed portions 14, of the resist layer 20 are subjected to a source (not shown) of broadband light, 15. Mask 13, is itself formed by depositing a 1000 xc3x85-5000 xc3x85 thick layer of chromium, or similar material, into a glass support slip and comprises a plurality of lines and other structures and features, and separations between features, some of which have minimum lateral dimensions (dimensions in the plane of the mask, perpendicular to separate pattern features) of less than 1 micron. The resist exposure source used herein was a high pressure mercury-vapor lamp emitting light over a spectral range of about 365 nm to 450 nm and providing a dose of approximately 80 millijoules/cm2 measured at a wavelength of 365 nm. In the next step in the process, illustrated in FIG. 1E, the photoresist is chemically xe2x80x9cdevelopedxe2x80x9d and the exposed portions, 14, of photoresist layer 20 are removed. What remains are the unexposed portions, 22, of the resist in an inverse image of the mask pattern wherein this inverse image comprises xe2x80x9cclearxe2x80x9d areas 23 of exposed silicon. Again, this step is performed using standard and well-known lithographic processes. It should be noted that the choice of a positive or negative image mask depends largely on the nature of the photoresist used, i.e., depending upon whether or not the exposed portion of the photoresist is removed or left intact after the resist has been developed. Either approach is possible, although, depending on the nature of the desired pattern, one is usually more preferred than the other. After cleaning and drying, the patterned substrate 10 is subjected to an anisotropic reactive plasma etching process, shown in FIG. 2A, such as the BOSCH or other similar etch-and-coat technique, wherein the exposed areas 23 of the silicon substrate 10 are etch to a depth d which is substantially greater than the width w of etched channels 25. This step provides the very high aspect ratio etched pattern shown in FIGS. 2B. As noted supra. the BOSCH process is a two step etch-and-coat process wherein the intervening coating step comprises coating the exposed silicon with a thin layer of a polymer film 24 which protects the walls and edges of the etched channel but is quickly destroyed on those surfaces which directly face the bombardment of the reactive plasma 26 shown in FIG. 2A. This action has the effect of etching channels or trenches in the exposed silicon which have a substantially uniform width and substantially parallel walls. The process continues until the desired etch depth d has been achieved. In the case of the present invention the desired depth was 30 microns but any depth, which achieves the stated intent of creating an x-ray blocking mask, is possible. After etching the silicon wafer 10 to the desired depth, the remaining resist layer 22 is removed and the part cleaned leaving substrate 10 with a pattern of etched surfaces 27 across top surface 11 of the wafer. The entire surface is subsequently covered with a thin electrically conductive metal film 30, as shown in FIG. 3B, in preparation for a much heavier coating. The chosen process for applying the first thin coating of FIG. 3B is a thermal evaporation or particle vapor deposition (PVD) process, although any other coating process which would provide a thin, continuous layer of conductive material would be equally effective. However, any such processes must be able to coat both the walls 28 and the bases 29 of the etched channels 25. Such methods could include, but are not limited to, sputtering and chemical vapor deposition or spraying coating methods, and only require that the coating process provide a continuous, adherent, and conductive layer. As disclosed herein, the film 30 is about a 250 xc3x85 (0.025 microns) layer of chromium with an overlaying layer of about 800 xc3x85 (0.08 microns) of gold. Any similar metal or combination of metals would be useful including most of the metals in the Transition series of metal listed in New IUPAC Group Numbers 4-12 of the Period Table of elements, alloys thereof, and certain of the metals of Groups 13 and 14, such as aluminum and tin. Film 30 is necessary to enable adherence of a final, thicker metal layer 31 which is deposited in a subsequent step, shown in FIG. 3C. In the present invention, layer 31 is also gold but as before, could be any similar metal selected from the list supplied above, providing that the etch depth d of the mask is adjusted to provide for a sufficiently thick layer of metal to effectively block or substantially attenuate the aforementioned synchrotron flux while remaining below a 100 microns thickness limit known to be about the limit at which silicon is no longer xe2x80x9ctransparentxe2x80x9d to such radiation but will itself begin to attenuate the x-ray beam and thus will begin to degrade to transmission and resolving power of the x-ray mask. Following the final step of depositing the thick x-ray blocking layer 31, the mask assembly is planarized, as shown in FIG. 4, to remove metal from across top surface 11 of supporting silicon substrate 10, and to provide a planarized surface 32. Planarizing is typically performed by lapping the top surface until the surface of the silicon is reached leaving only the embedded metal pattern 33 exposed. This is done to remove the xe2x80x9coverburdenxe2x80x9d x-ray blocking metal layer on the top surface of the substrate leaving only the metal deposited in etched channels 25. Planarized surface 32 is also intended to be as flat and smooth as possible since it is the surface which will lay against the surface of the material onto which the synchrotron radiation is to be illuminated. A final thinning step, illustrated in FIG. 5, is intended to reduce the thickness of silicon substrate 10 across a region 34 beneath the embedded metal pattern 33. Thinning is performed on the back side 12 of wafer 10 using a standard blanket etching techniques until the thickness of silicon everywhere underneath region 34 of the metal pattern 33 is reduced to about less than 100 microns. As explained above it is known that silicon is transparent or nearly transparent to synchrotron radiation of 10 KeV at thicknesses below about 100 microns. Finally, because a plurality of metal patterns would be embedded on each silicon wafer, the thinning step is most easily performed by reducing the thickness of the wafer across the entire surface under which such patterns have been created. Doing so however, will inevitably weaken the wafer to the point where it cannot be manually handled. In such cases, unetched areas in the form of struts spanning the diameter of the wafer are allowed to remain as strengthening members. At this point, the x-ray mask is complete. By implementing these steps, a mask having blocking structures with lateral dimensions of less than 1 micron are achievable. The mask is utilized by placing its planarized surface 32 directly onto the surface of the article which is to be exposed to the synchrotron radiation, and illuminating this assembly with the radiation. The foregoing description of the invention has been presented for purposes of illustration and description and is not intended to be exhaustive or to limit the invention to the precise form disclosed. Many modifications and variations are possible in light of the above teaching. The embodiments were chosen and described to best explain the principles of the invention and its practical application to thereby enable others skilled in the art to best use the invention in various embodiments and with various modifications suited to the particular use contemplated. The scope of the invention is to be defined by the following claims. |
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description | In applications where x-rays from a remote point source are being studied, the fall-off in photon flux with the square of the distance can seriously limit the sensitivity of the measurement. Two generic x-ray optic designs disclosed herein address the sensitivity problem. One design concept is based upon single or multiple reflection at grazing incidence from surfaces formed from nested cylindrical, conical, cylindrical spiral or conical spiral foils as shown in FIGS. 1 and 2. In these figures, a scanning electron microscope 10 generates a divergent beam of x-rays 12. The x-rays 12 impinge upon a single reflection cylindrical or cylindrical spiral foil concentrator 14 and are focused on a spectrometer 16. In FIG. 2 the diverging beam of x-rays 12 encounters a nested or multiple reflection conical or conical spiral foil concentrator 18 which similarly focuses the x-rays 12 on the spectrometer 16. A second design concept is shown in FIGS. 3 and 4 and takes advantage of multiple reflections in glass capillary bundles. In FIG. 3, the diverging beam of x-rays 12 passes through point-to-point capillary bundle 20 which focuses the x-rays 12 onto the spectrometer 16. In FIG. 4, multiple reflection point-to parallel, parallel-to-point capillary bundles 22 similarly focused the beam 12 onto the spectrometer 16. FIG. 4 also represents a point-to-parallel single or multiple reflection conical or conical spiral concentrator followed by a single or multiple reflection parallel-to-point conical or conical spiral concentrator. Details of the embodiments shown in FIGS. 1-4 will be described hereinbelow and the performance from experiments are presented. Each of the x-ray optics embodiments shown in FIGS. 1-4 is compact and capable of providing a significant enhancement in the solid angle of collection. These embodiments are particularly adapted to laboratory astrophysics and x-ray microanalysis applications in which they can significantly improve coupling of a cryostat which contains a high resolution x-ray microcalorimeter to a plasma machine or the scanning electron microscope 10. It will be readily apparent to those skilled in the art that the technology disclosed herein is applicable to space-borne astrophysical applications. Because the concentrators 14, 18, 20 and 22 focus a diverging beam, acceptable intensities are presented at the spectrometer 16. Two embodiments of the foil concentrators of the invention are shown in FIGS. 5 and 6. In FIG. 5, a cylindrical or conical concentrator 24 includes nested concentric cylinders or cones 26, 28, 30, etc. The concentric cylinders or cones are formed from a thin ribbon of a gold-coated plastic. The nested cylinders or cones 26, 28, 30, . . . , may also be made of glass, aluminum foil, silicon or germanium. A spiral concentrator 32 shown in FIG. 6 is formed of a long single ribbon 34 that is wound into a spiral. The ribbon 34 may be gold-coated plastic, aluminum foil or quartz ribbon. Suitable plastic materials for the embodiments in FIGS. 5 and 6 include polyester, polyimide, kapton, melinex, hostaphan, apilcal, mylar or any suitably smooth, flexible material. A particularly preferred plastic is available from the Eastman Kodak Company under the designation ESTAR(trademark). Such plastic foil may range from 0.004 to 0.015 inches thick, for example. The plastic material is coated with a thin layer of metal, preferably a high Z metal such as nickel, gold or iridium and may be coated with multilayers. A suitable thickness for the metal coating is approximately 800 xc3x85. Evaporation or sputtering is a suitable technology for applying the metal coating to the plastic ribbon material 34. The embodiments of FIGS. 5 and 6 may be configured for single reflection as illustrated in FIG. 1 or for multiple reflections as illustrated in FIG. 2. The embodiments shown in FIGS. 5 and 6 both use a point-to-point geometry to obtain significant gain and solid angle in the energy band of 0.1 keV to 10 keV. The gain depends upon the x-ray reflectivity, focal distance, the width of the ribbon material and the number of windings of the spiral or the number of nested cylinders. The x-ray reflectivity of the concentrators 24 and 32 can be improved by depositing multilayers of Wxe2x80x94C, Coxe2x80x94C, or Nixe2x80x94C for example, on the uncoated or metal-coated plastic which allow the designs to include larger grazing angles. An embodiment of the cylindrical spiral concentrator 32 has been built and tested in a microanalysis application at the Smithsonian Astrophysical Observatory in Cambridge, Mass. in which the distance between an x-ray source (scanning electron microscope, SEM) and an energy dispersive detector (lithium-drifted silicon detector and/or x-ray microcalorimeter) was approximately two meters. The constructed embodiment used single reflection in a point-to-point geometry. For the spiral concentrator 32 the ribbon was wound with a pitch of 0.05 inches and had 19 windings within an entrance aperture with diameter of 50 mm. For the cylindrical concentrator 24, the ribbon would be cut into 20 lengths to form concentric cylinders. FIG. 7 shows the results of a ray tracing computer program which simulated the shape of images produced by the cylindrical concentrator 24 with a ribbon width of 25 mm and focal length of 1.5 m. FIG. 8 depicts a simulated image expected from the cylindrical spiral concentrator 32. In contrast to the cylindrical geometry, the spiral optic 32 forms an annular image as shown in FIG. 8 because the ray that connects the center of the spiral to the reflecting surface of the ribbon is not the same ray that describes the normal vector at the ribbon surface. This relationship is schematically illustrated in FIGS. 9 and 10 in which the reflection geometry of the cylindrical optics of FIG. 5 and the spiral optics of FIG. 6 are compared. With reference to FIG. 11, the cylindrical spiral concentrator 32 comprises front and back discs 40 and 42 supported in spaced apart relation by a central hub 44. In this embodiment, each disc 40 and 42 has eight spokes which extend radially from the central hub 44. Holes (not shown) are drilled into the spokes to hold thin stainless steel pins 46. The pins 46 locate and support the gold-plated plastic ribbon 34 (not shown in FIG. 11). One end of the ribbon 34 is clamped to the central hub 44 and the other end is clamped to one of the outer ones of the support pins 46. FIG. 12 shows an assembled concentrator 32. It should be noted that the ribbon 34 may be supported by grooves machined into the radial spokes of the front and back discs 40 and 42. The structure supporting the ribbon 34 may be made of metal, plastic or a composite material. Suitable metals are aluminum, beryllium, stainless steel, titanium or tungsten. The spiral optic 32 shown in FIG. 12 was evaluated in the context of an x-ray microanalysis application. The optic (or concentrator) 32 was mounted on a kinematic base which was attached to a stage with five degrees of freedomxe2x80x94three translational and two rotational axes. The stage was located midway (52 inches) between the axis of a scanning electron microscope and a microchannel plate detector or lithium-drifted silicon detector. These instruments were used to measure the image characteristics and spectral transmission properties, respectively. FIG. 13 shows an image measured with an imaging detector at an energy of 1.5 keV. The annular image structure predicted by the simulation of FIG. 8 is evident. The spectral count rates obtained with and without the x-ray concentrator 32 are shown in FIG. 14. A lithium-drifted silicon detector was used for these measurements. The three peaks are Cu Lxcex1 at 930 eV, Cu Lxcex1 at 8.04 keV and Cu Lxcex2 at 8.9 keV, respectively. The ratio of the intensities recorded with and without the telescope is a measure of the gain provided by the x-ray optic 32 and is shown in FIG. 15. In this particular case, a 1 mm diameter aperture was placed over the detector to mimic the size of a smaller x-ray detector such as a microcalorimeter. The gain of approximately 200 below 2 keV means that, at a distance of 2 meters from the source, the telescope can provide an x-ray intensity that is equivalent to placing the detector fourteen times closer to the source (14 cm). An example of a high resolution microanalysis spectrum taken with a cryogenic microcalorimeter instead of a lithium-drifted silicon detector is shown in FIG. 16. Monolithic polycapillary glass optics have been adapted by others for many laboratory applications including microflorescence analysis and protein crystallography [references]. These tapered glass optics have made it possible to intercept x-rays from a point source over an angular range as much as 6 degrees and focus them to a spot with dimensions on the order of 0.2 mm FWHM. These monolithic polycapillary glass optics may be used for microanalysis with an SEM and an energy dispersive detector such as a lithium-drifted silicon detector, germanium detector or a cryogenically cooled microcalorimeter. As depicted schematically in FIGS. 3 and 4, there are two ways to produce point-to-point focusing with capillary bundles. The first as shown in FIG. 3 is with a single monolithic polycapillary bundle 20. We have tested such a capillary bundle with x-rays over an energy range extending to 6 keV. The optic used in this test had a point-to-point focal distance of 14 inches. The gain in intensity measured as a function of energy is as high as 400 as shown in FIG. 17. The second method uses two monolithic polycapillary bundles 22 shown in FIG. 4. The input optic intercepts the x-rays from the point source 10 and directs the radiation 12 into a parallel bundle. The output lens portion intercepts the parallel portion of x-rays and refocuses them to a spot. This technique has the advantage that the distance between the source and the image is variable and does not require a specific monolithic polycapillary bundle to be manufactured each time an experimental configuration is modified. A spectrum from a polycapillary glass optics setup is shown in FIG. 18 and the gain as a function of energy is presented in FIG. 19. The concentrators of the present invention may have application in the field of radiography, x-ray lithography and radiation therapy. For example, in conventional mammographic machines, a point source of Mo K x-rays forms a divergent beam that passes through the breast and is recorded on a photographic plate. Lesions in the breast tissue show up in the image as regions of contrasting intensity. Since the breast tissue is thick, the lesion can be located at varying distances along the beam path. The beam divergence will cause the recorded size of the lesion to vary according to its location along the beam path. This effect causes a loss of spacial resolution and can affect the resulting diagnosis of the mammogram. This effect would be absent if, instead of being divergent, the x-ray beam was parallel. Large diameter parallel x-ray beams are not commonly available since most conventional x-ray sources are derived from point-like geometries. The solution to this problem is to introduce an optical system between the source and the breast that makes a parallel beam from the x-rays diverging from the point source. This can be accomplished either by a set of nested cones that have been multi-layered to reflect Mo K x-rays with high efficiency as shown in FIG. 20 or a point-to-parallel bundle of glass capillary tubes. Both are suitable approximations to a parabolic lens and will provide a quasiparallel beam with small angular divergence. Some angular divergence is required to allow the x-rays reflected from successive cones to-overlap and remove any shadows of the cones. For this parallel beam, the image of a lesion will not be affected by its location along the beam path and a degree of uncertainty will be removed from the diagnosis. Similarly, as shown in FIG. 21, concentrators of the invention may form an optic for x-ray micro-lithography. Low energy x-rays are generally used for micro-lithography and multi-layering will not be necessary. A quasiparallel beam insures that the mask will be imaged with accuracy on a substrate. The lack of beam divergence means that it will be possible to construct features with thinner lines on the substrate. For applications where x-ray therapy is required to destroy lesions located deep in tissue, the normal approach is to use a finely collimated beam that intersects the location of the lesion. This approach causes all the tissue along the line of sight to receive roughly the same high dose of radiation. One approach to provide lower doses to the surrounding tissue than for the lesion is to have the radiation enter the body within the volume of a cone whose apex is located at the lesion. This can be achieved mechanically by rotating the patient about the apex of the cone centered on the lesion. The pencil beam always goes through the apex, but with a variety of directions thus reducing the exposure to the healthy surrounding tissue. Another approach that achieves the same goal is to use an optic that will refocus a diverging beam from an x-ray source as shown in FIG. 22. The focal point of the optic is located at the lesion and the conical, refocusing beam will put maximum intensity on the lesion and much less on the healthy surrounding material. The optic can be made as an approximate point-to-point lens. The approximation can be in the form of nested cylinders or two opposed sets of nested cones. In either case, the mirrors are made from thin foils that have been multilayered to reflect the x-rays of interest. It is recognized that modifications and variations of the disclosed invention may be apparent to those skilled in the art and it is intended that all such modifications and variations be included within the scope of the appended claims. |
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054105765 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to containers for disposing of low level radioactive waste and its detection and more particularly pertains to disposing of low level radioactive waste through specially configured containers adapted to detect its presence therewithin. 2. Description of the Prior Art The use of containers is known in the prior art. More specifically, containers heretofore devised and utilized for the purpose of supporting waste to be disposed are known to consist basically of familiar, expected, and obvious structural configurations, notwithstanding the myriad of designs encompassed by the crowded prior art which has been developed for the fulfillment of countless objectives and requirements. By way of example, U.S. Pat. No. 244,393 to Collica discloses a waste container for radioactive material. U.S. Pat. No. 4,495,139 to Janberg discloses a radioactive waste container with leak monitor. U.S. Pat. No. 4,760,268 to Noe discloses a container for low or medium activity radioactive waste. U.S. Pat. No. 4,894,550 to Baatz discloses a shielded radioactive waste container. Lastly, U.S. Pat. No. 4,996,019 to Catalayoud discloses a storage container for radioactive waste. In this respect, the containers for disposing of low level radioactive waste and its detection according to the present invention substantially depart from the conventional concepts and designs of the prior art, and in doing so provide an apparatus primarily developed for the purpose of disposing of low level radioactive waste through specially configured containers adapted to detect its presence therewithin. Therefore, it can be appreciated that there exists a continuing need for new and improved containers for disposing of low level radioactive waste and its detection which can be used for disposing of low level radioactive waste through specially configured containers adapted to detect its presence therewithin. In this regard, the present invention substantially fulfills this need. SUMMARY OF THE INVENTION In view of the foregoing disadvantages inherent in the known types of containers now present in the prior art, the present invention provides improved containers for disposing of low level radioactive waste and its detection. As such, the general purpose of the present invention, which will be described subsequently in greater detail, is to provide a new and improved container for disposing of low level radioactive waste and its detection and method which has all the advantages of the prior art and none of the disadvantages. To attain this, the present invention essentially comprises a new and improved container for disposing of low level radioactive waste and its detection comprising a container having a cylindrical side wall of an enlarged diameter and an enlarged height. The container has a bottom wall with its exterior periphery coupled to the lower edge of the side wall. The container also has a recess through the center of the base plate with an upwardly extending cylindrical support of a reduced diameter and shortened height extending upwardly from the aperture of the base plate. A liner is formed of a flexible material. The liner is configured to fit interiorly of the side wall with its upper edges extending over the upper edge thereof. The liner has a lower face adapted to be positioned on the interior face of the base plate. The liner also has an upwardly extending cylindrical extension adapted to be positioned over the upwardly extending interior cylinder of the container. A holder having a base, side walls and a cut-out in the exterior wall positioned on the exterior surface of the side wall adjacent the lower extent thereof for holding a meter. The meter is of the type having a probe for radioactive material positionable upwardly through the aperture of the base plate into the interior cylinder. A lid is provided in a circular configuration with a downwardly extending flange positionable over the exterior periphery of the container at its upper edge. There has thus been outlined, rather broadly, the more important features of the invention in order that the detailed description thereof that follows may be better understood and in order that the present contribution to the art may be better appreciated. There are, of course, additional features of the invention that will be described hereinafter and which will form the subject matter of the claims appended hereto. In this respect, before explaining at least one embodiment of the invention in detail, it is to be understood that the invention is not limited in its application to the details of construction and to the arrangements of the components set forth in the following description or illustrated in the drawings. The invention is capable of other embodiments and of being practiced and carried out in various ways. Also, it is to be understood that the phraseology and terminology employed herein are for the purpose of descriptions and should not be regarded as limiting. As such, those skilled in the art will appreciate that the conception, upon which this disclosure is based, may readily be utilized as a basis for the designing of other structures, methods and systems for carrying out the several purposes of the present invention. It is important, therefore, that the claims be regarded as including such equivalent constructions insofar as they do not depart from the spirit and scope of the present invention. Further, the purpose of the foregoing abstract is to enable the U.S. Patent and Trademark Office and the public generally, and especially the scientists, engineers and practitioners in the art who are not familiar with patent of legal terms or phraseology, to determine quickly from a cursory inspection the nature and essence of the technical disclosure of the application. The abstract is neither intended to define the invention of the application, which is measured by the claims, nor is it intended to be limiting as to the scope of the invention in any way. It is therefore an object of the present invention to provide new and improved containers for disposing of low level radioactive waste and its detection which have all the advantages of the prior art containers and none of the disadvantages. It is another object of the present invention to provide new and improved containers for disposing of low level radioactive waste and its detection which may be easily and efficiently manufactured and marketed. It is further object of the present invention to provide new and improved containers for disposing of low level radioactive waste and its detection which are of durable and reliable constructions. An even further object of the present invention is to provide new and improved containers for disposing of low level radioactive waste and its detection which are susceptible of a low cost of manufacture with regard to both materials and labor, and which accordingly are then susceptible of low prices of sale to the consuming public, thereby making such containers for disposing of low level radioactive waste and its detection economically available to the buying public. Still yet another object of the present invention is to provide new and improved containers for disposing of low level radioactive waste and its detection which provide in the apparatuses and methods of the prior art some of the advantages thereof, while simultaneously overcoming some of the disadvantages normally associated therewith. Still another object of the present invention is to dispose of low level radioactive waste through specially configured containers adapted to detect its presence therewithin. Lastly, it is an object of the present invention to provide new and improved containers for disposing of low level radioactive waste and its detection comprising a container having a cylindrical side wall of an enlarged diameter and an enlarged height. The container has a bottom wall with its exterior periphery coupled to the lower edge of the side wall. The container also has an aperture through the center of the bottom wall with an upwardly extending cylindrical support of a reduced diameter and shortened height extending upwardly from the aperture of the bottom wall. A liner is formed of a flexible material. The liner is configured to fit interiorly of the side wall with its upper edges extends over the upper edge thereof. The liner has a lower face adapted to be positioned on the interior face of the bottom wall. The liner also has an upwardly extending cylindrical extension adapted to be positioned over the upwardly extending interior cylinder of the container. These together with other objects of the invention, along with the various features of novelty which characterize the invention, are pointed out with particularity in the claims annexed to and forming a part of this disclosure. For a better understanding of the invention, its operating advantages and the specific objects attained by its uses, reference should be had to the accompanying drawings and descriptive matter in which there is illustrated preferred embodiments of the invention. |
description | This application is a 371 application of International Application No. PCT/GB2008/001920 filed Jun. 5, 2008, which claims priority to United Kingdom Patent Application No. 0713276.4 filed Jul. 9, 2007. Each of the foregoing applications is hereby incorporated herein by reference. The present invention relates to a transmission electron microscope. When imaging non-conducting specimens by transmission electron microscopy (TEM), beam-induced positive charge builds up on the specimen due to the ejection of secondary electrons. Transmission images of such charged specimens are degraded due to (1) electrostatic perturbation of the imaging optics and (2) charge-induced movement and modification of the specimen. These problems are a major limitation to a wide variety of imaging experiments in biology and materials science, including the imaging of frozen-hydrated specimens by cryomicroscopy. For specimens that are resistant to radiation damage by the imaging electron beam, images are often recorded after sufficient pre-exposure such that the positive charge build-up on the specimen reaches a steady-state because secondary electrons cannot escape the positive charge. For specimens that are not resistant to radiation damage, pre-exposure is not an option, because relevant structural details of the specimen must be recorded using the first few electrons that irradiate the specimen. Thus the charge on the specimen can change for the duration of the exposure. Image degradation by charging may be most problematic for cryomicroscopy of biological specimens precisely under imaging conditions that are otherwise most advantageous for imaging structural detail, such as when they are suspended in holes over ice, or at liquid helium temperature where specimen conductivity is reduced. Brink et al., Evaluation of charging on macromolecules in electron cryomicroscopy, Ultramicroscopy, 72 (1998) 41-52 describes charge build up on non-conducting specimens due to secondary electron emission. In particular, an experiment is disclosed in which a small diameter beam is used to charge up a specimen, and a wide diameter beam is then used to observe the charged area and eventually discharge it. It is suggested that some of the secondary electrons which are emitted across the entire region when the specimen is examined with the wide beam return to compensate the built up positive charge. Warrington, A simple charge neutralizer for the electron microscope, J. Sci. Instrum., 43 (1966) 77-78, proposes a charge neutralizer consisting of an earthed film of vacuum deposited carbon and aluminium supported above the specimen plane of the objective lens. The electron beam passes through the film before striking the specimen to be examined. Low energy electrons ejected from the film then discharge the non-conducting specimen. Beam-induced positive charge can build up on other non-conducting bodies, such as electron optical elements, located on the path of the electron beam, and degrade their performance. Examples of such bodies are phase plates and electron biprisms. US 2002/0011566 discloses an antistatic phase plate for use in phase-contrast electron microscopy, the phase plate being made of a thin film of conductive amorphous material. Frost, Image-plane off-axis electron holography: low-magnification arrangements, Meas. Sci. Technol., 10 (1999) 333-339, discusses measurements of the deflection angle at an electron biprism which indicate that the biprism fibre is positively charged by the imaging electron beam. The present invention aims to overcome or mitigate problems of beam-induced positive charge build up. In a first aspect, the present invention provides a transmission electron microscope (TEM) having: a target body position on the electron optical axis of the microscope, an electrically conductive body off the axis of the microscope, an electron source for producing an axial electron beam which, in use, impinges upon a target body located at the target body position, and a system for simultaneously producing a separate off-axis electron beam which, in use, impinges on the electrically conductive body causing secondary electrons to be emitted therefrom; wherein the electrically conductive body is located such that the emitted secondary electrons impinge on the target body to neutralise positive charge which may build up on the target body. As used herein, the term “separate off-axis electron beam” excludes any off-axis electron beam that may be produced by scattering or diffraction of the axial electron beam. By producing, simultaneously with the axial electron beam, a separate off-axis electron beam that causes charge-neutralising secondary electrons to be emitted, the TEM can be used e.g. to image a non-conducting specimen while at the same time operating to reduce charge build-up on the specimen. In other words, the separate off-axis electron beam and the off-axis electrically conductive body may be thought of as a dedicated system for reducing or eliminating charge build-up. Advantageously, and in contrast to the observations of Brink et al. ibid., the TEM operator can image a specimen with a narrow-diameter beam and still avoid problems of beam-induced positive charge build up. Further, the inconvenience and disturbance of a charge neutralizer positioned in the path of the axial electron beam according to the proposal of Warrington ibid. can be avoided. Preferably, the off-axis electrically conductive body is located adjacent to the target body position. The secondary electrons emitted from the body will then have a relatively short distance to travel before impinging on the target body, which can increase the flux of impinging electrons. Preferably, the off-axis electron beam is a paraxial electron beam. A paraxial electron beam can be defined as a beam that is focusable onto the electron optical axis by the TEM lenses, but has a minimum divergence angle that is greater than the maximum divergence angle of the axial electron beam. Conveniently, the off-axis electron beam can be produced by the same electron source that produces the axial electron beam, i.e. the system for simultaneously producing a separate off-axis electron beam can include the electron source. Advantageously, using this approach, a conventional TEM can readily be converted into a TEM according to the present invention. For example, the system for producing an off-axis electron beam may comprise an aperture body positioned between the electron source and the target body position, the aperture body having an axial aperture for transmission of the axial electron beam and further having an off-axis aperture for production of the off-axis electron beam. Such an aperture body may simply replace an existing condenser aperture body of an existing TEM. The system for producing an off-axis electron beam may produce a plurality of such beams which, in use, impinge on the electrically conductive body (or, more preferably, respective electrically conductive bodies). This makes it possible, for example, to neutralise build up of positive charge at respective target bodies at spaced positions on the electron optical axis position. Thus, the aperture body may have a plurality of off-axis apertures for production of respective off-axis electron beams, each off-axis electron beam, in use, impinging on the off-axis electrically conductive body or a respective off-axis electrically conductive body. Typically, the TEM has at least one condenser lens between the electron source and the target body position, and the aperture body may be positioned between the condenser lens and the target body position. Thus the aperture can limit the illuminating field of the condenser lens. Indeed, in a further aspect, the present invention provides a multi-aperture aperture body as discussed above. In other embodiments of the first aspect, the system for producing an off-axis electron beam may comprise a further electron source (or a plurality of further electron sources if a plurality of off-axis electron beams are to be deployed). This can increase the complexity and cost of the TEM, and may make it more difficult to convert a conventional TEM into a TEM according to the present invention. However, a further electron source for producing the off-axis electron beam can provide an advantage by allowing the off axis beam intensity to be varied independently of the axial beam, or for the off-axis beam to remain constant if the axial beam is pulsed. Typically, the target body position is a specimen position, the axial electron beam, in use, impinging upon a specimen, and the emitted secondary electrons impinging on the specimen to neutralise positive charge which may build up on the specimen. Conveniently, the off-axis electrically conductive body can then be provided by a specimen support which holds the specimen at the specimen position. However, the target body, or one of the target bodies, can be an electron optical element such as a phase plate or an electron biprism. In such cases the axial beam will typically impinge on a specimen, and that specimen may be another target body. When there are plural target bodies, each may have a respective electrically conductive body. FIG. 1 shows schematically a non-conducting specimen, and a TEM axial electron beam (block arrows, e) used to image the specimen. The electron beam causes secondary electrons (line arrows, SE) to be ejected from the specimen, leaving the specimen with a positive charge. This positive charge can cause electrostatic perturbation of the imaging optics and also can cause movement and Coulombic explosion of the specimen. FIG. 2 shows schematically a longitudinal section through a TEM according to an embodiment of the invention. An axial electron beam 1 produced by an electron source 2 impinges on a non-conducting specimen 3 held by specimen holder 4. A multi-hole aperture 5 positioned between TEM condenser lens 6 and the specimen has a central hole 7 for the axial electron beam and an off-centre hole 8 which produces a paraxial electron beam 9. The paraxial electron beam irradiates a grounded conductor 10 which is adjacent the specimen but off the axis of the TEM, the grounded conductor being integral with the specimen holder. The irradiation of the paraxial electron beam causes secondary electrons SE to be emitted by the grounded conductor, and some of these electrons in turn impinge on the non-conducting specimen to neutralise positive charge which has built up on the specimen. Thus the paraxial beam and the grounded conductor act as charge compensator for the specimen. A similar arrangement (not illustrated) can be used to neutralise positive charge on other bodies, such as a phase plate or an electron biprism, which are susceptible to charge build up due to secondary electron emission. FIG. 3 shows schematically a multi-hole aperture as viewed along the optical axis of a TEM. The aperture has a central hole 7 and six off-centre holes 8 circumferentially spaced around the central hole. Each off-centre hole can produce a respective paraxial beam. A typical diameter for the holes could be about 50 μm and typical centre-to-centre spacing for the holes could be about 200 μm. Such an aperture could be retrofitted to an existing microscope to allow it to produce several off-axis paraxial beams. FIG. 4 is a recording on film of the axial and paraxial electron beams produced using a condenser aperture having a 50 μm diameter central hole and one 100 μm diameter off-centre hole. FIG. 5(a) is an image recorded on a CCD camera of a seven-hole condenser aperture according to an embodiment of the present invention. The image was produced by placing the seven-hole aperture in the specimen holder of the electron microscope. Each hole is 50 μm diameter and the offset is 200 μm centre-to-centre. FIG. 5(b) is an electron microscope recording on a CCD camera of electron beams produced by the seven-hole aperture. The aperture is arranged to produce axial and paraxial beams on the specimen. The condenser lens demagnifies the image of the apertures onto the specimen plane. The paraxial beams are slightly elliptical compared to the axial beam due to spherical aberration of the condenser lens system. FIG. 6 is a recording on a CCD camera of a typical beam-sensitive (frozen-hydrated) specimen with the positions of beams indicated for imaging of a specimen using the seven-hole aperture. If a single hole aperture is required in place of the seven-hole aperture, both apertures can mounted on the microscope aperture holder so that one or the other can be shifted into position as needed. Alternatively, an extra occluding aperture could be inserted to obstruct the paraxial beams. The imaging protocol consisted of the identification of a region of interest in the specimen at low magnification (e.g. 10K times) using single beam illumination, focus determination by imaging the support adjacent to the specimen at a high magnification (e.g. 200K times) using single beam illumination, and then image recording of the region of interest at intermediate magnification (e.g. 60K times with focus parameters established during focus mode) using either the single or the seven-hole aperture. In this way the pre-exposure of the region of interest is minimized. The specimen is a holey carbon film covered with a thin film of vitreous ice containing material of interest at liquid nitrogen temperature (−195° C.). The region of interest is material located in the ice film over any of the holes. As FIG. 6 shows, in this case the axial beam position is within a hole. FIG. 6 shows a low magnification image recorded subsequent to focusing and exposure. The focus position is 2 μm away from the axial beam position and can be identified as a bright spot where the concentrated dose has removed a layer of ice from the carbon. The position of the six off-axis beams on the surrounding carbon is also indicated. They produce a characteristic charging “footprint” in the overlying ice (Brink et al. ibid.). Irradiation of the carbon support at the paraxial beam positions causes SE electrons to impinge on the axial beam position located at the thin film of ice over the hole in the carbon support. FIG. 7 shows two images of the same area of a typical single particle specimen in vitreous ice over a hole in a carbon support (a) with the seven-hole aperture used as a charge compensator and (b) without the compensator. In each case the axial beam was within the hole and did not irradiate the adjacent carbon support. In image (a) the paraxial beams from the compensator irradiated the adjacent carbon support. These beams were absent in image (b). The reduced image quality in the absence of the compensator (image (b)) is attributed to charging of the film resulting in image blurring and distortion by the mechanisms described above. While the invention has been described in conjunction with the exemplary embodiments described above, many equivalent modifications and variations will be apparent to those skilled in the art when given this disclosure. Accordingly, the exemplary embodiments of the invention set forth above are considered to be illustrative and not limiting. Various changes to the described embodiments may be made without departing from the spirit and scope of the invention. |
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047298685 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to pressurized water reactors and, more particularly, to a vibration arrestor for rod guides positioned within the inner barrel assembly of a pressurized water reactor. 2. State of the Prior Art Certain advanced designs of nuclear reactors incorporate at successively higher, axially aligned elevations within the reactor vessel, a lower barrel assembly, an inner barrel assembly, and a calandria, each of generally cylindrical configuration, and an upper closure dome. The lower barrel assembly may be conventional, having mounted therein, in parallel axial relationship, a plurality of fuel rod assemblies which are supported at the lower and upper ends thereof, respectively, by corresponding lower and upper core plates. Within the inner barrel assembly there is provided a large number of rod guides disposed in closely spaced relationship, in an array extending substantially throughout the cross-sectional area of the inner barrel assembly. The rod guides are of first and second types, respectively housing therewithin reactor control rod clusters (RCC) and water displacer rodlet clusters (WDRC); these clusters, as received within their respectively associated guides, generally are aligned with the fuel rod assemblies. The calandria includes a lower calandria plate and an upper calandria plate. The rod guides are secured in position at the lower and upper ends thereof respectively, to the upper core plate and the lower calandria plate. Within the calandria and extending between the lower and upper plates thereof is mounted a plurality of calandria tubes in parallel axial relationship and respectively aligned with the rod guides. A number of flow holes are provided in remaining portions of the calandria plates, intermediate the calandria tubes, through which passes the the reactor core outlet flow as it exits from its passage through the inner barrel assembly. In similar parallel axial and aligned relationship, the calandria tubes are joined to corresponding flow shrouds which extend to a predetermined elevation within the dome, and which in turn are connected to corresponding head extensions which pass through the structural wall of the dome and carry, on their free ends at the exterior of and vertically above the dome, corresponding adjustment mechanisms. The adjustment mechanisms have corresponding control lines which extend through the respective head extensions, flow shrouds, and calandria tubes and are connected to the respectively associated clusters of RCC rods and WDRC rods, and serve to adjust their elevational positions within the inner barrel assembly and, particularly, the level to which same are lowered into the lower barrel assembly and thus into association with the fuel rod assemblies therein, thereby to control the activity within the core. A critical design criterion of such reactors is to minimize wear of the rodlets at interfaces between the individual rodlets of a given cluster and known support plate structures within the rod guide through which the rodlets pass for support, and thus to reduce or eliminate the factors which produce wear, such as flow induced vibration and associated vibration of reactor internal structures. Because of the relatively dense packing of the rod guides within the inner barrel assembly, it is critical to maintain substantially uniform distribution of the outlet flow from the reactor core, and an axial direction of that flow through the upper barrel assembly. Even if a uniform, axial flow of the core outlet is achieved, the effects of differential pressure and temperature across the array of rod guides, or an individual rod guide, can produce significant reaction loads at the support points, or support connections, for the rod guides. These reactor loads, coupled with the flow induced vibrating create a high potential for wear of the rod guides, as well as the rodlets. Additionally, the provision of the calandria, and particularly the lower plate thereof, presents an interface with the top end of the rod guides which does not exist in conventional pressurized water reactors. That interface must be capable of accommodating differential thermal expansions between the lower calandria plate and the inner barrel in order to prevent large thermal stresses from developing. Furthermore, the bottom calandria plate and the upper core plate are essentially structurally independent; therefore, vibration of the internals can result in significant relative movement between the supporting connections of the rod guides at their lower and upper ends respectively to the upper core plate and the bottom calandria plate. The wear potential under these circumstances is great. Thus, split pin connections of conventional types are inappropriate for use as the supporting connections for the top ends of the rod guides since they would wear rapidly, with the result that the top ends of the rod guides would become loose. Rod guides having such loose top end connections are unacceptable because of the rapid rate of wear of the rodlets which would result. Other known mounting devices as well are inappropriate. For example, leaf springs cannot be used to support all of the rod guides because sufficient space is not available within the inner barrel assembly to provide leaf springs of the proper size for the large number of rod guides which are present, even if high strength material is used for the leaf springs. Beyond the unsuitability of existing, known structural support arrangements, further factors must be taken into account in the consideration of possible designs for the support of the top end of the rod guides within the inner barrel assembly. For example, both the RCC and the WDRC rod clusters should be removable without requiring that the guides be disassembled. This requirement imposes a severe space limitation in view of the dense packing of the guides and their associated rod clusters within the inner barrel assembly. For example, in one such reactor design, over 2,800 rods are mounted in 185 clusters, the latter being received within a corresponding 185 guides. The space limitation is further compounded by the requirement that unimpeded flow holes must be provided in the calandria plates for the core outlet flow. While these foregoing factors severly restrict the available space envelope in the horizontal cross-sectional dimension of the inner barrel assembly, axial or vertical limitations on this space envelope must also be considered. For example, the presence of the support members should not require any increase in the height of the vessel. From a maintenance standpoint, the support members should be visible for inspection and replaceable without undue effort. Additionally, the assembly load of the calandria must be less than its dead weight and must be accomplished without access to the support region. This avoids having to apply force to the calandria before installing the vessel head. While the supports for the rod guides must therefore satisfy a wide range of structural and functional requirements relating to, or imposed by, the inner barrel assembly itself, a further critical requirement is that the wear potential of the support structure itself must be minimized. This is a critical requirement in view of the potential for intense vibration arising out of the core outlet flow and the development of high contact forces due to differential pressure and both steady state and transient temperature conditions across both the array of rod guides and the individual rod guides. Conventional reactor designs do not present the support problems attendant the dense packing of rod guides and associated rod clusters in advanced reactor designs of the type herein contemplated. Thus, there is no known solution to the problems of adequately supporting the rod guides, consistent with the requirements and taking into account the environmental factors which exist in operation of such reactors as hereinabove set forth. SUMMARY OF THE INVENTION A pressurized water nuclear reactor, of the type with which the vibration arrestors for rod guides of the inner barrel assembly in accordance with the present invention are intended for use, employs a large number of reactor control rods, or rodlets, typically arranged in what are termed reactor control rod clusters (RCC) and, additionally, a large number of water displacer rods, or rodlets, similarly arranged in water displacer rod clusters (WDRC). For example, in one such reactor, an array of 185 such clusters containing a total of 2800 rodlets (i.e., the total of reactor control rods and water displacer rods) are mounted in parallel axial relationship within the inner barrel assembly. Each of these clusters, moreover, is received within a corresponding rod guide structure. In operation, it is desired to maintain the core outlet flow in an axial flow condition and in a substantially uniform distribution throughout the cross-sectional area of the inner barrel assembly, as it passes through the inner barrel assembly, and thus to prevent cross-flow conditions (i.e., core flow in a direction transverse of the rod guides). This is a critical requirement in reactors of such advanced designs in which the inner barrel is densely loaded with rodlets, as before noted. The geometry of the reactor vessel itself introduces a structural anomaly which is contrary to maintaining the desired, substantially uniform axial flow condition. Particularly, the circular configuration of the reactor vessel, including the inner barrel assembly, is geometrically incompatible with the generally rectangular or square cross-sectional configuration of the individual rod guides, and correspondingly of an array thereof as stacked in closely adjacent relationship within the inner barrel assembly. Thus, in the peripheral regions between the inside diameter of the cylindrical inner barrel assembly and the outer periphery of the array of rod guides, no rodlets are present, resulting in a nonuniform flow distribution and presenting at least the potential of turbulence and cross-flow conditions with attendant problems of vibration. A related, copending application of a common one of the co-inventors herein, entitled "Modular Former For Inner Barrel Assembly Of Pressurized Water Reactoring", and assigned to the common assignee hereof, discloses an invention relating to modular formers which are configured to be mounted in these peripheral regions, to provide hydraulic resistance and thereby to maintain a primarily axial direction, and substantially uniform distribution, of the core outlet flow, throughout the length of the rod guides within the inner barrel assembly. Thus while the state of the art, in the design of the inner barrel assembly of such advanced types of pressurized water reactors, has addressed the problem of attempting to maintain relatively stable conditions by minimizing cross-flow, e.g. by maintaining substantially uniform distribution and axial direction of the core output flow throughout the inner barrel assembly, there remains the critical problem of properly supporting the rod guides within the inner barrel because of remaining excitation forces from internal vibration and axial flow turbulence, consistent with the objectives and the structural and operating conditions and parameters as hereinabove set forth. The vibration arrestors in accordance with the present invention, for use with rod guides of the inner barrel assembly of a pressurized water reactor, afford a highly efficient and effective structure for satisfying the critical design criteria relating to flow induced vibrations of structural components and lateral force effects, as particularly relate to the rod guides within the inner barrel assembly. In one preferred use or application of the vibration arrestors in accordance with the present invention, they are employed in combination with a flexible rod guide support structure which is the subject of a copending application of a common inventor hereof, entitled "Flexible Rod Guide Support Structure for Inner Barrel Assembly of Pressurized Water Reactor," assigned to the common assignee hereof. Particularly, the flexible rod guide support structures as disclosed in the referenced, copending application, comprise, as major components, interdigitized matrices of top plates for the rod guides, flexible linkages which interconnect the top plates in a concatenated arrangement, pin stops between the continuous top plates of the two matrices, mounting extensions from the calandria which engage the top plates of one matrix, and rod guide leaf springs which are mounted on the calandria and which exert a force against the top plates of the one matrix to restrain lateral movement. These components are configured in a pattern that is repeated across the interface between the tops of all the rod guides in the array and the bottom plate of the calandria. Each such flexible linkage is attached to a respective WDRC rod guide top plate and to each of the RCC guide top plates which contiguously surround the respective, given WDRC rod guide top plate. Thus, each WDRC guide is attached, or concatenated, laterally to its surrounding RCC rod guides via the flexible linkage. This concatenated assembly of linkages creates a stiff structure between the guides in a plane perpendicular to the axis of the rod guides. Thus, the guides are essentially bound together laterally; however, the linkages in the out-of-plane direction, i.e., axially, are flexible and thus accommodate relative axial motion between guides to permit bowing of adjacent guides. This capability of flexibility in one plane compensates for local differences in height of adjacent guides due to differential thermal expansion and bowing due to pressure differential across the guide. Thus, the flexible linkages are flexible in a direction parallel to the axis of the rod guide, but rigid in a plane perpendicular to the axis of the rod guide. Lateral loads exerted on the rod guides are reacted into the calandria either by the calandria extensions or by the leaf springs, at each of the RCC plates. The rod guide leaf springs, as mounted on the calandria plate and pressed against the RCC top plates, generate sufficient lateral frictional force such that fluctuating steady state loads exerted on the guides do not cause slippage. Moreover, the mounting extensions from the calandria provide overall lateral support during events such as seismic, which can exceed the lateral frictional force of the leaf springs, and provide alignment between the rod guides and the calandria, there being one extension for each of the RCC guides. Collectively, the calandria extensions react the seismic loads from the rod guides. Alignment of the RCC clusters in the rod guide top plates further is controlled by the calandria extensions. The vibration arrestors of the present invention may be of differing embodiments, two specific embodiments thereof being disclosed herein, and in either embodiment may be employed as an improvement, in the alternative to the leaf springs disclosed in the referenced, copending application. Whereas leaf springs of the type disclosed therein are appropriate choices for the function required thereof, as above described, in view of the prior experience in the use thereof in connection with fuel rod assemblies, the leaf springs present certain obstacles or disadvantages which are overcome by the vibration arrestors of the present invention. For example, the leaf springs introduce numerous individual parts (in the referenced, exemplary reactor vessel design, in excess of 2,000 parts), adding considerably to the time and cost of initial assembly and continuing maintenance expense, for a given reactor vessel installation. The vibration arrestors in accordance with the present invention have an optimum design for use with rod guide top plates of the general configuration and mounting provisions of the RCC top plates hereinabove described, and achieve a substantial reduction in the number of parts--at the level of an order of magnitude smaller--while yielding superior structural performance at reduced stress levels. More particularly, the vibration arrestors of the present invention comprises a central hub of generally annular configuration, functioning as a mounting base, and integral spring arms which extend therefrom in a pattern which is symmetrical about the center of the hub. In one specifically disclosed embodiment of the arrestors, a single pair of two such spring arms extend in aligned and oppositely oriented directions from the hub (and thus angularly displaced by 180.degree. about the hub). In a second embodiment, two such pairs of spring arms are formed integrally with the hub and extend therefrom in quadrature relationship. The arrestors are formed of metal of constant thickness, and each of the spring arms has a simple taper in the width dimension along the generally radial length thereof. The hub includes a central aperture by which it is received over a corresponding calandria extension. A clamping, or stiffening, ring having an outer periphery corresponding to the hub portion of the arrestor is received over the calandria extension in superposed relationship with the hub, the ring and the hub having holes extending therethrough which are positioned in alignment with corresponding threaded bores in the lower calandria plate for receiving attachment bolts thereby to secure the arrestor to the calandria. The calandria extensions both position and laterally support the respective vibration arrestors, and thus prevent the attachment bolts from reacting lateral loads which may be imparted on the spring arms and transmitted thereby through the hub to the calandria. Moreover, the symmetrical configuration of the vibration arrestors prevents bending torques from being applied to the attachment bolts, because the compression loads applied to the ends of the symmetrically oriented spring arms are substantially the same and thus exert no net external moment. Due to the symmetrical configuration of the vibration arrestors, forces applied to the spring arms generate primarily an internal moment in the hub of the arrestor, which is made of sufficiently high strength material to withstand the stress. The clamping ring moreover reinforces the arrestor hub and prevents localized stresses therein due to attachment bolt preload effects. There results primarily only tensile loads on the bolts, with minimal, if any, shear and moment forces which is a highly desirable and acceptable condition. The vibration arrestors of the present invention thus achieve not only a reduction in the number of parts and corresponding time and cost of assembly, relative to the leaf spring implementation, but additionally the clamping ring attachment structure and the balanced reaction to compression loads afforded by the symmetrical configuration thereof provide improved operational characteristics and enhanced reliability by substantially eliminating the bending moment on the attachment bolts. While an exemplary application of the vibration arrestors of the present invention may be in assemblage with a flexible rod guide support structure of the type disclosed in the above identified and similarly entitled copending application, the vibration arrestors as well may be employed independently with rod guides of any desired type, and are not restricted in use to the specific interleaved matrices of first and second different types of rod guides, as disclosed in that copending application. Thus, for example, where employed independently with a given type of rod guides, such as the RCC rod guides, alternative mounting means may be employed for other types of rod guides, such as the WDRC rod guides; in such an installation, the WDRC rod guides may be supported independently, for example, by the top end support structure disclosed in the copending application entitled: "Top End Support for Water Displacement Rod Guides of Pressurized Water Reactor," having a common coinventor herewith and assigned to the common assignee hereof. The vibration arrestors of the present invention thus afford greatly enhanced beneficial effects, corresponding and relative to the leaf springs as described in the earlier-referenced copending application; thus, the arrestors as well react lateral force components on the associated rod guides, even in the event of wear of the calandria extensions, and thus suppress top end lateral motion and correspondingly prevent any increase in the excitation of the associated rodlets. This assures that rodlet wear does not increase, despite the potential of slippage due to inadvertent wear of the rod guide support, the need for significant gaps to permit assembly, and resultant increased tolerances between adjoining parts. Likewise, the vibration arrestors of the present invention increase, by more than an order of magnitude, the allowable wear depth on the calandria extension before alignment between the rodlet clusters and respective rod guides is compromised. Further, regardless of the gap size between the calandria extension and the respective rod guide top plate, the lateral excitation of rodlets within the respective rod guides is not affected. These and other advantages of the vibration arrestors for the rod guide supports in the inner barrel assembly of a pressurized water reactor, in accordance with the present invention, will become more apparent from the following detailed description and drawings. |
050739140 | claims | 1. A stereoscopic X-ray apparatus comprising: means for irradiating X-rays in two directions to an object to be examined; means for picking-up images created by X-rays having passed said object to provide X-ray images picked-up in the two directions; means for detecting the positional relation of the object with respect to said irradiation means and picking-up means; and stereoscopic display means for adjusting the positional relation between two X-ray images output from said picking-up means according to the positional relation of said object detected by said detection means and stereoscopically displaying the X-ray images. and means for shifting the display position of the X-ray image in the right or left direction by changing the readout addresses for respective picture elements of the X-ray image in said memory with respect to the write-in addresses thereof. means for comparing said distance with a standard distance and deriving a difference therebetween; and means for shifting said two X-ray images in the right and left directions, respectively, by a distance equal to half the difference. 2. An apparatus according to claim 1, wherein said irradiating means is a stereoscopic X-ray tube having two foci and irradiating X-rays from the two foci in two directions intersecting at a target point in said object. 3. An apparatus according to claim 1, wherein said irradiating means includes an X-ray tube, means for revolving said X-ray tube around said object by a preset angle, and means for irradiating X-rays which intersect at a target point in said object from said X-ray tube before and after the revolution of said X-ray tube. 4. An apparatus according to claim 1, wherein said stereoscopic display means includes means for alternately displaying said two X-ray images on the same display screen. 5. An apparatus according to claim 4, further comprising glasses means to be put on an observer and having shutters driven to interrupt or transmit the image impinging on the right and left eyes of the observer in synchronism with the alternate display operation of said stereoscopic display means. 6. An apparatus according to claim 1, wherein said stereoscopic display means includes means for simultaneously displaying said two X-ray images side by side. 7. An apparatus according to claim 6, further comprising glasses means to be put on an observer, for guiding the right and left X-ray images to the right and left eyes of the observer 8. An apparatus according to claim 1, wherein said stereoscopic display means includes means for shifting at least one of the X-ray images picked-up in the two directions in the right or left direction to adjust a distance between said two X-ray images to be displayed. 9. An apparatus according to claim 8, wherein said stereoscopic display means includes means for enlarging the X-ray images and then shifting at least one of them. 10. An apparatus according to claim 1, wherein said stereoscopic display means includes a memory for storing X-ray images picked-up in two directions; 11. An apparatus according to claim 1, wherein said stereoscopic display means includes means for detecting a distance between said object and said picking-up means and a distance between said irradiation means and said picking-up mean to detect a distance between said X-ray images on an image picking-up plane of said picking-up means; |
062164457 | claims | 1. A pulsed plasma thruster comprising: vapor producing solids; heat producing means arranged adjacent said solids; a thruster housing having a thrust discharge chamber with a plurality of openings, a thrust nozzle, and a fuel propellant in said thrust discharge chamber; passageways leading from said solids to said thrust discharge chamber within said housing, the passageways arranged so that vapors from said solids are received through said openings of said thrust discharge chamber; first and second electrodes extending from said thrust discharge chamber through said housing; and a power source coupled to said first and second electrodes and configured to enable spark breakdown between the electrodes of said thrust discharge chamber, the power source configured to control the voltage-current shape of spark breakdown that results in a plasma arc capable of ablating said fuel propellant, and ionizing said solid vapors and fuel propellants within said thrust discharge chamber to create a thrust force outwardly directed from said thrust nozzle. solids capable of producing a high pressure vapors; heat generating elements adjacent said solid and operably configured to generate heat that causes said solids to sublime; an ignition chamber forming a passageway from said solid to a thrust discharge chamber, said ignition chamber having a plurality of holes for guiding vapors from said solid to said thrust discharge chamber; said thrust discharge chamber having first and second ends, said first end coupled to said ignition chamber, said thrust discharge chamber further including two oppositely positioned electrode plates and two oppositely positioned fuel propellants, each of said electrode plates coupled to corresponding electrode terminals a nozzle coupled to said second end of said thrust discharge chamber; and a power processor unit coupled to said electrode plates through said electrode terminals and configured to provide an ignition voltage that causes a plasma arc to occur in the gap between said electrode plates; wherein said electrode plates are sized and shaped to assure transfer of said plasma arc in the direction of said nozzle to cause a predictable amount of said fuel propellants to be ablated and create a thrust force that exits said nozzle. heating the subliming solid to create a high pressure vapor; directing the high pressure vapor in the direction of said thrust discharge chamber; and applying a DC ignition signal to said electrodes to spark a breakdown of the PTFE propellants and cause a transition of the spark to a useful plasma arc. 2. The thruster of claim 1 wherein said passageways are configured to separate said solids from said thrust discharge chamber so that sparks and plasma do not interact with said solids. 3. The thruster of claim 1 wherein said passageways includes a plurality of holes adjacent said solids. 4. The thruster of claim 3 wherein said holes are arranged so that vapors are optimally fed into said thrust discharge chamber. 5. The thruster of claim 1 further comprising a means of varying the angle of said thrust nozzle with respect to a central axis extending through said thrust discharge chamber. 6. The thruster of claim 1 wherein said heat producing means are MEMS micro heaters capable of independently producing sufficient quantities of heat to cause said solids to sublime at times and locations to reduce ignition voltages and increase the PPT efficiency at desired values of propellant velocity. 7. The thruster of claim 1 wherein said first and second electrodes are made of a slightly radioactive material. 8. The thruster of claim 1 wherein said power source delivers a maximum DC voltage signal of 300 volts. 9. The thruster of claim 1 wherein said power source is configured to control the shape and magnitude of said ignition signal in three segments corresponding to an open circuit to a constant voltage segment, a constant voltage segment, and a constant current segment. 10. The thruster according to claim 1 further comprising a means of varying the spacing as a function of axial distance between said electrodes. 11. The thruster according to claim 1 wherein the fuel propellant comprises PTFE. 12. A pulsed plasma thruster comprising: 13. The pulsed plasma thruster according to claim 12 further including an insulating layer extending substantially over said thrust discharge chamber and said nozzle. 14. The pulsed plasma thruster according to claim 13 further including a housing surrounding said insulating layer with openings that allow access to said electrode terminals. 15. The pulsed plasma thruster according to claim 12 wherein the spacing between said electrode plates is less than 50 micrometers. 16. The pulsed plasma thruster according to claim 12 further comprising a UV variable intensity light source predisposed to provide an ignition signal that sparks vapors within said ignition chamber. 17. The pulsed plasma thruster according to claim 12 further comprising a means of varying the angle of said nozzle with respect to said thrust discharge chamber. 18. The pulsed plasma thruster according to claim 12 wherein said power processing unit is capable of producing multiple volt-amp signal forms that effect the shape of said ignition voltage. 19. The pulsed plasma thruster according to claim 18 wherein said power processing system produces an ignition voltage signal in three segments corresponding to an open circuit to constant voltage segment, a constant voltage segment and a constant current segment. 20. The pulsed plasma thruster according to claim 12 wherein said fuel propellants comprise PTFE propellants. 21. The pulsed plasma thruster according to claim 12 wherein said heat generating element comprises a micro-heater capable of heating said solid to create a vapor that travels into said thrust discharge chamber and exerts a pressure on said electrode plates. 22. The pulsed plasma thruster according to claim 12 wherein said holes of said ignition chamber are sized and located to provide uniform feeding of vapors subliming from said solid to said discharge thruster. 23. The pulsed plasma thruster according to claim 12 further comprising a plurality of heating elements embedded in said fuel propellants. 24. The pulsed plasma thruster according to claim 23 wherein said heating elements comprise variable temperature MEMS-based micro-heaters that can control the amount of said fuel propellants ablated. 25. The pulsed plasma thruster according to claim 12 wherein said thrust discharge chamber is arranged to prevent either of an ignition voltage and plasma arc from interacting with said solid. 26. The pulsed plasma thruster according to claim 12 wherein said electrode plates are evenly spaced about a central axis extending through said thrust discharge chamber. 27. The pulsed plasma thruster according to claim 12 further comprising first and second electrode positioning devices predisposed about said first and second electrode plates, respectively, for varying the spacing of said electrode plates as a function of the distance to a central axis extending through said thrust discharge chamber. 28. The pulsed plasma thruster according to claim 13 wherein said insulating layer is shaped to increase the local field strengths existing between said electrode plates. 29. The pulsed plasma thruster according to claim 12 wherein said ignition voltage less than 300V. 30. The pulsed plasma thruster according to claim 14 wherein said housing includes a means of mounting said thruster to a mass to be propelled. 31. A method of operating a pulsed plasma thruster having a heat generator means predisposed adjacent a subliming solid, a thrust discharge chamber formed of two oppositely positioned electrode plates and two oppositely positioned PTFE propellants, a device separating the subliming solid from the thrust discharge chamber, a power source coupled to the electrode plates and capable of providing an ignition voltage, the method comprising the steps of: 32. The method of operating a pulsed plasma thruster according to claim 31 wherein the step of applying a DC ignition signal is performed by controlling the shape and magnitude of the output of the power source. 33. The method of operating a pulsed plasma thruster according to claim 32 wherein the shape is controlled in three segments corresponding to an open circuit to constant voltage segment, a constant voltage segment and a constant current segment. 34. The method of operating a pulsed plasma thruster according to claim 32 wherein the step of directing the high pressure vapor in the direction of said thrust discharge chamber is performed in a manner that creates pressure between the two electrode plates of the thrust discharge chamber. 35. The method of operating a pulsed plasma thruster according to claim 32wherein the step of directing the high pressure vapor in the direction of said thrust discharge chamber is performed so that the vapor is fed uniformly to the thrust discharge chamber. 36. The method of operating a pulsed plasma thruster according to claim 32 further comprising the step of transferring an initial spark to a plasma arc in the gap defined by the separation of the electrode plates. 37. The method of operating a pulsed plasma thruster according to claim 36 wherein the step of transferring is performed in such a manner that an amount of PTFE propellant is ablated. 38. The method of operating a pulsed plasma thruster according to claim 37 wherein the amount of PTFE propellant is ablated in a controlled manner. 39. The method of operating a pulsed plasma thruster according to claim 32 wherein the step of applying a DC ignition signal to said electrodes is performed by focusing an ultraviolet light source on the high vapor pressure. 40. The method of operating a pulsed plasma thruster according to claim 32 further comprising the step of adjusting the spacing of the electrode plates to effect the force delivered by the propellants. 41. The method of operating a pulsed plasma thruster according to claim 32 wherein the step of applying a DC ignition signal to said electrodes is performed by limiting the DC voltage signal to less than 300 volts. |
summary | ||
claims | 1. A safety system for a nuclear plant, the safety system comprising:a plurality of catalytic recombiner elements each for triggering a recombination reaction with oxygen upon hydrogen being entrained in onflowing gas flows of a gas medium;said recombiner elements being configured to act as ignition elements and to create a pressure pulse in the gas medium resulting from an ignition during the recombination reaction;said recombiner elements being interconnected by flow paths and including a first recombiner element and a second adjacent recombiner element; andsaid recombiner elements and said flow paths each being configured to cause the pressure pulse of said first recombiner element:to trigger a gas displacement process preceding a flame front,to have a flow rate of at least 5 m/s, andto cause intensified heating of said second recombiner element and as a result an ignition of the gas flow even before the flame front reaches said second recombiner element. 2. The safety system according to claim 1, wherein the flow rate for the triggered gas displacement process is predefined as at least 10 m/s. 3. The safety system according to claim 1, wherein at least one of said recombiner elements is constructed to trigger an ignition in a circulating gas flow in a natural convection mode only when a content of entrained hydrogen is at least 6% by volume. 4. The safety system according to claim 1, wherein at least one of said recombiner elements is constructed to trigger an ignition in a circulating gas flow in a natural convection mode only when a content of entrained hydrogen is at least 8% by volume. 5. The safety system according to claim 1, which further comprises a building spray system. 6. The safety system according to claim 5, wherein said building spray system injects water as required. 7. The safety system according to claim 5, wherein said recombiner elements and said building spray system are adapted to one another to initially form high-temperature target ignition zones and then carry out steam deinerting, in the event of a response. 8. A nuclear plant, comprising a safety system according to claim 1. 9. The safety system according to claim 1, wherein said recombiner elements and said flow paths are configured such that the velocity of the gas flow of the pressure pulse is more than twice as high as the gas flow velocity present in a convection operating mode. |
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047059508 | claims | 1. A specimen-exchanging apparatus comprising a specimen chamber subjected to a vacuum, a specimen-exchanging chamber connected with said specimen chamber, said specimen-exchanging chamber being openable to ambient air, valve means for sealing and releasing the connection between said specimen chamber and said specimen-exchanging chamber means for exhausting said specimen-exchanging chamber, rotating means for rotating a feed screw means in clockwise and counterclockwise directions, said feed screw means extending within said specimen-exchanging chamber along the longitudinal coaxial direction of both said specimen-exchanging chamber and said specimen chamber, guide means disposed in parallel with said feed screw means, a removable nut means threaded through said feed screw means for moving slidably in a direction from said specimen-exchanging chamber to said specimen chamber and in an opposite direction from said specimen chamber to said specimen-exchanging chamber by a rotataional direction of said feed screw means, said nut means being formed monolithically with a moving means for removing a specimen holder mounted on said guide means and retaining a specimen for movement linearly between said specimen-exchanging chamber and said specimen chamber, and said moving means being provided with a specimen-chucking means for loading and unloading said specimen holder from said nut means. 2. A specimen-exchanging apparatus according to claim 1, wherein said specimen-chucking means includes a lever means for accomplishing a loading and unloading of said specimen holder by switching position thereof. 3. A specimen-exchanging apparatus according to claim 2, further comprising a rotatable shaft means extending through a wall of said specimen-exhanging chamber for rotating said rotating means from the outside to the inside of the specimen-exchanging chamber in a manner so as to maintain a vacuum in the specimen-exchanging chamber. 4. The specimen-exchanging apparatus according to claim 2 further including means for rotating said lever means disposed outside said specimen-exchanging chamber. |
abstract | A method for assembling a secondary collimator including a first face plate having a first surface and an opposing second surface is provided. The method includes 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. The method also includes coupling 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. |
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description | The present invention refers to a fuel assembly configured to be positioned in a nuclear water reactor, especially a heavy water reactor or a light water reactor, LWR, such as a boiling water reactor, BWR, or a pressurized water reactor, PWR, of a nuclear plant. More precisely, the present invention refers to a fuel assembly configured to be positioned in a nuclear water reactor, comprising an upstream minor portion defining an upstream end, a downstream minor portion defining a downstream end, a main portion connecting the upstream portion and the downstream portion, a plurality of elongated fuel rods arranged in parallel with a longitudinal axis extending through the upstream end and the downstream end, a flow interspace between the upstream end and the downstream end, the flow interspace being configured to permit a flow of coolant through the fuel assembly along a flow direction from the upstream end to the downstream end in contact with the fuel rods, and at least one elongated tube forming an internal passage extending through the main portion in parallel with the fuel rods and permitting a stream of the coolant through the internal passage, wherein the elongated tube comprises a bottom, an inlet to the internal passage at the upstream portion and an outlet from the internal passage at the downstream portion. The purpose of the elongated tube, permitting a stream of the coolant through the internal passage, may be to provide non-boiling moderating water in the fuel assembly, or to provide a guide for a control rod to be introduced into the fuel assembly when the operation of the nuclear water reactor is to be interrupted. The elongated tube may have different cross-sectional shapes, such as a circular shape, an oval shape, a polygonal shape, a cruciform shape, etc. US 2003/0128798 discloses a fuel assembly for a nuclear water reactor. The fuel assembly comprises a lower tie plate with a screening plate positioned below a network section in the lower tie plate cavity. The screening plate is arranged substantially horizontally so that the lower tie plate is divided into upper and lower parts by the screening plate. Tubular filters are attached to the screening plate so that the tubular filters have openings below and above the screening plate. The top ends of the tubular filters are closed. U.S. Pat. No. 6,175,606 discloses a fuel assembly for a nuclear reactor, comprising a plurality of fuel rods and a debris filter, which is provided in the flow of coolant between an upstream end and the fuel rods, and configured to catch debris particles in the coolant flow. The filter device comprises a plurality of through-holes for the coolant. The purpose of the coolant in a nuclear water reactor is to function as a cooling fluid and a moderator. It is important to secure a flow of the coolant through the fuel assembly in order to ensure a proper cooling of the fuel and a proper moderation of the neutrons. The purpose of the tubular filters and the debris filter, discussed above, is to catch debris particles in the coolant, and thus to prevent debris particles from being caught at a higher position in the fuel assembly, especially in the spacers where the debris particles may cause fretting on the cladding of the fuel rods. Fretting may result in a primary defect, a small hole through the cladding, and at a later stage in a secondary defect, i.e. break of the fuel rod, which may cause uranium leakage into the coolant. In case of a secondary defect, the operation of the reactor has to be interrupted and the failed fuel rod be replaced. Such a replacement is time-consuming and expensive. Debris particles in the coolant may of course also cause defects to other components in a nuclear plant, for instance pumps. The debris filters, used in the fuel assemblies that are marketed and sold today, are dimensioned to catch debris particles above a certain size, for instance, approximately 7 mm if the particles are wire particles. Smaller debris particles are regarded harmless and/or have a low probability to get caught in the spacers. A problem is that such smaller debris particles will circulate indefinitely, and albeit the probability to get caught is low, the debris particles will have millions of chances to get caught. There is no natural sink for such circulating smaller debris particles in the reactor. An object of the present invention is to remedy the problems discussed above and to provide a natural sink for debris particles, in particular small debris particles which are not caught by the debris filter. This object is achieved by the fuel assembly initially defined, which is characterized in that the elongated tube comprises an inlet pipe forming the inlet, that the inlet pipe has an inlet end and an outlet end, that the outlet end is located inside the internal passage at a distance from the bottom, thereby forming a space in the internal passage between the outlet end and the bottom. By means of such an inlet pipe extending into the elongated tube, debris particles, in particular small debris particles, which are brought by the coolant into the internal passage, may in an efficient manner be caught in the space immediately above the bottom of the elongated tube. When leaving the outlet end of the inlet pipe and entering the internal passage, a wake, or local sub-pressure, will be formed in the internal passage permitting any possible particles to be separated from the coolant stream and to sink towards the bottom into the space, where they will remain. They will thus be prevented from causing any defects to the fuel assembly, and to any further components of the nuclear plant. According to an embodiment of the invention, the elongated tube has an inner diameter, wherein the inlet pipe at the outlet end has an outer diameter being smaller than the inner diameter of the elongated tube. Consequently, the inlet pipe will have a flow area which is smaller than the flow area of the inner passage, and thus the velocity of the coolant stream will decrease when the coolant enters the internal passage, causing a pressure drop. In such a way said wake or local sub-pressure will be ensured. According to a further embodiment of the invention, the inlet end forms an opening which extends along or in a plane being non-parallel to the longitudinal axis. The opening of the inlet end of the inlet pipe will thus be directed towards the flow of coolant, which ensures entrance of a significant stream of coolant into the inlet pipe and the internal passage. Advantageously, the plane of the opening may extend transversally to the longitudinal axis, and thus to the flow of coolant. In particular, the plane of the opening may be perpendicular, or substantially perpendicular, to the longitudinal axis. According to a further embodiment of the invention, the inlet pipe extends through the bottom. This is advantageous from a manufacturing point of view. Furthermore, such an extension of the inlet pipe may minimize the space required for the inlet pipe outside the elongated tube, and thus any negative influence on the flow of coolant. According to a further embodiment of the invention, the space is an annular space around the inlet pipe. According to a further embodiment of the invention, the inlet pipe is concentric with the elongated tube. According to a further embodiment of the invention, the elongated tube comprises a bottom end plug forming said bottom, wherein the inlet pipe extends through the bottom end plug. The inlet pipe and the bottom end plug may thus form a common element, which may be attached to the end of the elongated tube in a convenient manner. According to a further embodiment of the invention, the outlet end is located a distance of at least 0.2 m downstream the bottom. Thus the space will have a height of at least 0.2 m, which is regarded to be sufficient to create a sufficient sub-pressure to permit possible debris particles to sink to the bottom of the elongated tube. Preferably, the distance is at least 0.3 m, more preferably at least 0.4 m and most preferably at least 0.5 m. According to a further embodiment of the invention, the elongated tube is cylindrical. According to a further embodiment of the invention, the elongated tube may have a circular cross-section. However, the elongated tube may have any possible cross-section, such as an oval cross-section, a square or rectangular cross-section, a triangular cross-section, a polygonal cross-section, or a cruciform cross-section. According to a further embodiment of the invention, the elongated tube comprises at least one magnet provided to attract magnetic material towards the bottom. Such at least one magnet may comprise a permanent magnet, but also an electromagnet may be possible. According to a further embodiment, the at least one magnet may comprise one, two, three, four or even more magnets arranged at the bottom of the elongated tube. According to a further embodiment of the invention, the fuel assembly comprises a debris filter at or in the upstream minor portion upstream the fuel rods. Such a debris filter will catch all debris particles above a certain size. According to a further embodiment of the invention, the inlet end is located upstream the debris filter. In this case, the inlet pipe will thus extend through the debris filter. All debris particles may thus be collected in the space of the elongated tube. According to a further embodiment of the invention, the inlet end is located downstream the debris filter. In this variant, only debris particles not caught by the debris filter, i.e. relatively small debris particles, will be collected in the space of the elongated tube. According to a further embodiment of the invention, the fuel assembly is configured to be positioned in a boiling water reactor, wherein the elongated tube comprises a water rod for conveying non-boiling water through the internal passage. According to a further embodiment of the invention, the fuel assembly comprises at least two elongated tubes each comprising a water rod for conveying non-boiling water through the respective internal passage. For instance, the fuel assembly may comprise three elongated tubes, each comprising a water rod for conveying non-boiling water through the respective internal passage. According to a further embodiment of the invention, the fuel assembly is configured to be positioned in a pressure water reactor, wherein the elongated tube comprises a guide tube for receiving a control rod. For instance, the fuel assembly may comprise a plurality of elongated tubes each comprising a guide tube for receiving a respective control rod. FIG. 1 discloses a first embodiment of a fuel assembly 1 configured to be positioned in a nuclear water reactor, and more precisely a boiling water reactor, BWR. The fuel assembly 1 has an elongated shape and extends along a longitudinal axis x between an upstream end 1a and a downstream end 1b of the fuel assembly 1. During normal use of the fuel assembly 1 in the reactor, the upstream end 1a forms a lower end, whereas the downstream end 1b forms an upper end of the fuel assembly 1. A plurality of elongated fuel rods 2 are arranged in parallel with the longitudinal axis x. Each fuel rod 2 comprises a cladding tube and nuclear fuel (not disclosed), for instance in the form of fuel pellets, contained in the cladding tube. The fuel rods 2 may comprise full length fuel rods and part length fuel rods. The part length fuel rods may for instance have a length being ⅓ and/or ⅔ of the length of the full length fuel rods. Only full length fuel rods are shown in FIG. 1. The fuel assembly 1 comprises an upstream minor portion 3 defining the upstream end 1a, and a downstream minor portion 4 defining the downstream end 1b. A main portion 5 is provided between and connects the upstream minor portion 3 and the downstream minor portion 4. The borders between the main portion 5 and the upstream minor portion 3, and between the main portion 5 and the downstream minor portion 4 are indicated by dashed transversal lines in FIG. 1. The main portion 5 has an axial length being longer than the axial length of each of the upstream minor portion 3 and the downstream minor portion 4. The axial length of the main portion 5 may also be longer than the sum of the axial lengths of the upstream minor portion 3 and the downstream minor portion 4. The fuel assembly 1 defines a flow interspace 6 between the upstream end 1a and the downstream end 1b. The flow interspace 6 is configured to permit a flow of coolant through the fuel assembly 1 along a flow direction F from the upstream end 1a to the downstream end 1b in contact with the fuel rods 2. In the first embodiment, the fuel assembly 1 also comprises two elongated tubes 7 forming a respective internal passage 8, see FIG. 2, extending through the main portion 5 in parallel with the fuel rods 2. The elongated tubes 7 permit a stream of the coolant through the respective internal passage 8. In the first embodiment, the elongated tubes 7 form so called water rods enclosing the stream of coolant, especially of non-boiling water. It should be noted that the number of elongated tubes 7 may be another than two, for instance one, three, four, five, six or more. In the first embodiment, the elongated tubes 7 are cylindrical. In particular, the elongated tubes 7 may have a circular cross-section. The fuel rods 2 are held by means of spacers 9 in a manner known per se. In the first embodiment, the spacers 9 are attached to the elongated tubes 7. The fuel assembly 1 of the first embodiment also comprises a casing 10 enclosing the fuel rods 2, the spacers 9 and the flow interspace 6. The elongated tubes 7 are attached to a bottom plate 11 provided beneath the fuel rods 2. Furthermore, the elongated tubes 7 may also be attached to a top plate 12 at the downstream end 1a. The top plate 12 comprises a handle 13. The bottom plate 11, the elongated tubes 7, the top plate 12 and the spacers 9 form a support structure. The support structure may be lifted via the handle 13 and carries the weight of the fuel rods 2. In the first embodiment, each elongated tube 7 comprises a tube part 7a attached to the bottom plate 11, and a solid part 7b forming a massive rod attached to the top plate 12 as can be seen in FIG. 1. The fuel assembly 1 comprises a bottom piece 14, frequently designated as a transition piece. The bottom piece 14 extends to the upstream end 1a and defines an inlet opening 15 for the flow of coolant. The bottom piece 14 may be attached to the bottom plate 11 or to the casing 10. The fuel assembly 1 comprises a debris filter 16 at or in the upstream minor portion 3 upstream the fuel rods 2. The debris filter 16 is provided between the upstream end 1a and the fuel rods 2. In the first embodiment, the debris filter 16 is provided between the upstream end 1a and the bottom plate 8. The debris filter 16 may be supported by or attached to the bottom piece 11. In the first embodiment, each of the elongated tube 7 comprises a bottom 20, an inlet 21 to the internal passage 8 at the upstream minor portion 3 and an outlet 22 from the internal passage 8 at the downstream minor portion 4. The outlet 22 is located at the downstream end of the tube part 7a of the elongated tube 7. The elongated tube 7 also comprises an inlet pipe 23, which forms the inlet 21 to the internal passage 8, as can be seen in FIG. 2. Advantageously, the flow area of the inlet 21 is smaller than the flow area of the outlet 22. The inlet pipe 23 has an inlet end 24 and an outlet end 25. The outlet end 25 is located inside the internal passage 8 at a distance from the bottom 20. In such a way, a space 26 is formed in the internal passage 8 between the outlet end 25 and the bottom 20. The elongated tube 7 has an inner diameter D. The inlet pipe 23 at the outlet end 25 has an outer diameter d, which is smaller than the inner diameter D of the elongated tube 7. The inlet end 24 of the inlet pip 23 forms an opening which extends along a plane p which is non-parallel to the longitudinal axis x. In the first embodiment, the plane p is perpendicular to the longitudinal axis x. In the first embodiment, the inlet pipe 23 extends through the bottom 20 of the elongated tube 7, and especially concentrically through the bottom 20. In the first embodiment, the space 26 mentioned above is an annular space around the inlet pipe 23 of the elongated tube 7. The elongated tube 7 comprises a bottom end plug 27, which forms the bottom 20. The inlet pipe 23 thus extends through the bottom end plug 27. The inlet pipe 23 and the bottom end plug 27 mays form a common element, which is attached to the end of the elongated tube 7, for instance by being inserted and welded. The outlet end 25 of the inlet pipe 23 is located at a distance from the bottom 20. In particular, the outlet end 25 is located 0.2 m, or at least 0.2 m, downstream the bottom 20. More specifically, the distance downstream the bottom 20 may be at least 0.3 m, at least 0.4 m or at least 0.5 m. The distance downstream the bottom 20 may be at the most 1 m. Furthermore, the elongated tube 7 may optionally comprise at least one magnet 28 provided to attract magnetic material, i.e. magnetic debris particles, towards the bottom 20. In the first embodiment, two magnets 28 are indicated, but one, three, four, five or oven more magnets may be provided close to the bottom 20. The magnets are located in the space 26. The magnets 28 may be permanent magnets. The magnet or magnets 28 may have any suitable shape, for instance one annular magnet provided around the inlet pipe 28 would be possible. In the first embodiment, the inlet end 24 of the inlet pipe 23 is located upstream the debris filter 16. This is illustrated for the inlet pipe 23 to the right in FIG. 1. In a second embodiment, the inlet end 24 of the inlet pipe 23 is located downstream the debris filter 16. This is illustrated for the inlet pipe 23 to the left in FIG. 1. In the second embodiment, the inlet end may be located at an upstream side of the bottom plate 11. FIG. 3 illustrates a third embodiment, which differs from the first embodiment in that the inlet pipe 23 extends eccentrically in the internal space 8 of the elongated tube 7. Furthermore, the inlet end 24 of the inlet pipe 23 is located at the upstream surface of the bottom plug 27. The inlet end 24 is thus located downstream the debris filter 16, and possibly also downstream the bottom plate 11 depending on how the elongated tube 7 is attached to the bottom plate 11. FIG. 4 illustrates a fourth embodiment, which differs from the first embodiment in that the inlet pipe 23 is curved or angled. The inlet pipe 23 extends through the side of the elongated tube 7 into the internal passage 8, as can be seen in FIG. 4. The space 26 is thus not annular. FIG. 5 illustrates a fifth embodiment, which differs from the first embodiment in that the inlet pipe 23 is curved or angled. However, the inlet pipe 23 extends eccentrically through the bottom 20, and through the bottom plug 27. Furthermore, the outlet end 25 of the inlet pipe 23 extends vertically, or substantially vertically. It is to be noted, the configuration of the inlet pipe 23 may be further modified. Especially, all the various features of the inlet pipe 23 and its attachment to the elongated tube 7 shown in the first to fifth embodiments may be combined with each other. FIG. 6 illustrates a sixth embodiment, which differs from the first embodiment in that the fuel assembly 1 is configured to be positioned in a pressure water reactor. It should be noted, that the same reference signs have been used for similar or corresponding elements in all the embodiments disclosed. Also the fuel assembly 1 according to the sixth embodiment has an elongated shape and extends along a longitudinal axis x between an upstream end 1a and a downstream end 1b of the fuel assembly 1. During normal use of the fuel assembly 1 in the reactor, the upstream end 1a forms a lower end and the downstream end 1b forms an upper end of the fuel assembly 1. A flow interspace 6 is provided between the upstream end 1a and the downstream end 1b. A plurality of fuel rods 2 is provided in the flow interspace 6 between the upstream end 1a and the downstream end 1b. The fuel rods 2 are held by means of spacers 9. In the sixth embodiment, the spacers 9 are attached to a number of elongated tubes 37, two of which are shown in FIG. 6. The elongated tubes 37 form a respective internal passage 8 extending through a main portion 5 in parallel with the fuel rods 2 and permitting a stream of the coolant through the internal passage 8. Each elongated tube 37 forms a so called guide tube configured to receive a control rod when the operation of the PWR is to be interrupted. The control rods are introduced into the elongated tubes 37 from the downstream end 1b, i.e. from above. In contrast to a fuel assembly 1 for a BWR, the fuel assembly 1 according to the sixth embodiment has no casing, but still comprises the flow interspace 6 as explained above. The elongated tubes 37 are attached to a bottom plate 11 provided beneath the fuel rods 2, and to a top plate 12 at the downstream end 1a. The bottom plate 11, the elongated tubes 7, the top plate 12 and the spacers 9 form a support structure which carries the weight of the fuel rods 2. The fuel assembly 1 also comprises a bottom piece 14. The bottom piece 14 extends to the upstream end 1a and defines the inlet 15 for the flow of coolant. The bottom piece 14 may be attached to the bottom plate 12. The fuel assembly 1 of the sixth embodiment also comprises a debris filter 16 at the upstream minor portion 3 upstream the fuel rods 2. As explained above in connection with the previous embodiments, each of the elongated tubes 37 according to the sixth embodiment comprises a bottom 20, an inlet 21 to the internal passage 8 at the upstream minor portion 3 and an outlet 22 from the internal passage 8 at the downstream minor portion 4. The elongated tube 37 also comprises an inlet pipe 23, which forms the inlet 21 to the internal passage 8, as can be seen in FIG. 2. Advantageously, the flow area of the inlet 21 is smaller than the flow area of the outlet 22. The inlet pipe 23 has an inlet end 24 and an outlet end 25. The outlet end 25 is located inside the internal passage 8 at a distance from the bottom 20. In such a way, a space 26 is formed in the internal passage 8 between the outlet end 25 and the bottom 20. It should be noted, that the configuration of the inlet pipe 23 and the arrangement of the inlet pipe 23 in the elongated tube 37 in the sixth embodiment may be as in the first to fifth embodiments disclosed in FIGS. 2-5 and discussed above. The present invention is not limited to the embodiments disclosed, but may be varied and modified within the scope of the following claims. |
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051924960 | abstract | The fuel assembly has fuel arrangement of a square lattice of 10 lines by 10 rows. Four water rods having a large diameter are arranged in central region of horizontal cross section wherein arrangement of 12 fuel rods is possible. The four water rods having a large diameter are so arranged adjacently in a circle as to form an internal between each other. First coolant flow path which is extended toward axial direction is formed by surrounded with the water rods having a large diameter. The first coolant flow path leads to third coolant flow path which is formed around the fuel rods through second coolant flow path which is the interval between the water rods having a large diameter.. The fuel assembly is able to optimize the moderator to fuel atom number density ratio and to reduce the pressure loss because a portion to be an excessively moderated region is utilized as the first coolant flow path. |
046648828 | claims | 1. A continuous segmented fuel and moderator rod for use with a water cooled and moderated nuclear fuel assembly, said rod comprising: (a) a lower fuel region containing a column of nuclear fuel; (b) a moderator region, disposed axially above the fuel region, said moderator region having means for admitting and passing said water moderator therethrough for moderating an upper portion of the nuclear fuel assembly, said moderator region being separated from said fuel region by a water tight separator. (a) a fuel bundle including a plurality of elongate nuclear fuel rods held in a spaced lateral array, each of said fuel rods being positioned in one of a plurality of predetermined lattice positions defined by an upper and a lower tie plate; (b) a flow channel for enclosing said fuel bundle; (c) a base member, secured to said flow channel and having means for supporting said fuel bundle within said flow channel, said base member further including nozzle for admitting moderator/coolant into said fuel assembly; (d) at least one continuous segmented fuel and moderator rod disposed in said fuel bundle in one of said fuel rod lattice positions, said segmented rod including a lower fuel region containing a column of nuclear fuel and an upper moderator region separated from said lower fuel region by a water tight seal, said moderator region having an aperture means for passing said moderator/coolent into and through said moderator region for improving an H/U ratio of an upper portion of said fuel assembly. (a) passing moderator/coolant through a fuel bundle of said fuel assembly; (b) providing at least one continuous, fuel and moderator rod having an upper moderator region and a lower fuel region; (c) passing a portion of said moderator/coolant through said moderator region of the segmented rod while maintaining said moderator at a substantially constant void fraction in the range of between 0-20% to thereby improve the H/U ratio at said upper portion of the fuel assembly. 2. The segmented rod of claim 1 further comprising a cladding for containing said lower fuel region and said moderator region, and said lower fuel region further containing a fission gas plenum region disposed between said column of neclear fuel and said moderator region. 3. The segmented rod of claim 2 wherein said separator comprises a disk-shaped separator affixed to said cladding to separate said moderator region from said fuel region. 4. The segmented rod of claim 2 wherein said means for admitting and passing comprises at least one inlet hole in the cladding of said moderator region. 5. The segmented rod of claim 4 wherein said means for admitting and passing further comprises at least one outlet hole in said cladding near the top of said moderator region for passing flowing moderator out of said moderator region. 6. The segmented rod of claim 4 wherein said moderator comprises subcooled water and said at least one inlet hole in sized and positioned to admit subcooled water having a void fraction of between about 0-20%. 7. The segmented rod of claim 5 wherein said at least one outlet hole is positioned to discharge moderator into a region of said nuclear fuel assembly where the void fraction is between about 20-70%. 8. A nuclear fuel assembly for a boiling water reactor (BWR) core comprising: 9. The nuclear fuel assembly for claim 8 wherein said segmented rod further comprises a cladding for containing said column of nuclear fuel and said moderator region and an upper end plug for mating an upper end of said segmented rod with said upper tie plate and a lower end plug for sealing the bottom of said cladding and for mating a lower end of said segmented rod with said lower tie plate. 10. The nuclear fuel assembly of claim 8 wherein said segmented rod further comprises a fission gas plenum region disposed between said column of nuclear fuel and said moderator region and separated from said moderator region by said water tight seal. 11. The nuclear fuel assembly of claim 10 wherein said water tight seal comprises a disk welded to said cladding and wherein said aperture means comprises at least one inlet flow aperture disposed above said disk. 12. The nuclear fuel assembly for claim 11 wherein said at least one inlet flow aperture is axially located within said fuel assembly at a position where a void fraction of the coolant is between about 0-20%. 13. The nuclear fuel assembly of claim 11 wherein said aperture means further comprises at least one flow outlet aperture for passing moderator out of said moderator region of said segmented rod. 14. A method of moderating a normally undermoderate upper portion of a boiling water reactor (BWR) fuel assembly comprising the steps of: |
claims | 1. A radiological image capturing system, comprising:a radiological image capturing apparatus that includes an X ray tube to emit X rays, a subject placing plate to place a subject thereon, a plurality of gratings, and an X-ray detector to detect the X rays passed through the plurality of gratings, so as to capture an X-ray image of a subject to output X-ray image data representing the X-ray image of the subject;a controlling apparatus that is structured to determine whether or not a distortion is generated on any one of the gratings based on at least one of Moiré stripe images captured by detecting the X rays passed through the plurality of gratings by employing the X-ray detector in such a state that no subject is placed on the subject placing plate;an image processing apparatus that is structured to apply an image processing to the X-ray image data, outputted by the radiological image capturing apparatus, so as to create and output processed X-ray image data; andan image outputting apparatus structured to output a processed X-ray image based on the processed X-ray image data outputted by the image processing apparatus;wherein the Moiré stripe images includes a first Moiré stripe image and a second Moiré stripe image that is captured after the first Moiré stripe image is captured; andwherein the controlling apparatus is structured to determine whether or not the distortion is generated on any one of the gratings based on both the first Moiré stripe image and the second Moiré stripe image, while, the image processing apparatus is structured to conduct an operation for correcting the X-ray image data representing the X-ray image that is captured in a state that the subject is actually placed on the subject placing plate, based on the second Moiré X-ray image data. 2. The radiological image capturing system of claim 1,wherein the plurality of gratings includes such a grating that is structured to convert the X rays emitted from the X ray tube into multi-radiant sources. |
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description | The present invention relates to a novel material based on uranium, gadolinium and oxygen exhibiting a crystalline phase enriched in gadolinium. Such a material is particularly advantageous in the context of its use as burnable neutron poison in a nuclear fuel element, for example used in light water nuclear reactors. As in any type of industry, the production of nuclear electricity has to respond to economic realities. For nuclear power stations without inline reloading of fuel, as is the case with light water nuclear reactors (LWRs), such as, for example, pressurized water reactors (PWRs), reducing the production costs involves lengthening the operating campaigns of the reactor. Thus, the economically damaging effect of shutdowns in order to reload with fuel and of the maintenance time of the power stations may be limited. The elongation of the duration of operation, in other words the use of long cycles, requires, however, an additional reserve of reactivity of the fuel, that is to say an increase in the enriching of the starting fuel. However, it is imperative to be able to compensate for this increase in the reactivity of the fuel by an excess of negative reactivity at the start of the cycle. Currently, this increased need for negative reactivity, in particular in pressurized water reactors, is essentially provided by the presence of boron in the form of boric acid, dissolved at variable concentrations in the coolant or moderator of the primary circuit. The boron acts as poison for the neutrons. The homogeneous distribution of the boron in the core exhibits the advantage of not interfering with the power distribution of the nuclear reactor. On the other hand, the increase in the content of boron dissolved in the primary circuit is capable of causing some problems, in particular in terms of deterioration in the coefficient of reactivity of the moderator, of corrosion and of safety of the plant, as set out in the document FR 2 789 404, and of increase in the production of tritium. In fact, the dissolved boron is capable of dilating during an increase in temperature, thus inducing a positive contribution to the coefficient of reactivity of the reactor (αm). The amount of dissolved boron thus absolutely has to be maintained below a maximum limit, in order to observe the criterion of a negative coefficient of reactivity of the moderator (αm<0) under all the conditions of operation of the reactor. Furthermore, the introduction of additional amounts of boric acid H3BO3 may bring about problems of direct corrosion but also, as explained in the document FR 2 789 404, of indirect corrosion brought about, for example, by the lithium introduced as pH control agent for compensating for the amount of boric acid. Moreover, the risk of untimely dilution of the boron is regarded as one of the main initiators of reactivity insertion accident or “RIA” during safety studies on reactors. Finally, neutron activation reactions on boron constitute the main source of production, in the primary circuit, of tritium (10B+n→3H+24He), an undesirable radionuclide element, and thus the aim is to reduce the discharges to the environment for reasons of safety and of radiation protection. Consequently, for the purpose of reducing the amount of boron necessary for the control of the reactivity, in particular the cycle start, various burnable poisons, other than soluble boron, or used in conjunction with the latter, have been envisaged. The term “neutron poison” is understood to mean an element having a high power of capturing neutrons and used to compensate, at least in part, for the excess reactivity of fissile environments. Reference is made to “burnable” neutral poison, also known as “temporary neutron-absorbing or -capturing material”, to denote a poison which gradually disappears during the operation of the nuclear reactor. These poisons are generally based on gadolinium, erbium, samarium, europium or other isotope which, subsequent to neutron capture, produces an isotope of low effective absorption cross section. Among burnable poisons, gadolinium in the solid state is the most widely used. Advantageously, as the density of gadolinium varies only very slightly with temperature, it does not bring about a positive contribution to the coefficient of reactivity of the moderator αm. Of the 25 known isotopes of gadolinium (from 142Gd to 164Gd), only seven are stable. Among them, the most absorbent are 155Gd and 157Gd, these two isotopes representing nearly 100% of the absorption related to natural gadolinium. Subsequent to the neutron capture, the 155 and 157 isotopes of gadolinium are transmuted into 156 and 158 isotopes respectively, these two isotopes being, on the other hand, of very low effective cross sections. This property makes gadolinium a burnable poison of choice. On the other hand, unlike boron, the concentration of gadolinium in the solid state in the fuel cannot be controlled by an external system. The gadolinium diminishes and disappears with the consumption of the fuel. To date, gadolinium is generally used in the oxide form, Gd2O3, mixed in appropriate proportions with uranium oxide UO2, or a mixture of fissile materials, in order to form fuel pellets of a certain number of rods in a nuclear fuel assembly. Two major forms of inclusion of gadolinium in nuclear ceramics are distinguished: homogeneous inclusions and heterogeneous inclusions. In the context of homogeneous inclusions, mention may in particular be made of homogeneous pellets consisting of a (U,Gd)O2 solid solution. However, the Gd2O3 contents of these pellets do not exceed 20% by weight, in order to remain within the domain of the FCC (face-centered cubic) solid solution of the UO2—Gd2O3 system. By way of example, the document FR 2 536 571 describes the use, in uranium oxide pellets, of gadolinium oxide Gd2O3 as neutron-absorbing material, in a percentage by weight of less than 12%. The solid solutions of the UO2—Gd2O3 system show, however, a large decrease in the thermal conductivity with respect to the UO2, which makes it necessary to decrease the generation of power by reducing the enriching in 235U and consequently increases the penalty with regard to the generation of energy during the fuel cycle. As regards the forms of heterogeneous inclusions, gadolinium may be distributed in the nuclear fuel pellets in the form of macromasses or also be present therein according to a specific radial distribution. A distribution of the macromass type of Gd2O3, as envisaged by Balestrieri [1], exhibits the advantage of a higher thermal conductivity than that obtained with a UO2—Gd2O3 solid solution mentioned above. However, gadolinium oxide Gd2O3 is liable to present problems of incompatibility with the UO2 (in particular in terms of anisotropic expansion) and of solubility in water (under typical operating conditions of a pressurized water reactor, for example), which hinders its use in UO2 pellets. Furthermore, several alternative forms of radial distribution of gadolinium in fuel pellets, in or not in combination with other burnable poisons, have already been provided. Thus, the document U.S. Pat. No. 4,671,927 describes the use, in pellets for nuclear fuel rods, of a hybrid mixture of Gd2O3 (from 1 to 20% by weight) and boron carbide particles. The document U.S. Pat. No. 4,587,087 provides, for its part, nuclear fuel pellets comprising a core of fissile material, for example of uranium oxide, coated with a first layer comprising mainly boron, optionally in combination with other burnable poisons, and with a second layer of hydrophobic material formed mainly of niobium. The document U.S. Pat. No. 3,122,484 for its part employs a layer of cadmium, gadolinium or samarium at the surface of UO2 plates (MTR type). Also, the document U.S. Pat. No. 4,668,468 provides for the use, in a few rods of the assembly, of pellets exhibiting radial poisoning according to different alternative distribution forms, so as to minimize the amount of neutron poison necessary for the control of the reactivity. It more particularly describes pellets comprising, either in their internal region or in their external region, a homogeneous mixture of gadolinium and enriched uranium. Likewise, in the document U.S. Pat. No. 4,678,629, pellets are provided which exhibit a cylindrical internal part comprising from 4 to 8% by weight of Gd2O3 mixed with natural or depleted uranium and an annular external part formed of UO2 enriched in 235U. Finally, all the processing alternative forms presented in these documents use Gd2O3 without chemical combination with UO2 (Gd2O3 macromasses, for example) or a mixture of UO2 and Gd2O3 in which the gadolinium oxide Gd2O3 does not exceed 20% by weight of the mixture. These inclusion forms become known materials of the U—Gd—O phase diagram: UO2, (U,Gd)O2 solid solutions and Gd2O3. There indeed exists another phase listed in the literature [2] [3] of the U—Gd—O system, which phase is enriched in gadolinium. This phase Gd6UO12, better known under the name “Delta” phase, has a crystallographic structure of the rhombohedral type and belongs to the R3 space group (it may also be indexed under a hexagonal structure). Two types of methods for obtaining the Gd6UO12 phase at the laboratory scale are reported in the literature: starting from comilled UO2 and Gd2O3 powders [2], or starting from U3O8 [3]. Unfortunately, the Delta phase proves to be unstable under high-temperature sintering conditions typically employed for the preparation of the UO2 pellets. It is thus not possible to envisage its use as burnable poison of nuclear fuel pellets. Currently, an increase in the enriching of the fuels is envisaged by the main designers/constructors/operators of nuclear reactors, and also the limitation of/reduction in the use of boron. Consequently, in the light of this increase in enriching and of the problems described above of the current techniques for supplying negative reactivity, a need remains to optimize the poisoning of UO2 ceramics of nuclear assemblies with gadolinium, so as to make possible control of the reactivity under normal operating conditions of a nuclear reactor, while reducing, indeed even while suppressing, the use of boron in the reactor. The present invention is targeted specifically at providing a novel U—O—Gd material exhibiting a crystalline phase enriched in gadolinium and compatible with its use as burnable neutron poison of a nuclear fuel element, for example for pellets. Thus, the present invention relates, according to a first of its aspects, to a material based on uranium (U), gadolinium (Gd) and oxygen (O) exhibiting a crystalline phase with a crystallographic structure of cubic type, with a Gd/[Gd+U] atomic ratio ranging from 0.6 to 0.93, the uranium being present therein in the +IV and/or +V oxidation state. Surprisingly, the inventors have discovered that it is possible to access a U—O—Gd material exhibiting a crystalline phase rich in gadolinium and stable under the sintering conditions of a fuel pellet. As described in detail in the continuation of the text, the inventors have more particularly demonstrated two different crystalline phases, which will be referred to in the continuation of the text as “phase C1” and “phase C2”, and also a two-phase domain which is a mixture of these two crystalline phases. The material according to the invention may advantageously be used as burnable neutron poison of a nuclear fuel element. In fact, as described in detail in the continuation of the text, it is produced under the same sintering conditions as the uranium oxide; it is cosinterable with UO2, which allows it to be used in nuclear fuel pellets. Thus, according to another of its aspects, the present invention is targeted at the use of a material as defined above as burnable neutron poison of a nuclear fuel element, for example for a light water or heavy water nuclear reactor. The material according to the invention may thus be employed in pellets used for nuclear fuel assembly rods or also in nuclear fuels of plate type. As illustrated in the examples which follow, it proves to be possible, by adjusting the number of rods according to the invention, the proportion of burnable poison according to the invention or the pellets constituting them, and/or the isotopic vector of the gadolinium, to design nuclear fuel assemblies which make it possible to approach the behavior of an “ideal” reactor, in other words to optimally operate the reactor. In particular, the change in the reactivity of the reactor may be better controlled with the use of pellets in accordance with the invention than with conventional homogeneous pellets composed of a (U,Gd)O2 solid solution. Furthermore, advantageously, the use of the material according to the invention makes it possible to reduce, indeed even to dispense completely with, the use of other absorbing/neutron poisons. In particular, it makes it possible to reduce, indeed even to completely suppress, the use of boron dissolved in the primary cooling circuit of the reactor and/or dissolved in the moderator. The use of the material according to the invention as burnable neutron poison thus makes it possible to overcome the disadvantages, discussed above, brought about by recourse to large amounts of boron, in particular to reduce the problems of corrosion, of tritium production and of risk of reactivity accident. Other characteristics, advantages and forms of application of the material according to the invention, will more clearly emerge on reading the detailed description which will follow of the implementation examples of the invention and on examining the appended drawings. In the continuation of the text, the expressions “between . . . and . . . ”, “ranging from . . . to . . . ” and “varying from . . . to . . . ” are equivalent and are understood to mean that the limits are included, unless otherwise mentioned. Unless otherwise mentioned, the expression “comprising a(n)” should be understood as “comprising at least one”. Material of the Invention As mentioned above, the material based on uranium (U), gadolinium (Gd) and oxygen (O) according to the invention is characterized by a crystalline phase with a crystallographic structure of cubic type, with a Gd/[Gd+U] atomic ratio ranging from 0.6 to 0.93, the uranium being present therein in the +IV and/or +V oxidation state. According to a first alternative embodiment, the material according to the invention exhibits a crystalline phase, subsequently denoted “cubic 1” or “C1” phase, the Gd/[Gd+U] atomic ratio of which is between 0.79 and 0.93. This C1 phase more particularly exhibits a crystallographic structure of cubic type with a unit cell parameter (a1), close to that of c-Gd2O3 (unit cell parameter of approximately 10.83 Å), between 10.8 and 10.9 Å. According to a second alternative embodiment, the material according to the invention exhibits a crystalline phase, subsequently denoted “cubic 2” or “C2” phase, the Gd/[Gd+U] atomic ratio of which is between 0.6 and 0.71. This C2 phase more particularly exhibits a crystallographic structure of cubic type with a unit cell parameter (a2), close to that of UO2 (unit cell parameter of approximately 5.47 Å), between 5.3 and 5.5 Å. According to yet another alternative embodiment, the material according to the invention is of two-phase type, exhibiting both a cubic 1 phase and a cubic 2 phase as defined above. In other words, the material according to this third alternative form exhibits a crystalline phase with an overall Gd/[Gd+U] atomic ratio strictly of greater than 0.71 and strictly of less than 0.79. The uranium present in a material of the invention, according to one or other of the abovementioned alternative forms, may be natural uranium (mixture of 234U, 235U and 238U). According to another specific embodiment, it may be uranium, the natural isotopic composition of which is modified, in particular uranium isotopically enriched in 235U or uranium isotopically depleted in 235U. Likewise, the gadolinium present in a material according to the invention may be natural gadolinium (152Gd/154Gd/155Gd/156Gd/157Gd/158Gd/160Gd mixture). According to another specific embodiment, it may be gadolinium, the natural isotopic composition of which is modified in its 155Gd/Gdtotal ratio and/or in its 157Gd/Gdtotal ratio, in particular with an increased 155Gd and/or 157Gd content in comparison with natural gadolinium. By way of example, the gadolinium of the material according to the invention may exhibit the following isotopic vectors: 100% 155Gd; 50% 155Gd+50% 157Gd. As illustrated in the following example 4 and in FIGS. 7.a and 8, it is possible, by varying the isotopic vector of gadolinium, more specifically the 155Gd/Gdtotal and/or 157Gd/Gdtotal isotopic ratios, to obtain controlled kinetics of exhaustion of the gadolinium. Preparation of the Material According to the Invention According to another of its aspects, the present invention relates to a process for the preparation of a material as defined above comprising a stage of sintering, at a temperature ranging from 1200 to 2200° C. and under a reducing atmosphere, a powder formed of a mixture of uranium oxide and gadolinium oxide (Gd2O3) in proportions such that the gadolinium is present in the final powder in a Gd/[Gd+U] atomic ratio ranging from 0.6 to 0.93. The uranium oxide may more particularly be uranium dioxide (UO2) or a higher oxide, such as U3O8 (triuranium octaoxide). According to a specific embodiment, the powder is formed by mixing a first uranium dioxide UO2 powder and a second gadolinium oxide Gd2O3 powder in a Gd2O3/(UO2+Gd2O3) ratio by weight of greater than or equal to 40% by weight. Of course, it is up to a person skilled in the art to adjust the proportions of uranium oxide and gadolinium oxide employed in order to obtain the desired proportion of gadolinium in the final powder. The sintering stage may be carried out on the powder, prior to the use thereof in a nuclear fuel element, for example for a pellet. Alternatively, this sintering stage may be carried out during the preparation of the nuclear fuel element. In particular, in the case of the preparation of a pellet according to the invention, as described in detail in the continuation of the text, the sintering may be carried out on the powder deposited in the form of a slip at the surface of a pellet comprising at least one fissile, indeed even fertile, material (crude pellet pressed simply by compaction, or presintered pellet) by overall sintering of the pellet. The material of the invention is then formed directly on the pellet, on completion of the sintering. The stage of sintering under a reducing atmosphere may be carried out by any technique known to a person skilled in the art. It may be carried out by heating the powder at a temperature ranging from 1200 to 2200° C., in particular from 1600 to 1800° C. “Reducing atmosphere” is understood to mean an atmosphere exhibiting an oxygen potential pO2, of less than −300 kJ/mol, in particular between −550 kJ/mol and −300 kJ/mol, during the sintering. The reducing atmosphere may more particularly be an atmosphere incorporating hydrogen, for example an argon atmosphere (with potentially a few ppm of O2 impurity) to which 5 mol % of hydrogen has been added. Generally, the duration of the sintering may be greater than or equal to 1 hour, in particular ranging from 3 to 8 hours. Applications of the Material as Burnable Poison As specifically indicated, the material according to the invention finds a particularly advantageous application as burnable neutron poison of a nuclear fuel element. It may, for example, be used to control the reactivity in the context of the operation of light water nuclear reactors (LWRs), for example for pressurized water reactors or boiling water reactors, or also for heavy water (deuterium oxide D2O) reactors. Generally, “nuclear reactor” is understood to mean the usual sense of the term to date, namely power stations for the production of energy from nuclear fission reactions using fuel elements in which fission reactions take place which release the calorific power, the latter being extracted from the elements by heat exchange with a heat-exchange fluid which provides for the cooling thereof. The material may thus be employed in nuclear fuel elements conventionally encountered in nuclear plants of the type consisting of rods formed of a plurality of pellets stacked on one another or also of plate type, as described more specifically in the continuation of the text. Of course, the invention is in no way limited to the specific alternative embodiments described below. Nuclear Fuel Pellet The present invention relates, according to another of its aspects, to a nuclear fuel pellet comprising a material as defined above. As mentioned above, the pellets denote ceramic fuel elements of cylindrical shape, the stacking of which in a cladding tube forms a rod of a nuclear assembly. According to a particularly preferred embodiment, a pellet according to the invention is a heterogeneous pellet, as represented in FIG. 1, formed of at least an internal part, in particular cylindrical (1), comprising at least one fissile, indeed even fertile, material and coated with an annular external part (2) formed in all or part of a material according to the invention. Preferably, the pellet of the invention is of cylindrical form, like the pellets conventionally encountered in nuclear fuel rods. Other forms may, of course, be envisaged, for example an overall elliptical form, as described in the application FR 2 953 637. A pellet of the invention may exhibit the dimensions of conventional pellets. For example, it may have a radius between 3.8 mm and 4.4 mm (for example: 4.05 to 4.25 mm) and a height between 3 and 20 mm, typically between 12 and 16 mm. According to a specific embodiment, the annular external part (2) exhibits a thickness (e) ranging from 0.05 to 7.5% of the total radius (R) of said pellet, in particular ranging from 1 to 3.5%. The annular external part may thus exhibit a thickness (e), measured along the axis of the radius of the pellet, between 2 and 300 μm, in particular between 30 and 250 μm. As illustrated in example 4 and in FIGS. 7.a and 8, the thickness of the annular external part (2) and thus the proportion of burnable neutron poison according to the invention of the pellet may advantageously be adjusted in order to control as best as possible the change in the reactivity of the reactor and to approach an optimum change. “Fertile” material is understood to mean a material composed of fertile atoms, in other words of atoms, the nucleus of which may be converted, directly or indirectly, into a fissile nucleus by capture of neutrons. A fertile material may, for example, be 238U. “Fissile” material is understood to mean a material, the nuclei of which are capable of undergoing fission by absorption of neutrons, such as, for example, 235U. The cylindrical internal part (1), also known as “core”, of the pellets according to the invention may be formed in all or part of uranium oxide (UO2), plutonium oxide (PuO2), thorium oxide (ThO2) or a mixture of these fissile materials, such as, for example, (U, Pu)O2. According to a specific embodiment, the core of the pellets is formed of uranium oxide. Just as described above, the uranium may be natural uranium or uranium, the isotopic vector of which is modified, for example uranium isotopically enriched in 235U. Preparation of a Pellet in Accordance with the Invention According to a first embodiment, a heterogeneous pellet according to the invention may be formed by pressing the powders. More particularly, a pellet according to the invention may be molded according to the structure represented in FIG. 1, by compression of a first powder comprising at least one fissile material and dedicated to forming the core of the pellet and of a second powder formed in all or part of a material according to the invention and dedicated to forming the annular external part. The pellet thus molded is subsequently sintered according to techniques known to a person skilled in the art, under reducing conditions, for example under an argon (or nitrogen) atmosphere to which hydrogen has been added. The sintering is more preferably carried out at a temperature ranging from 1200 to 2200° C., in particular from 1600 to 1800° C. This embodiment is particularly advantageous for forming pellets with an annular external layer with a thickness, along the axis of radius, of at least 50 μm. According to another specific embodiment, the annular external part (2) may be formed by deposition of a slip at the surface of a pellet dedicated to forming the core of the final pellet according to the invention. Thus, according to another of its aspects, the present invention relates to a process for the manufacture of a heterogeneous nuclear fuel pellet (10) according to the invention comprising at least the stages consisting in: (i) having available a powder comprising a material according to the invention or having available a powder formed of a mixture of uranium oxide such as UO2 or a higher oxide, such as U3O8, and gadolinium oxide Gd2O3 in proportions such that the gadolinium is present in a Gd/[Gd+U] atomic ratio ranging from 0.6 to 0.93; (ii) preparing a slip from the powder of stage (i); (iii) depositing the powder in the slip form on the surface of a pellet (1) comprising at least one fissile, indeed fertile, material; and (iv) sintering the pellet obtained on conclusion of stage (iii) under reducing atmosphere and at a temperature between 1200° C. and 2200° C. Thus, a pellet (10) according to the invention may be prepared from a powder of material according to the invention formed prior to the use thereof in stage (i) of the process of the invention. Alternatively, the material according to the invention may be produced directly on the pellet, during the overall sintering in stage (iv) of the pellet under reducing conditions. The slip in stage (ii) may be formed conventionally by mixing the powder of stage (i) with a liquid medium capable of being able to be easily removed by heating or natural evaporation, in particular ethanol. The pellet (1), at the surface of which the slip is deposited and which is dedicated to forming the core of the heterogeneous pellet (10) of the invention, is preferably a pressed pellet. It may be prepared by any conventional method known to a person skilled in the art for the preparation of nuclear fuel pellets. For example, the pellet (1) may be formed via the following stages: preparing the fuel powder, for example uranium oxide powder, referred to as pelleting stage; compacting the fuel powder in the pellet form by cold pressing or any other means, optionally while using a lubricant, such as zinc stearate, ammonium stearate or ethylene bis(stearamide) (composed of 76.8% C+13.3% H+5.1% O+4.8% N, sold by Hoechst under the name Ceridust). The pellet (1) may be nonsintered, sintered or presintered. The sintering may more particularly be carried out by heating the pressed pellet at high temperature, in particular at a temperature of greater than 1200° C., in particular under reducing atmosphere. “Presintered” pellet is understood to mean a pellet which has been subjected to a heat treatment below the sintering conditions, for example at a temperature ranging from 1000 to 1500° C., in particular of approximately 1200° C., especially under reducing atmosphere. The presintering advantageously makes it possible to improve the cohesion of the pellet, without achieving complete densification. The slip obtained on conclusion of stage (ii) may be deposited at the surface of the core pellet (1) by immersing the pellet in the slip. Stage (iii) may include the drying of the layer of slip deposited at the surface of the pellet, for example by leaving the pellet, at the surface of which the slip is deposited, in the open air for a period of time ranging from 5 to 30 minutes and optionally by heating between 40 and 90° C., typically between 50 and 60° C. The sintering in stage (iv) may be carried out under an argon atmosphere to which hydrogen has been added, for example under an argon atmosphere to which 5 mol % of hydrogen has been added. Preferably, this sintering stage is carried out at a temperature ranging from 1200° C. to 2200° C., in particular from 1600 to 1800° C. According to a specific embodiment, this sintering stage is carried out for a period of time of greater than or equal to 1 hour, in particular ranging from 3 to 8 hours. Nuclear Fuel Rod and Assembly The pellets according to the invention, as described above, may be employed in nuclear fuel rods. “Nuclear fuel rod” is intended to denote, conventionally, a tubular fuel element, having a small diameter, closed at its two ends, constituent of the core of a nuclear reactor and comprising a fissile or fertile material. A rod is more particularly formed of a plurality of fuel pellets stacked on one another and of a cladding surrounding the stack of pellets. For example, the cladding of a rod provided for a pressurized water reactor (PWR) may be made of zirconium alloy or of M5 (ZrNbO) alloy. Several rods form an assembly and several assemblies form the core of a nuclear reactor. The invention thus relates, according to another of its aspects, to a nuclear fuel rod comprising fuel pellets as defined above. A rod according to the invention may be composed exclusively of identical or different pellets in accordance with the invention. According to another specific embodiment, it may comprise, in addition to the pellets in accordance with the invention, other pellets not comprising the material of the invention as poison, for example homogeneous UO2 pellets. The invention also relates to a nuclear fuel assembly, for example used for a light water reactor, comprising fuel rods according to the invention as defined above. The number and the positioning of the rods according to the invention in a conventional assembly formed of a lattice of rods may be adjusted so as to result in optimum control of the reactor, as illustrated in example 4 which follows and in FIGS. 7.a and 8. Preferably, the rods enriched in gadolinium according to the invention are distributed uniformly in the fuel assembly. An appropriate distribution of the fuel assemblies comprising gadolinium according to the invention in the core of the nuclear reactor makes it possible to achieve a more uniform radial distribution of the power, this being the case throughout an operating cycle of the core before reloading. By way of example, a conventional assembly formed of a 17×17 lattice may incorporate from 4 to 64 rods according to the invention, it being possible for the other rods to be formed of conventional pellets not comprising poison, for example homogeneous UO2 pellets enriched in 235U. FIG. 3 diagrammatically represents, by way of example, a fuel assembly formed of a 17×17 lattice, composed of 265 fuel rods, including 25 rods (GD) formed of pellets according to the invention, the other rods (U) being formed of homogeneous UO2 pellets. The fuel rods are held by a structure comprising 24 guide tubes (TG). Of course, the material according to the invention may be employed as burnable neutron poison in nuclear fuel elements other than rods. The exposition which follows describes, by way of example, the use of the material of the invention in nuclear fuels of plate-type geometry. Nuclear Fuel of Plate-Type Geometry According to yet another alternative embodiment, the invention relates to a nuclear fuel element of plate-type geometry comprising one or more fissile, indeed even fertile, regions covered, at least in part, with a material according to the invention. These nuclear fuels of plate-type geometry are generally employed in low-power reactors. FIG. 9.a diagrammatically represents a comprehensive view of such a nuclear fuel assembly (100) comprising a stack of clad plates (FIG. 9.b) comprising a fissile material. The cladding is generally composed of an aluminum alloy. As represented in FIG. 9.c, the assembly may comprise one or more plates (103) comprising a fissile region (101), for example of UO2, covered at least in part with a layer (102) of material enriched in gadolinium according to the invention. The other plates (104) may be standard plates not comprising poison. According to a specific embodiment, as represented in FIG. 10.a, a plate (103) may more particularly be composed of a lattice of nuclear fuel pellets (111) for example of UO2 pellets, located in a lattice of cells (6), inserted between two plates (7,8). Such a fuel element structure of plate type is described in more detail in the document FR 2 889 765. In the context of this alternative embodiment, a few pellets, indeed even all of the pellets (111) incorporated in the fuel element of plate type may be covered, at least in part, with a layer (112) formed of a material according to the invention, as displayed, in cutaway view, in FIG. 10.b. The invention will now be described by means of the following examples and figures, given by way of illustration and without any limitation of the invention. Preparation of a Material According to the Invention Different mixed powders formed of a mixture of UO2 and Gd2O3 with a content by weight of Gd2O3 varying from 50% to 90% (for example: 55%, 65%, 69%, 80%, 82.4%) are compacted and then sintered under a reducing atmosphere of Ar, 5% H2, at 1700° C. for 4 hours, in order to give dense pellets. The results of the analyses, by X-ray diffraction, SEM and energy dispersive analysis (EDS), of the pellets thus obtained are presented in table 1 below. Crystalline phases with a crystallographic structure of cubic type are detected in the pellets thus obtained and more particularly: a crystallographic structure of cubic type with a unit cell parameter of approximately 5.43 Å, entitled cubic 2 (C2) phase, for a Gd/[Gd+U] atomic ratio between 0.5 and 0.71; a crystallographic structure of cubic type with a unit cell parameter of approximately 10.8 Å, entitled cubic 1 (C1) phase, for a Gd/[Gd+U] atomic ratio between 0.79 and 0.93; and a region of phase separation of these two phases for a Gd/[Gd+U] atomic ratio between 0.71 and 0.79. TABLE 1Powders12345Content by weight (%) of5565698082.4the starting powerGd2O3/(Gd2O3 + UO2)Z (Gd/(Gd + U)) atomic0.660.740.750.850.87ratio of the starting powderCrystallographic structurecubiccubiccubiccubiccubicC2C2 andC2 andC1C1C1 two-C1 two-phasephaseUnit cell parameter (Å) (1)~5.43mixedmixed~10.86~10.85Oxidation state of the+4/+5+4/+5+4/+5+4/+5+4/+5uraniumStatesolidsolidsolidsolidsolid(1) obtained by X-ray diffraction analysis Preparation of a Fuel Pellet According to the Invention by Pressing Powders (i) Powder of Material According to the Invention A powder of material according to the invention is prepared, as described in example 1, by sintering a mixture of UO2 and Gd2O3, in a Gd2O3/(UO2+Gd2O3) ratio by weight of 80%, at 1700° C. and under a reducing atmosphere of Ar, 5% H2, for 4 hours. (ii) Preparation of the Heterogeneous Pellet A pellet is molded according to the structure presented in FIG. 1 with a cylindrical internal part formed from a uranium oxide powder and an annular external part formed from the powder enriched in Gd obtained above. In order to distribute the powders according to FIG. 1, it is possible to use a thin partition made with two concentric rings. On completion of the filling, the thin partition is withdrawn and the pressing is carried out. The cylindrical core exhibits a radius (R1) of approximately 4 mm; the annular external part exhibits a thickness (e) of approximately 50 to 250 μm as a function of the supply of negative reactivity desired. The pellet is subsequently sintered under reducing conditions with an atmosphere of Ar, 5% H2, for 4 hours. Preparation of a Fuel Pellet According to the Invention by Deposition of a Layer Formed of a Powder Having a High Content of Gd (i) Powder of Material Two possibilities are selected: A—A powder of material according to the invention is prepared, as described in example 1, by sintering a mixture of UO2 (indeed even of U3O8) and Gd2O3 in a Gd2O3/(UO2+Gd2O3) ratio by weight of 80%. B—A powder is prepared by mixing UO2 (indeed even U3O8) and Gd2O3 in a Gd2O3/(UO2+Gd2O3) ratio by weight of 80%. (ii) Preparation of the Heterogeneous Pellet A pellet composed of fissile materials (1) is shaped by compaction with a cylindrical geometry. In order to give cohesion of the powder, presintering of this pellet may be carried out. An annular external part formed from the powder enriched in gadolinium, obtained in stage (i) according to mode A or mode B, is deposited, for example in the form of a slip (formed from the powder and ethanol), on the cylindrical surface and then the slip is dried. The pellet is subsequently sintered under reducing conditions with an atmosphere of Ar, 5% (molar) H2, for 4 hours. The cylindrical core, with fissile/fertile elements, exhibits a radius of approximately 4 mm; the annular external part, with the gadolinium, exhibits a thickness of approximately 30 to 250 μm as a function of the supply of negative reactivity desired. FIG. 2 shows the pellet observed in cross section by optical microscopy. Use of a Material According to the Invention as Burnable Poison in Nuclear Reactors for the Supply of Negative Reactivity and/or for the Reduction/Suppression of the Requirements for Boron and Other Neutron Poisons/Absorbing Materials The neutron performance of different 17×17 assemblies of nuclear fuel is modeled using the APOLLO2 computing code. i. Principles and Definitions of the Notions Used Kinfini: multiplication factor of the neutrons in an infinite medium (without taking into account the escapes); Ktrue: multiplication factor of the neutrons in a finite (true) medium. The difference between Kinfini and Ktrue is thus related to the amount of neutrons which escape from the reactor, without multiplying, in other words:Kinifini=Ktrue×factorgeometric (Eq. 1) the factorgeometric depending mainly on the geometry of the core but also on the nature of the materials. The reactivity ρ, expressed in pcm (percent mille), is another way of expressing the multiplication factor (infinite or true) mathematically, ρ = K - 1 K × 1 · 10 5 ( Eq . 2 ) Thus, equation 1 may be expressed as:ρinf=ρtrue+ρescapes (Eq. 3) Critical Reactor A “critical” reactor is a reactor in which the population of neutrons is constant and different from zero (without taking into account external sources), in other words a reactor for which Ktrue=1.00, or, expressed in terms of reactivity, by employing equation 2, ρ=0 pcm. The calculations carried out with the abovementioned computing code give us the Kinfini. It is found that, for the imaginary reactor, the term ρescapes is approximately −2500 pcm. Thus, a “critical” reactor within the meaning of the modelings carried out exhibits: ρtrue=0 pcm and ρinfini=2500 pcm; which is reflected, in multiplication factor, by: Ktrue=1.00 and Kinfini=1.025 (curve 7 in FIG. 4). By way of example, FIG. 4 represents the change in the multiplication factor (Kinfini) of an imaginary reactor for an assembly composed of homogeneous UO2 pellets enriched in 235U to 4.9%, with management by ¼ (4 operating cycles, in other words, in each cycle, a quarter of the assemblies [the ones most used] are exchanged for fresh assemblies), modeled using the computing code (curve 1). It is compared with the reactivity of the same reactor using boron (curve 5) as neutron poison (2000 ppm of boron diluted in the water of the heat-exchange fluid). In the case of the use of boron, the reactor may remain critical starting from the operating point (6) by decreasing the concentration of boron in the heat-exchange fluid (“critical boron” operating method). Multiplication Factor of an Ideal Reactor Curve 2 on the graph of FIG. 4 represents the change in the infinite multiplication factor of an “ideal” reactor. This “ideal” core is a reactor without a neutron penalty, with an initial superreactivity of 2000 pcm (thus with a Kinfini=1.050) and without penalty over the length of the cycle (operating point (4)). Thus, as represented in FIG. 4, an “ideal” reactor” within the meaning of the modelings carried out is a reactor which has a true reactivity, ρtrue of +2000 pcm up to the operating point (3). This superreactivity makes it possible to operate the reactor (for example in order to rise in power). Multiplication Factor of an Ideal Assembly FIG. 5 exhibits the change in the multiplication factor in an infinite medium Kinfini under hot conditions (that is to say, while considering the effect of the temperature) of an imaginary fuel assembly based on UO2 as a function of the mean burnup of the assembly, for 17×17 assemblies. The relationship between the change in the Kinfini of an assembly (Kassemblyinf) and the change in the Kinfini of a reactor (Kcoreinf), with management of N cycles, is given by the approximation below: K core inf ( x ) = 1 N ∑ i = 0 i = N - 1 [ K assembly inf ( x + L cycle × i ) ] ( Eq . 4 ) with:x the burnup of the assemblies in the 1st cycle,N the total number of cycles that the assemblies are used in the reactor,Kcoreinf(x) the Kinfini of a reactor, with management at N cycles, as a function of the burnup x,Kassemblyinf(x+Lcycle×i) the Kinfini of an assembly, as a function of a burnup,Lcycle the length of a cycle (in burnup units). In particular, Lcycle confirms that Kcoreinf(Lcycle)=1.025, in order for the reactor to be critical at the end of the cycle. By employing equation 4, it is possible to plot (FIG. 6) the change in the Kinfini of an “ideal” assembly and of a “critical” assembly, so that, employed in a reactor (with management by quarter, thus N=4), they respectively give an “ideal reactor” behavior and “critical reactor” behavior, as defined above. ii. Neutron Effect Obtained with Different Assemblies The change in the Kinfini for various 17×17 assemblies are presented in FIG. 7.a: imaginary fuel assembly based on UO2 (curve 1); assemblies employing 40 or 52 rods formed of heterogeneous pellets in accordance with the invention, the annular coating of which exhibits a thickness of 50, 60 or 150 μm and various Gd isotopic vectors. The other rods of the assembly are formed of homogeneous UO2 pellets enriched in 235U to 4.9% (curves 2 to 5); and “ideal” and “critical” assemblies as defined in the preceding point i. (curves 6 and 7). All these curves consider a boron-free imaginary reactor, that is to say a concentration of 0.0 ppm of boron in the heat-exchange fluid/moderator. For comparative purposes, the neutron effect obtained for assemblies incorporating rods of conventional homogeneous pellets composed of a (U,Gd)O2 solid solution comprising 8% by weight of Gd2O3 is presented in FIG. 7.b. It emerges from FIG. 7.a that it is possible, by adjusting the number of rods in accordance with the invention, the thickness of the burnable poison layer of the pellets constituting them and the isotopic vector of the gadolinium, to control the change in the reactivity of the reactor so as to approach an optimum change. Also, the comparison of FIGS. 7.a and 7.b shows that the change in the reactivity of the assembly may be better controlled with pellets in accordance with the invention than with conventional homogeneous pellets, since the curve of reactivity for an assembly according to the invention more closely approaches the “ideal” curve at the cycle end. iii. Reactivity of the Reactor The behavior of a reactor employing the assemblies of the invention is shown in FIG. 8. It emerges from FIG. 8 that the effect on the reactivity makes it possible to reduce, indeed even to suppress, the use of boron in the reactor. [1] Balestrieri thesis, 1995; [2] Tang and al., Order-to-disorder phase transformation in ion irradiated uranium-bearing delta-phase oxides RE6U1O12 (RE=Y, Gd, Ho, Yb, and Lu), Journal of Solid State Chemistry, 183(4), 844-848; [3] Tang and al., Microstructural evolution in irradiated uranium-bearing delta-phase oxides A6U1O12 (A=Y, Gd, Ho, Yb, and Lu), Journal of Nuclear Materials, 407(1), 44-47. |
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053533208 | claims | 1. In a water-cooled nuclear reactor comprising: a core of nuclear fuel; a reactor pressure vessel containing said nuclear fuel core therein and comprising an inlet; containment spaced outwardly from and surrounding said reactor pressure vessel; a pool of water located within said containment and outside said reactor pressure vessel at an elevation which is higher than an elevation of said nuclear fuel core; and means for channeling water from said pool to said inlet of said reactor pressure vessel by the force of gravity in response to a predetermined condition inside said reactor, said water channeling means comprising a supply pipe having a flow passage of first diameter extending from a first end to a second end thereof, said second end of said supply pipe flow passage being in flow communication with said pool via said first end of said supply pipe flow passage; and a nozzle having a flow passage extending from a first end to a second end, said first end of said nozzle flow passage being in flow communication with said first end of said supply pipe flow passage via said second end of said supply pipe flow passage and being in flow communication with said inlet of said reactor pressure vessel via said second end of said nozzle flow passage, the improvement wherein: a core of nuclear fuel; a reactor pressure vessel containing said nuclear fuel core therein and comprising an inlet; containment spaced outwardly from and surrounding said reactor pressure vessel; a pool of water located within said containment and outside said reactor pressure vessel at an elevation which is higher than an elevation of said nuclear fuel core; and means for channeling water from said pool to said inlet of said reactor pressure vessel by the force of gravity in response to a predetermined condition inside said reactor, said water channeling means comprising a supply pipe having a flow passage of first diameter extending from a first end to a second end thereof, said second end of said supply pipe flow passage being in flow communication with said pool via said first end of said supply pipe flow passage; and a nozzle having a flow passage extending from a first end to a second end, said first end of said nozzle flow passage being in flow communication with said first end of said supply pipe flow passage via said second end of said supply pipe flow passage and being in flow communication with said inlet of said reactor pressure vessel via said second end of said nozzle flow passage, the improvement wherein: 2. The nuclear reactor as defined in claim 1, wherein said throat has said second diameter for a length of at least 10 cm. 3. The nuclear reactor as defined in claim 1, wherein said second section has a maximum diameter which is less than a diameter of said inlet, and said nozzle flow passage further comprises a third section situated between said second section and said inlet, said third section having the shape of a bell mouth which changes gradually in diameter from said maximum diameter of said second section to said inlet diameter. 4. The nuclear reactor as defined in claim 3, wherein said nozzle flow passage further comprises a fourth section situated between said throat and said supply pipe, said fourth section having the shape of a circular cylinder of said first diameter. 5. The nuclear reactor as defined in claim 1, wherein said second section has a half-angle of about 4.degree.. 6. In a water-cooled nuclear reactor comprising: 7. The nuclear reactor as defined in claim 6, wherein said nozzle flow passage comprises a throat and first and second sections in flow communication with said throat, said throat having a second diameter which is smaller than said first diameter, said first section having the shape of a bell mouth which changes gradually in diameter from said first diameter to said second diameter, and said second section having the shape of a truncated cone having a minimum diameter substantially equal to said second diameter, said first section being situated between said throat and said supply pipe and said second section being situated between said throat and said inlet of said reactor pressure vessel. 8. The nuclear reactor as defined in claim 7, wherein said throat has said second diameter for a length of at least 10 cm. 9. The nuclear reactor as defined in claim 7, wherein said second section has a maximum diameter which is less than a diameter of said inlet, and said nozzle flow passage further comprises a third section situated between said second section and said inlet, said third section having the shape of a bell mouth which changes gradually in diameter from said maximum diameter of said second section to said inlet diameter. 10. The nuclear reactor as defined in claim 9, wherein said nozzle flow passage further comprises a fourth section situated between said throat and said supply pipe, said fourth section having the shape of a circular cylinder of said first diameter. 11. The nuclear reactor as defined in claim 7, wherein said second section has a half-angle of about 4.degree.. 12. A reactor pressure vessel comprising an inlet in flow communication with a nozzle, said nozzle having a flow passage extending from a first end to a second end, said first end of said nozzle flow passage being in flow communication with said inlet of said reactor pressure vessel via said second end of said nozzle flow passage, wherein said nozzle flow passage comprises a throat and first and second sections in flow communication with said throat, said throat having a predetermined diameter, said first section having the shape of a bell mouth which gradually decreases in diameter from a first diameter to said predetermined diameter, and said second section having the shape of a truncated cone having a minimum diameter substantially equal to said predetermined diameter, said first section being situated between said throat and said first end of said nozzle flow passage and said second section being situated between said throat and said inlet. 13. The reactor pressure vessel as defined in claim 12, wherein said throat has said predetermined diameter for a length of at least 10 cm. 14. The reactor pressure vessel as defined in claim 12, wherein said second section has a maximum diameter which is less than a diameter of said inlet, and said nozzle flow passage further comprises a third section situated between said second section and said inlet, said third section having the shape of a bell mouth which changes gradually in diameter from said maximum diameter of said second section to said inlet diameter. 15. The reactor pressure vessel as defined in claim 14, wherein said nozzle flow passage further comprises a fourth section situated between said throat and said first end of said nozzle flow passage, said fourth section having the shape of a circular cylinder of said first diameter. 16. The reactor pressure vessel as defined in claim 12, wherein said second section has a half-angle of about 4.degree.. |
063209241 | description | BEST MODE FOR CARRYING OUT THE INVENTION Referring now to FIG. 1, there is illustrated a nuclear fuel bundle assembly, generally designated 10, including a plurality of fuel rods 12 supported between upper tie plate 14 and a lower tie plate 16. Fuel rods 12 pass through a plurality of fuel rod spacers 18 at vertically spaced positions along the fuel bundle. The spacers 18 provide intermediate support to retain the elongated fuel rods 12 in spaced relation relative to one another and to restrain the fuel rods from lateral vibration. With reference to FIG. 1, a 10.times.10 array of fuel rods is illustrated, while FIG. 6 illustrates an 11.times.11 array. It will be appreciated, however, that the invention hereof is applicable to various arrays of fuel rods of different numbers, for example, 8.times.8, 9.times.9, etc. Each fuel rod 18 is formed of an elongated tubular cladding material, with the nuclear fuel and other materials sealed in the tube by end plugs. The lower end plugs register in bores formed in the lower tie plate 16, while the upper end plugs are disposed in cavities in the upper tie plate 14. Additionally, the fuel rod assembly includes a channel 20 of substantially square cross section, sized to form a sliding fit over the upper and lower tie plates and the spacers, so that the nuclear fuel bundle, including the channel 20, tie plates 14, 16, rods 12 and spacers 18 can be removed from the reactor core (not shown). Turning now briefly to FIGS. 6 and 7, there is illustrated a spacer 18 constructed in accordance with the present invention, and having a plurality of individual ferrules 22 and springs 24, each ferrule 22 having a single associated spring 24, which bears against a single fuel rod 12' in a respective ferrule. The ferrules 22 are arranged in a square matrix in which each ferrule receives a fuel rod 12' and maintains the fuel rods spaced and restrained relative to adjoining fuel rods. The spring 24 of each ferrule biases its associated fuel rod in a lateral direction against hard stops 26 opposite the spring, whereby each fuel rod 12' is maintained in a predetermined position relative to one another and within the marginal band 28 of the spacer 18. The marginal band 28 normally includes inwardly directed flow tabs 30, but the latter form no part of this invention. With reference back to FIG. 2, each spacer ferrule 22 has a generally hollow, generally cylindrical configuration. The wall of each cylindrical ferrule is indented at circumferentially spaced locations along one side of the ferrule to form the inwardly directed, axially extending hard stops 26. It will be appreciated that the stops 26 extend the full height of the ferrule, although the stops could be provided at axially spaced locations along the height of the ferrule. As best illustrated in FIG. 2, each ferrule 22 includes a central spring opening 30 opposite the hard stops 26. Spring opening 30 is defined partially by upper and lower circumferential bands 32, 34. Between these two bands, the opening 30 is centrally located and has a substantially "I" shape which includes relatively wider upper and lower portions 36, 38 connected by a narrower axially extending stem portion 40. The opening 30 is more specifically formed by upper and lower horizontal edges 42, 44, upper vertical edges 46, 48 which are connected to the vertical stem edges 50 and 52 by means of horizontal steps or shoulders 54, 56. Similarly, lower vertical edges 58 and 60 are connected to the vertical stem edges 50 and 52 by step or shoulder surfaces 62, 64. As already noted above, the opening 30 is directly opposite and centered relative to the hard stops 26. FIG. 3 illustrates the spring 24 in detail. The spring also has a substantially "I" shape which includes, generally, a planar body including upper and lower horizontal flanges 64, 66, respectively, interconnected by a vertical stem 68. The spring 24 also includes a pair of T-shaped cutouts, one of which is upright and located in the upper half of the spring, while the other is inverted and located in the lower half of the spring. Thus, the upper cutout 70 includes a horizontal portion 74 and a vertical portion 76 whereas the lower cutout 72 includes a horizontal portion 78 and a vertical portion 80. These openings increase the flexibility and resilience of the spring 24 as will be appreciated by those of ordinary skill in the art. The spring 24 is substantially planar except as noted below. In the upper half of the spring, the horizontal band 82 lying adjacent the horizontal portion 74 of the opening 70 is formed to include a radially outward tab or projection 84 in substantially axial or vertical alignment with the vertical portion 76 of the T-shaped cutout 70. A similar radially outwardly extending projection 86 is formed in the lower band 88 lying adjacent the horizontal portion 78 of the T-shaped cutout 72. Both projections 84 and 86 extend outwardly to substantially the same degree. In the center of the stem portion of the spring, between the vertical portions 76, 80 of the T-shaped cutouts 70, 72, respectively, there is formed a radially inwardly directed projection 89 which is further formed to include a radially inwardly extending dimple 90 which is designed to engage a fuel rod in the associated ferrule 22. With further reference now also to FIGS. 4 and 5, the spring 24 may be aligned and seated within the opening 30 such that the projection 88 and dimple 90 extend radially inwardly into the interior space of the ferrule 22, but the upper and lower flanges 64 and 66 remain on the outside of the ferrule 22. The size of the opening 30 is designed so that the spring 24 is relatively snugly received in the opening, with upper and lower edges 92, 94, respectively, supported by upper and lower edges 42, 44. respectively, of the opening 30. At the same time, vertical edges 96, 98 of the projection 88 are supported by the edges 50, 52 of the opening 30. In this way, both vertical and lateral movement of the spring 24 relative to the ferrule opening 30 is substantially precluded. It will also be noted that the movement of the spring 24 toward or away from the fuel rod 100 is limited by the edges 46, 48, 58 and 60 of the opening 30. Limiting the spring's freedom of movement is achievable because both the ferrule opening shape and the spring shape are formed by stamping, which is a reliable manufacturing process for dimensional stability. Typically, when the spacer ferrules 22 are placed within the spacer band prior to welding, the ferrules 22 are aligned as shown in FIG. 6 with all of the hard stops 26 and spring openings 30 similarly aligned. During placement of the ferrules, the springs 24 are located within the respective openings 30 and the ferrules 22 are then welded together at diametrically opposed locations between the hard stops 26 and the opening 30. These welds are shown in FIG. 7 by letter designations W.sub.1 and W.sub.2, with the understanding that all of the ferrules are welded in a similar manner. It will be appreciated that when a fuel rod 12' is inserted within the ferrule 22, the rod 12' contacts the hard stops 26 and the dimple 90. In this regard, FIG. 7 illustrates the fuel rod 12' in phantom while the spring 24 is shown in its relaxed position. In this way, it is to be understood that when the fuel rod is inserted within the ferrule, the dimple 19 and indeed the entire vertical stem of the "I" spring will be flexed radially outwardly to accommodate the rod, with the resulting bias of the fuel rod 12' against the hard stops 26. With all of the springs and ferrules similarly aligned as shown in FIGS. 6 and 7, the bundles can be oriented horizontally for shipment in such a way that any loads which occur during bundle shipment to customers are taken by the hard stops 26 on the ferrules and not by the springs. In addition, the unique "I" shape of the springs 24 in accordance with this invention, results in only a minimal protrusion of material into the subchannel flow. This, in combination with the tight capture of the spring within the ferrule cutout, securely holds the spring against the water induced movement that could otherwise cause wear. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiment, it is to be understood that the invention is not to be limited to the disclosed embodiment, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims. |
claims | 1. A specimen holder support device for use with an electron microscope having a wall defining a vacuum tight specimen chamber, said specimen holder support device comprising: a cylindrical support member extending through said wall in an x direction and mounted to said wall; a spherical bearing surface formed on an inner surface of said support member; a generally cylindrical swinging member having a cylindrical bearing on an outer surface inserted in said support member and swingable about said spherical bearing in Y and Z directions perpendicular to each other and the X direction; a specimen holder slidably mounted in said swinging member via an O-ring and having a front end extending inwardly beyond said spherical bearing surfaces; an X-motion drive mechanism engaging with the front end of said specimen holder; and a specimen holder movement-arresting means spring biasing said specimen holder relative to said swinging member for alleviating force applied to said X-motion drive mechanism from said specimen holder when a vacuum is pulled on said chamber. 2. The specimen holder support device of claim 1 , wherein said holder movement-arresting means is mounted to an outer surface of said swinging member and connected to said specimen holder via a hole formed in said outer surface, and wherein said specimen holder is biased away from said specimen chamber by said holder movement-arresting means. claim 1 3. The specimen holder support device of claim 2 , wherein a means for detecting axial position of said specimen holder is mounted to said holder movement-arresting means. claim 2 4. A specimen holder support device for use with an electron microscope having a wall defining a specimen chamber whose interior is maintained as a vacuum, said specimen holder support device comprising: a holder mounting member having a spherical bearing supported to said wall defining said specimen chamber, a swinging member having an inner end rotatably held by said spherical bearing, a holder-receiving hole formed in said swinging member and acting to place the interior of said specimen chamber in communication with outside, and a slider-receiving groove formed in an outer-end portion of said swinging member axially of said swinging member; a specimen holder held in said holder-receiving hole slidably and hermetically and having an inner-end portion and a guide pin, said inner-end portion having a specimen holding portion for holding a specimen, said inner-end portion being positioned in said specimen chamber, said guide pin being mounted on an outer surface of an outer-end portion located outside said wall defining said specimen chamber, said guide pin protruding into said slider-receiving groove; a holder inner end-positioning drive mechanism abutting against the specimen holding portion of said specimen holder to place said specimen holding portion in position axially of said swinging member; a slider received in said slider-receiving groove so as to be movable axially of said swinging member and having a pin engagement portion and a movement-arresting portion shaped to protrude outwardly from said swinging member, said pin engagement portion engaging with the guide pin of said specimen holder; a slider-arresting member having a slider-arresting portion for coming into engagement with said movement-arresting portion of said slider from a side of an inner end of said holder mounting member; and a spring for biasing said slider-arresting member to arrest movement of said slider toward said inner end. 5. The specimen holder support device of claim 4 , wherein said slider-arresting member consists of a lever having first and second ends and held so as to be rotatable about the axis of rotation of the arresting member adjacent to the outer surface of said swinging member, said slider-arresting portion being formed at said first end, said spring being connected to said second end. claim 4 6. The specimen holder support device of claim 5 , wherein there is further provided a linear gauge for measuring the amount of rotation of said lever. claim 5 7. A specimen holder support device for use with an electron microscope having a wall defining a vacuum tight specimen chamber, said specimen holder support device comprising: a cylindrical support member extending through said wall in an x direction and mounted to said wall; a spherical bearing surface formed on an inner surface of said support member; a generally cylindrical swinging member having a cylindrical bearing on an outer surface inserted in said support member and swingable about said spherical bearing in Y and Z directions perpendicular to each other and the X direction; a specimen holder slidably mounted in said swinging member via an O-ring and having a front end extending inwardly beyond said spherical bearing surfaces; an X-motion drive mechanism for moving said specimen holder; and a specimen holder movement-arresting means spring biasing said specimen holder relative to said swinging member for alleviating force applied to said X-motion drive mechanism from said specimen holder when a vacuum is pulled on said chamber. |
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summary | ||
description | This application is a continuation of Ser. No. 13/706,374 filed Dec. 6, 2012 the entire disclosure of which is hereby incorporated by reference. The invention relates to an X-ray apparatus, comprising an electron beam source, emitting an electron beam, a target, onto which the electron beam is directed, thus forming a focal spot on the target, X-ray optics, collecting X-rays emitted from the focal spot and forming an X-ray beam, and a sample position at which the X-ray beam is directed. Such an X-ray apparatus is known from U.S. Pat. No. 7,929,667 B1. By means of X-rays, samples may be investigated in a destruction-free and efficient manner. X-rays may interact with a sample in numerous ways in order to obtain analytical information about the sample, with X-ray diffraction (XRD) and X-ray fluorescence (XRF) being two important methods. In general, high X-ray intensities are useful to obtain high signal to noise ratios in X-ray analysis experiments. X-rays are typically generated by directing an electron beam onto a target. The deceleration of the beam electrons (resulting in Bremsstrahlung) as well as the refilling of depleted deep electron shells of the target material (resulting in characteristic X-rays) leads to X-ray emissions within the focal spot of the electron beam on the target. In order to provide X-rays of a particular wavelength, monochromators may be used. Further, if the sample is significantly spaced apart from the focal spot, it is useful to focus X-rays by suitable optics such as Göbel mirrors or Montel optics. U.S. Pat. No. 6,249,566 B1 proposes to combine a microfocus X-ray source with Montel type optics to focus X-rays onto a sample. Apparent focal spot sizes of about 30 μm or less are proposed. A particular high brightness X-ray source has been proposed in U.S. Pat. No. 7,929,667 B1, wherein an electron beam is focused on a jet of liquid metal, such as gallium. Higher power loads on the target due to the electron beam and thus high brightness levels are possible as the target is already liquid and can dissipate quickly the generated heat from the focal spot. A multilayer X-ray focusing element may be used to shape an X-ray beam. Focal spot sizes of about 10-15 μm are mentioned. U.S. Pat. No. 6,711,233 B2 also proposes an X-ray source wherein an electron beam is directed onto a liquid metal jet target. It is proposed to match the size of the electron beam with the size of the jet, with the jet having a diameter of about 1-100 μM. When combining a microfocus X-ray source with X-ray optics, it is necessary to align these components with respect to each other. Alignment in this sense means that a certain aspect of the beam properties downstream the mirror is maximized. Depending on the aimed application this aspects can for example be flux density or integral flux. In classical x-ray systems this is achieved by changing the x-ray optics and mechanically repositioning the x-ray optics. However, mechanically moving the X-ray optics on the μm range to match the focal spot of the X-ray source with the focus of the X-ray optics is difficult in practice, in particular due to backlash of alignment mechanics. It is the object of the invention to provide an X-ray apparatus wherein aligning the X-ray optics with respect to a microfocus X-ray source is simplified. This object is achieved, in accordance with the invention, by an X-ray apparatus as mentioned in the beginning, characterized in that the X-ray apparatus further comprises an electrostatic or electromagnetic electron beam deflection device, suitable for moving the focal spot on the target, and in that in any direction the extension of the focal spot is smaller at least by a factor F, with F=1.5, than the extension of the target. By means of the electron beam deflection device, the focal spot can be moved on the target. The X-ray apparatus is aligned when the focal spot of the electron beam on the target overlaps the focus of the x-ray optics. Instead of mechanically moving the X-ray optics, the inventive apparatus allows moving the focal spot, what can conveniently be done with electric means (such as changing the voltage between electrodes or changing an electric current through an electromagnetic coil), in particular without using alignment mechanics. An electric alignment is highly reproducible, allows a higher precision, and is in particular not subject to backlash effects. Accordingly, the inventive apparatus can be aligned in a fast and simple way. In accordance with the invention, the focal spot has a size (extension) S with S*F≤T, with F=1.5 and T being the size (extension) of the target; this equation is valid in any direction (i.e. S and T are measured in the same direction, but this direction can be chosen arbitrarily; further below SX, SY, SZ and TX, TY, TZ as the sizes of the focal spot and the target in directions x, y, z are discussed in more detail). This means that the focal spot has a minimum available alignment range in any direction without leaving the target. This requirement ensures that, after a coarse prealignment of the target and the X-ray optics by mechanical means, a fine alignment via the electron beam deflection device is easily feasible. The extension of the focal spot (and the electron beam) may be determined as the full width at half maximum (FWHM) of the electron intensity distribution. The extension of the X-ray beam may be determined as the full width at half maximum of the photon intensity distribution. It should be noted that the electron beam deflection device may be included in the electron beam source (then the electron beam source needs control inputs for adjusting the beam deflection), but typically is separate from the electron beam source. The electron beam has a (maximum) diameter small enough to qualify for a microfocus X-ray source, such as 100 μm or less, preferably 30 μm or less. The X-ray apparatus is typically an X-ray analysis apparatus, with a sample to be analyzed (typically a single crystal sample, thin film sample, or a powder sample) being located at the sample position. A typical X-ray measurement to be done with an inventive X-ray apparatus is X-ray diffraction (XRD), in particular single-crystal X-ray diffraction, high resolution thin film analysis, grazing incidence diffraction, microdiffraction as well as (grazing incidence) small-angle scattering. In a preferred embodiment of the inventive apparatus, the target is a liquid metal jet target. This allows a particularly high brilliance. The target material at the focal spot is continuously replaced, what avoids a local overheating (e.g. evaporation) of the target. Further, the jet is a simple way to provide a curved target surface (see below), typically with a circular curvature. Preferred is a further development of this embodiment, wherein in a direction transverse to the liquid metal jet target propagation direction and transverse to the propagation direction of the electron beam, the extension of the focal spot is smaller at least by a factor FT, with FT=2, preferably FT=5, than the extension of the liquid metal jet target. This increases the available alignment range of the focal spot on the target. Further, a curvature of the target has a stronger effect on the apparent spot size and the self-absorption of the target. In a highly particularly preferred embodiment, the target has a curved surface, in particular having a radius of curvature R, with 0<R≤10 mm, preferably with 0<R≤1 mm. The curved surface allows an adjustment of the apparent spot size of the focal spot by moving the focal spot on the target. When the electron beam hits the target surface perpendicularly or almost perpendicularly, an X-ray beam emitted at about 90° with respect to the electron beam will appear to have a small focal spot (small apparent spot size). On the other hand, when the electron beam hits the target surface at a flat or even at an almost tangential angle, an X-ray beam emitted at 90° with respect to the electron beam will appear to have a large focal spot (large apparent spot size); however the latter X-ray beam will experience less self-absorption. Note that the target may have parts with a non-curved surface, too, in accordance with this embodiment. At least a part of the curved surface of the target faces the electron beam source, such that the focal spot may be moved across said part. Note that the radius of curvature may change in said part. In a further development of this embodiment, the electron beam deflection device is suitable for moving the focal spot on the target in a plane in which the target surface is curved. Then the adjustment of the focal spot size via the target can be done particularly simple in an electrical way. Particularly preferred is an embodiment, wherein the X-ray apparatus further comprises an electrostatic or electromagnetic electron beam focusing device, suitable for changing the spot area of the focal spot at least by a factor FS, with FS=2, preferably FS=5. The electron beam focusing device allows a widening and narrowing of the focal spot on the target. By this means, further characteristics of the resulting X-ray beam may be adjusted, such as the size, shape, divergence or (integral) intensity, without changing the e-beam power. It should be noted that the electron beam focussing device may be included in the electron beam source (then the electron beam source needs control inputs for adjusting the beam focusing), but typically is separate from the electron beam source. The electron beam focussing device may be integrated with the electron beam deflection device. In a preferred further development of this embodiment, the electron beam focusing device comprises one or more electromagnetic coils and/or one or more charged electrodes. These simple elements have shown good results in practice. The electron movement may be influenced via magnetic fields generated by the coils or electric fields at the electrodes. Further preferred is an embodiment wherein the electron beam deflection device is suitable for moving the focal spot on the target by at least a distance D, with D=50 μm, preferably D=200 μm. These ranges are typically well suited for both a simple relative alignment of the target and the electron beam, and for adjusting the apparent spot size at a curved target surface. In an advantageous embodiment, the electron beam deflection device is suitable for deflecting the electron beam in two independent directions perpendicular to a propagation direction of the electron beam, in particular wherein said two independent directions are perpendicular to each other. Two linear independent movement directions give access to an area of alignment on the target. Perpendicular orientation of the independent directions simplifies accessing a particular spot on the target. Further preferred is an embodiment wherein the electron beam deflection device comprises one or more electromagnetic coils and/or a one or more charged electrodes. These simple elements have shown good results in practice. The electron movement may be influenced via magnetic fields generated by the coils or electric fields at the electrodes. Particularly preferred is an embodiment wherein the X-ray optics comprises a multilayer mirror, in particular a Montel mirror or a Göbel mirror or a mirror having a single reflective surface curved with respect to both a sagittal and a meridional direction of incident X-rays, and/or capillary X-ray optics. These elements allow an accurate focusing or collimation of X-rays, proven in practice. In particular, the X-ray optics may comprise a double curved mirror as described in U.S. Pat. No. 7,248,670 B2. In a preferred embodiment, the factor F=2, preferably F=5. This increases the available alignment range of the focal spot on the target. Further, a curvature of the target has a stronger effect on the apparent spot size and the self-absorption of the target. Preferred is further an embodiment wherein the X-ray optics is positioned to collect X-rays emitted from the focal spot at essentially 90° with respect to a propagation direction of the electron beam hitting the target. In this orientation, high X-ray intensity levels can be obtained, and spot size adjustment via a curved target surface works well. The X-ray optics are typically arranged at an angle of (and including) 85° through 95° with respect to the electron beam, and use X-rays from an angle interval of 10° or less, typically 5° or less. Further within the scope of the present invention is a method for aligning an X-ray apparatus, in particular an inventive X-ray apparatus as described above, wherein the apparatus comprises an electron beam source, emitting an electron beam, a target, onto which the electron beam is directed, thus forming a focal spot on the target, X-ray optics, collecting X-rays from a focus of the X-ray optics, characterized in that the focal spot is moved on the target by deflecting the electron beam by means of an electric and/or magnetic field until the focal spot overlaps with the focus of the X-ray optics. After a coarse mechanical prealignment, this fine alignment is simple to perform by electric means, is highly reproducible and precise and not subject to mechanical backlashes. Typically the fine alignment includes iteratively or continuously changing the focal spot position while simultaneously monitoring the photon flux at a detector arranged downstream (after) the X-ray optics. Also within the scope of the present invention is a method for aligning an X-ray apparatus, in particular an inventive X-ray apparatus as described above, wherein the apparatus comprises an electron beam source, emitting an electron beam, a target, onto which the electron beam is directed, thus forming a focal spot on the target, X-ray optics, collecting X-rays from a focus of the X-ray optics, characterized in that the focal spot is moved on the target by deflecting the electron beam by means of an electric and/or magnetic field, and/or the spot area of the focal spot is changed by changing the focusing of the electron beam by means of an electric and/or magnetic field, until the photon flux or the photon flux density of an X-ray beam formed by the X-ray optics is maximized. Again, this alignment is simple to perform by electric means and is highly reproducible Typically the alignment includes iteratively or continuously changing the focal spot position. The photon flux density may, for example, be measured at a sample position or a detector position downstream the X-ray optics. If the X-ray optics is of focusing type, the photon flux density is the optimization parameter and is typically measured at the image focus (second focus) of the X-ray optics. If the X-ray optics is of collimating type, the photon flux per solid angle is the optimization parameter wherein the divergence and flux can be measured anywhere downstream the X-ray optics. It should be noted that within the inventive methods mentioned above, the extension of the focal spot is typically always smaller at least by a factor F, with F=1.5, preferably F=2, most preferably F=5, than the extension of the target in any direction. In a preferred variant of this latter inventive method, the apparatus is switched between two operation modes wherein in a first of the operation modes the photon flux is maximized, and wherein in a second of the operation modes the photon flux density is maximized. By changing the operation modes, the apparatus may be adapted to dedicated analysis measurements without the need to change the X-ray optics. Thus, this inventive method is very time-saving and cost-saving. With the photon flux density being maximized, diffraction data from a limited local area may be well obtained. With the photon flux maximized, diffraction data may be obtained with a high signal to noise ratio in short time. The change of operation modes may in particular be done by moving the focal spot on a curved target surface to a different position. In an advantageous further development, the target of the apparatus is chosen as a target with a curved surface having a radius of curvature R, with 0<R≤1 mm. This simplifies changing the operation modes by moving the focal spot on the target. Further advantages can be extracted from the description and the enclosed drawing. The features mentioned above and below can be used in accordance with the invention either individually or collectively in any combination. The embodiments mentioned are not to be understood as exhaustive enumeration but rather have exemplary character for the description of the invention. The invention is shown in the drawing. Overview of the Invention The invention proposes an X-ray apparatus with an X-ray source, in particular a microfocus X-ray source, which allows for a continuous variation of the position of the electron beam on the target, in particular a liquid metal jet target, preferably in two directions. In other words, the position of the focal spot of the electron beam is variable. To alter the spot position, the electron beam can be deflected by applying an electric and/or magnetic field to the electron beam. As an advantage of the variable spot position, it is possible to align the X-ray source and a subsequent X-ray optics in a fast and comfortable way. In the state of the art, the alignment is done only mechanically. Due to the backlash of the mechanics in and/or at the optics housing it is difficult and time consuming to optimize the alignment (which is done by increasing the photon flux of the primary beam). However, by varying the spot position on the target, the relative position of the X-ray optics and the focal spot can be changed and thus optimized, in particular such that the photon flux or the photon flux density is maximized. As the spot position is not varied mechanically, but electro-magnetically via electrodes or coils (e.g. in the source), this alignment procedure is very reproducible with an accuracy in the μm-range. Preferably, the target has a curved surface, for example wherein the target is of liquid metal jet type, what is preferred for the invention. By moving the electron beam perpendicular to the flow direction of the jet, the projected size of the X-ray emission area can be changed continuously. A combination of said microfocus X-ray source with curved (in particular elliptical or parabolic shape) multilayer mirrors allows to tailor the size, shape, divergence and intensity of the X-ray beam at the sample position. These properties of the X-ray beam may be changed continuously, allowing to adapt the X-ray beam to the needs of the experiment without the need of swapping optics. The optimization of the X-ray beam properties further results in an improved data quality and a shortened measurement time. When the electron beam is positioned close to the center position of the jet, the take-off angle of the X-ray beam is small, and X-ray self-absorption in the target is high, resulting in a small apparent source size with reduced integral flux, but increased flux density (“flux density maximization”). This small FWHM size of the X-ray source is the optimum X-ray beam condition for analyzing small samples; using focusing optics most of the photons are in the center of the X-ray beam hitting the small sample. By this, a diffracted intensity from the sample is maximized and the background noise is reduced, as the amount of photons that do not hit the sample, but just contribute to the background noise, is low. When shifting the electron beam away from the center towards the edge of the liquid metal jet target, the take-off angle is increased, enlarging the apparent spot size and reducing the self-absorption in the metal jet target. Consequently, using focusing optics the FWHM of the X-ray beam is increased and the peak intensity (flux density) is decreased (“Flux maximization”). Compared to the flux density maximization, the integral flux is now increased, as the self-absorption of the generated X-ray photons in the jet is reduced by placing the electron beam closer to the edge of the jet. This is the optimum condition for analyzing larger samples. It should be noted that by changing the position of a typical focal spot on a typical jet, the integral intensity can be changed by about 20%, and the flux density can be changed by about 50% with ease, compare FIG. 6. For this diagram, at different sample sizes, the flux and the flux density were maximized each, and the ratio of the mean fluxes incident on the respective sample diameter was determined at these alignment positions. According to the results, depending on the sample diameter used in the respective experiment, either a flux density or a flux optimized alignment is preferable. Preferably, the inventive X-ray apparatus is further capable of changing the size of the focal spot of the electron beam on the target by changing the focusing of the electron beam (“variable spot size”). In other words, the electron beam is widened or narrowed by electromagnetic means. This way the (microfocus) X-ray source is capable of changing the e-beam spot size on the metal jet target. It was found that the electron power density can be increased when the e-beam spot size decreases, without overheating the target. This can be used to increase the photon flux density, at the expense of integral photon flux. Small e-beam spots will result in small apparent X-ray spot sizes, advantageous for smaller samples, while larger e-beam spots will allow larger X-ray spot sizes at higher X-ray flux, advantageous for larger samples. Together with X-ray optics, this enables the system to control the size of the X-ray spot size on the sample position, the divergence of the X-ray beam and the integral flux downstream the X-ray optics. Description of Inventive Experimental Setups Shown in the Figures FIG. 1a shows schematically an embodiment of an inventive X-ray apparatus 1. An electron beam source 2 emits an electron beam 3. The electron beam 3 hits a target 4, here of a solid and flat type. A typical solid target material for use with the invention is copper. The area where the electron beam 3 hits the target 4 is called a focal spot 5. At the focal spot 5, X-rays are generated. X-ray optics 6, here of Montel type with two graded multilayer mirrors in a side by side orthogonal configuration, within an optic housing 6a, collect X-rays from a focus 7 of the X-ray optics 6 (compare focal length f1 on the entry side) and its close vicinity, thus forming an X-ray beam 8 directed to a sample position 9, where a sample to be investigated (not shown) is located. Note that the X-rays are collected at an angle δ of about 90° with respect to the electron beam propagation direction (here negative z). Beyond the sample position 9, an X-ray detector (not shown) is located. In the example shown, the X-ray beam 8 is focused at the sample position 9 (compare focal length f2 on the exit side); however it is also possible to parallelize (or otherwise shape) the X-ray beam 8 by means of the X-ray optics 6, in accordance with the invention. In the configuration shown, with the electron beam 3 being undeflected (i.e. propagating linearly), the focus 7 of the X-ray optics 6 deviates slightly from the focal spot 5 of the electron beam 3. Accordingly, only a small percentage of the X-rays generated at the target 4 or its focal spot 5, respectively, is collected by the X-ray optics 6. In order to increase the percentage of collected X-rays, the electron beam 3 may be deflected by means of an electron beam deflection device 10, here comprising a pair of charged electrodes (alternatively or in addition, the electron beam can be deflected by a magnetic field, generated by an electromagnetic coil). The deflection device 10 can deflect (shift) the electron beam 3 continuously in two orthogonal directions x, y perpendicular to its propagation direction z by adjusting a control voltage at the electrodes (or alternatively or in addition, adjusting a current at electromagnetic coils). In the embodiment shown, the defection device 10 is separate from the electron beam source 2; however, the deflection device 10 may also be integrated into the electron beam source 2. In FIG. 1b, the electron beam deflection device 10 has been activated in order to move the focal spot 5 on the target 4. After proper adjustment of the deflection device 10, namely slightly moving the focal spot 5 over a distance D basically in negative y direction or deflecting the electron beam 3 by a small angle α to the right, respectively, the focal spot 5 overlaps with the focus 7 of the X-ray optics 6. Thus a high percentage of the X-rays generated at the focal spot 5 may be collected by the X-ray optics 6 and directed to the sample position 9. Note that the optimum position of the focal spot 5 is typically found by maximizing the photon flux or the photon flux density downstream the X-ray optics 6, such as at the sample position. It should be noted that the width of the electron beam as well as the width of the X-ray beam is shown enlarged in the figures, in order to increase comprehensibility. A typical distance D over which the focal spot 5 can be moved on the target 4 is about 200 μm. FIG. 1c illustrates a variant of the embodiment of the X-ray apparatus 1 of FIG. 1a, but comprising an electron beam focusing device 11 (here comprising an electromagnetic coil assembly) in addition to the deflection device 10. Note that the focusing device 11 may be integrated into the deflection device 10 and/or into the electron beam source 2, if desired. The electron beam focusing device 11 allows to change the focusing of the electron beam 3, i.e. the width of the electron beam 3 on the target 4, by changing the electric currents through the coils of the coil assembly. By this means, the area of the focal spot can be adjusted directly. In the figure, the solid lines of the strongly narrowing electron beam 3 belong to a focal spot 5a with a small focal spot area Aa, whereas the dashed lines of the electron beam 3 only slightly narrowing belong to a focal spot 5b with a rather large focal spot area Ab; note that the areas Aa, Ab are shown in a projection each. Typically, the focusing device 11 allows an area change by a factor of up to five. By altering the focusing of the electron beam 3, some properties of the X-ray beam 8 at the sample position 9 can be altered, such as the beam divergence or the integral photon flux, without changing the electron beam power. FIGS. 2a and 2b illustrate focal spots 5a, 5b on a curved target 4 for an electron beam 3 for different positions of the focal spots 5a, 5b on the target 4. The figures show cross-sections through the target 4, here a circular liquid metal jet propagating in the x direction with a radius of curvature R, in a plane (yz-plane) including the electron beam propagation direction (negative z) and perpendicular to the jet propagation direction x. In this plane, the target surface 12 is curved. The electron beam 3 can be moved at least within this plane, i.e. here basically in y direction. If the electron beam 3 hits the target 4 basically perpendicular to the curved target surface 12 (compare angle β of about 80°), as shown in FIG. 2a, the apparent focal spot size SZ in z direction is rather small, in particular smaller than the focal spot size SY in y direction. Since the X-rays originate from a small area, a high photon flux density can be achieved. On the other hand, if the electron beam 3 hits the curved target surface 12 under a relatively flat angle (compare angle γ of about 35°), as shown in FIG. 2b, the apparent focal spot size SZ in z direction is rather large, in particular larger than the focal spot size SY in y direction. In the configuration of FIG. 2a, with a small focal spot size SZ in z direction, self-absorption of X-rays is rather large when the X-ray beam is taken on the left side in y direction (i.e. perpendicular to the electron beam propagation direction): X-rays generated on the right hand side of the focal spot 5a have to pass much target material before leaving the target 4. In contrast, in the configuration of FIG. 2b, with a large focal spot size SZ in z direction, self-absorption is relatively weak: Even X-rays generated on the right hand side of the focal spot 5b have to pass only few target material before leaving the target 4. Accordingly, the latter configuration yields a high integral photon flux. Preferably, an inventive X-ray apparatus is switchable between the two configurations of FIG. 2a and FIG. 2b electrically, for a quick change of X-ray beam characteristics between different measurements. Towards this end, a switching element (schematically indicated in the drawing of FIGS. 2a and 2b with a double arrow) is disposed, structured and dimensioned to switch the apparatus between two operation modes. In the first operation mode, the photon flux is maximized and in the second operation mode, a photon flux density is maximized. In the example shown, the diameter 2*R of the target 4 (representing its extension both in y and z) is more than a factor F of F=5 larger than both SY and SZ for the two shown configurations. FIG. 3a shows schematically parts of an embodiment of an inventive X-ray apparatus 1, wherein the electron beam 3 emitted by an electron beam source 2 passes through an electron beam deflection device 10 (or a combined electron beam deflection and focusing device), suitable for deflecting the electron beam 3 in x and y direction, and hits a liquid metal jet target 4 at a focal spot 5 where X-rays are generated. A continuous stream of liquid metal (for example consisting of gallium) is pumped through a circuit 13 by means of a pump 14 and directed via a nozzle 15 into a funnel type recovery unit 16; between the nozzle 15 and the recovery unit 16, the free metal stream constitutes the jet type target 4. If needed, the circuit 13 includes a tempering stage for heating and/or cooling the metal within the circuit 13 (not shown). Note that the jet has typically a diameter of about 50-250 μm, whereas the electron beam diameter is typically 100 μm or less. Marked with a dashed box are the parts of the X-ray apparatus 1 which should be located in a vacuum chamber 17; in particular, the electron beam 3 should only propagate inside the vacuum chamber 17. FIG. 3b illustrates an embodiment of an inventive X-ray apparatus 1 similar to the one shown in FIG. 1a, but with a curved target 4, namely a liquid metal jet target 4 (as shown in FIG. 3a, for example). The jet propagates in x direction, i.e. perpendicular to the electron beam 3 and the X-ray beam 8. By means of the electron beam deflection device 10 (or a combined electron beam deflection and focusing device) the electron beam 3 can be deflected in x and y direction. FIG. 3c shows an X-ray apparatus 1 also similar to the one shown in FIG. 1a, again with a liquid metal jet target 4. Here the size of the focal spot 5 can be changed by means of an electron beam focusing device 11. By changing the focal spot size (in x- and/or y-direction), the properties of the X-ray beam 8 downstream the X-ray optics 6 can be altered, in particular to obtain desired properties at the sample position 9. In particular, the properties of the X-ray beam 8 can be altered such that alternatively a maximum photon flux or a maximum photon flux density of the X-ray beam 8 can be obtained, in accordance with the invention. FIG. 4a and FIG. 4b illustrate in more detail a focal spot 5 of an electron beam on a target 4, here a liquid metal jet target, and their extension proportions in accordance with the invention. FIG. 4a shows a front view perpendicular to the z direction in which the electron beam propagates; FIG. 4b shows a cross-section in the plane perpendicular to the jet propagation direction x. The size (or extension) SX of the focal spot 5 in x direction is here more than five times smaller than the size (or extension) TX of the target 4 in x direction (Note that typically, the jet is a some tens of mm in x direction, which is the direction in which the jet propagates). In the example shown, the size (or extension) SY of the focal spot 5 is about 3 times smaller than the size (or extension) TY of the target 4 in y direction. The size (or extension) SZ of the focal spot in z direction (resulting from the propagation depth of electrons in the target material) is about 5 times smaller than the size (or extension) TZ of the target 4 in z direction here. So all in all, for all directions (x, y, z), the size of the focal spot 5 is at least about a factor F, with F=3, times smaller than the size of the target 4. Note that in accordance with the invention, a factor F=1.5 is sufficient, but a factor F=2 is preferred, and a factor F=5 is particularly preferred. FIG. 4c illustrates a focal spot 5 of elliptical shape. Here, too, the sizes SX, SY (and SZ, not shown) are at least about a factor F, with F=3, times smaller than the corresponding size TX, TY (and TZ, not shown) of the target 4. The target 4 is here of liquid metal jet type again. An elliptical electron beam may be the preferred choice because it can produce a circular “X-ray spot” (apparent focal spot) when viewed along the y direction (at an 90° angle with respect to the electron beam and the metal jet propagation direction); towards this direction the X-ray optics is placed then, receiving an X-ray beam with circular cross-section. FIG. 5 illustrates another embodiment of an inventive X-ray apparatus 1, similar to the one shown in FIG. 1a, but with capillary optics used as X-ray optics 6 for directing the X-ray beam 8 to the sample position 9. The capillary optics include one or more hollow, bent tubes (“capillaries”), at the internal surfaces of which total reflection of the x-rays occurs, so the X-rays may be guided by means of the capillaries (not shown in detail). The target 4 is of liquid metal jet type. In summary, the present invention proposes to align the focal spot of an electron beam and the focus of X-ray optics by deflecting the electron beam, thus allowing to do without mechanical fine alignment of the X-ray optics in an X-ray apparatus. Furthermore this invention allows to change the maximized X-ray beam properties downstream the X-ray optics by controlling shape and position of the focal spot on the target, in particular a target with a curved surface. |
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062194003 | abstract | An X-ray optical system includes an X-ray illumination system having at least two mirrors, a driving system for moving the at least two mirrors, a detecting system having at least one sensor, for detecting at least one of tilt and incidence position of X-rays in the X-ray illumination system, and a control system for controlling movements of the at least two mirrors by the driving system on the basis of detection by the detecting system. |
description | This application is a Continuation-In-Part application of, and claims the benefit of U.S. patent application Ser. No. 13/078,729, filed Apr. 1, 2011, which is hereby incorporated by reference in its entirety. The present disclosure relates to electromagnetic control of plasmas in fusion power reactor environments. In particular, it relates to liquid lithium (Li) first walls for electromagnetic control of plasmas in fusion power reactor environments. The present disclosure relates to an apparatus, system, and method for liquid lithium first walls for electromagnetic control of plasmas in fusion power reactor environments. In one or more embodiments, a method is disclosed for maintaining liquid lithium on a surface area of internal walls of a reactor chamber. The method involves installing at least one layer of at least one tile on the surface area of the internal walls of the reactor chamber. In one or more embodiments, a portion of the tile(s) facing the interior of the reactor chamber includes a plurality of channels. The method further involves applying an electric charge to the liquid lithium. Also, the method involves flowing the liquid lithium into the tile(s). In addition, the method involves circulating the liquid lithium through an interior network of the tile(s) to allow for the liquid lithium to flow into the channels and to reach the external surface of the tile(s) that faces the interior of the reactor chamber. Further, the method involves outputting the circulated liquid lithium from the tile(s). As illustrated above, in one or more embodiments, the interior network of the tile(s) and the channels of the tile(s) are manufactured from a high-temperature resistant, porous open-cell material. In at least one embodiment, the high-temperature resistant, porous open-cell material is a ceramic foam or a metal foam, and the channels are hydraulically and electrically separated from one another by solid ceramic strips. In other embodiments, the high-temperature resistant, porous open-cell material is a ceramic foam or a metal foam, and the channels are hydraulically and electrically separated from one another by ceramic foam strips treated with a lithium-blocking, electrically insulating coating. In at least one embodiment, as illustrated above, the tile(s) is manufactured from a high-temperature resistant, porous open-cell material. In some embodiments, the high-temperature resistant, porous open-cell material is a ceramic foam, and the channels are hydraulically and electrically separated from one another by strips of the ceramic foam that are treated with a lithium-blocking, nonconductive coating. As illustrated above, in one or more embodiments, the disclosed method further involves installing at least one magnetic coil between the tile(s) and the surface area of the internal walls of the reactor chamber. In at least one embodiment, at least one voltage source is used to provide the electric charge. In one or more embodiments, the reactor chamber is employed in a fusion reactor. In at least one embodiment, at least one tile has an irregular shape. In some embodiments, at least one tile has a regular shape. In at least one embodiment, as illustrated above, at least one tile contains at least one open cell in the interior of the tile(s), and the liquid lithium is circulated throughout the interior of the tile(s) via the open cell(s). In some embodiments, at least one tile has a constant porosity. In some embodiments, at least one tile has a varied porosity. As illustrated above, in one or more embodiments, at least one tile includes an input plenum, and the liquid lithium is inputted into the tile(s) via the input plenum. In at least one embodiment, the input plenum is a hollow piece of metal. In some embodiments, at least one tile includes an output plenum, and the liquid lithium is outputted from the tile(s) via the output plenum. In at least one embodiment, the output plenum is a hollow piece of metal. In one or more embodiments, the flow rate of the circulation of the liquid lithium within the interior network of at least one tile is varied over time. As illustrated above, in one or more embodiments, a system is disclosed for maintaining liquid lithium on a surface area of internal walls of a reactor chamber. The system involves at least one tile, where a portion of the tile(s) facing the interior of the reactor chamber includes a plurality of channels. The system also involves the reactor chamber, where at least one layer of the tile(s) is installed on the surface area of the internal walls of the reactor chamber. In addition, the tile(s) allows for electrically charged liquid lithium to be flowed into the tile(s). Additionally, the tile(s) further allows for the liquid lithium to be circulated throughout an interior network of the tile(s) to allow for the liquid lithium to flow into the channels and to reach an external surface of the tile(s) that faces the interior of the reactor chamber. Also, and the tile(s) further allows for the circulated liquid lithium to be outputted from the tile(s). In at least one embodiment, as illustrated above, a tile is disclosed for maintaining liquid lithium on a surface area of internal walls of a reactor chamber. The tile is manufactured from a high-temperature resistant, porous open-cell material. The tile includes a plurality of channels. The tile also includes at least one open cell in the interior of the tile for circulating electrically charged liquid lithium within the interior of the tile and the channels of the tile. The features, functions, and advantages can be achieved independently in various embodiments of the present inventions or may be combined in yet other embodiments. The methods and apparatus disclosed herein provide an operative system for electromagnetic control of plasmas in fusion power reactor environments. Specifically, this system relates to liquid lithium (Li) first walls for electromagnetic control of plasmas in fusion power reactor environments. In particular, the disclosed system uses a high-temperature, high-porosity open-cell material to maintain liquid lithium in a fusion energy reactor, thereby reducing damage to the internal reactor surfaces that have complex shapes. For example, the Boeing Rigid Insulation (BRI) material, which is a porous open-cell ceramic material with a high temperature tolerance and a good material compatibility, may be employed by the disclosed system. In addition, the disclosed system also provides for high-neutron flux exposure for the lithium for purposes of tritium breeding. BRI material is a porous, ceramic, fiber insulating material that comprises a unique combination of ceramic fibers, which are sintered together to form a low density, highly porous material with very low thermal conductivity. In addition, BRI material exhibits a high tensile strength and an outstanding dimensional stability. In particular, BRI material is manufactured from a combination of silica (SiO2) and alumina (Al2O3) fibers, and boron-containing powders, which help to aid in the fusion and sintering of the silica and alumina fibers. The preponderance of the insulative capability of the BRI material is provided by the silica fiber and the alumina fiber orientation. The BRI material exhibits very low thermal conductivity, particularly in the through-the-thickness direction. Further details discussing the composition of BRI and the method of producing BRI are disclosed in U.S. Pat. No. 6,716,782, which is expressly incorporated herein by reference. There are known difficulties in maintaining plasma control in fusion energy reactors. Among them, plasmas can be unstable at high power densities. Liquid lithium is known to help stabilize plasmas in reactor vessels. The plasma consists mostly of positive ions and negative electrons, and its outer sheath, near the reactor walls, is cooler than its core. In the sheath, the ions have a higher probability of acquiring electrons from the plasma and, thus, becoming neutral atoms than do ions in the core. Neutral atoms cannot be confined by magnetic fields, thus neutral atoms have a high probability of crossing the magnetic field that confines the plasma, and hitting the reactor vessel walls. In this process, the neutral atoms carry some energy from the plasma to the walls, thus causing a slight further cooling of the plasma sheath and a slight heating of the walls. In a fusion plasma, most of these neutral atoms are hydrogen, but other materials can be present, such as helium made by the fusion reactions and heavy elements (contaminants) that can be spalled off the reactor structure by accidental plasma impingement on the structure. If the walls are made of high temperature-tolerant ceramics or metals, the neutral atoms will stick to the walls for a short time, then drift back into the plasma sheath. However, the atoms reentering the plasma sheath from the walls are now quite cold in comparison to the plasma sheath, thus they cause considerable cooling of the plasma in the sheath. Normally, the plasma sheath is cooler than the plasma core, but if the sheath is cooled too much, the differential in temperature between the plasma core and the sheath increases the instability of the plasma. Lithium on the inside wall of the reactor tends to absorb and not release neutral atoms that drift into it. By absorbing and holding the neutral atoms that contact the walls, the lithium prevents the atoms from getting back into the plasma sheath as cold atoms, which helps the sheath to stay warmer and makes the overall plasma more stable. Currently, in experiments, liquid lithium is drizzled down the inside of the side walls of the reactor vessel from channels that lie just above where the side walls are nearly vertical (i.e. the channels lie just above the “equator” or midsection of the torus reactor vessel). Because of gravity, the liquid lithium does not stay in place, but rather runs down the side walls of the vessel from the channels, and is collected by other channels and drains that lie farther down in the vessel that remove the lithium. This particular method is able to coat the side walls of the vessel from the equator of the reactor to most of the way down to its bottom because gravity causes the lithium to flow down from the channels to the bottom of the reactor. But, this method clearly is not able to coat the side walls that are above the equator of the vessel. In the lowest parts of reactors, liquid lithium has also been used in pools and on coarse horizontal screens, neither of which method can be effectively applied to the upper walls. The disclosed system allows for liquid lithium to be maintained on the surface of all the inner walls of the reactor vessel. An additional advantage of the use of lithium on the reactor walls is that it is a low atomic number (low-Z) material. If high atomic number (high-Z) materials, such as iron from steel in reactor walls, enter into the plasma, their atoms can become electronically excited by absorbing kinetic energy from ions in the plasma. Typically, the excited high-Z materials lose their extra energy by radiating it as electromagnetic energy (photons). The plasma is transparent to most wavelengths of electromagnetic energy; thus, most photons emitted by excited high-Z materials escape from the plasma and are absorbed by the reactor walls. The net effect is an overall energy loss from the plasma and is called radiative cooling. The plasma gets colder and the reactor walls get hotter. That is the opposite of what is needed to maintain the fusion power reactions. Low-Z materials, such as lithium, have so few electrons that they have very few ways in which they can radiate energy, therefore, low-Z materials cause relatively little radiative cooling of the plasma. A further advantage to the use of lithium on the insides of fusion reactor walls is that one of the two elements of reactor fuel, tritium, is very rare naturally, but can be made efficiently by exposing lithium to the flux of high energy neutrons produced by the fusion reactor. Thus, if lithium can be placed in regions of the reactor close to the plasma where the neutron flux is most intense, the production of tritium from the lithium can be efficient. Because of liquid lithium's tendency to hold on to atoms of other materials in it, cycling the lithium through the reactor provides an effective way to introduce pure lithium into the reactor, produce tritium in the lithium, and remove the tritium from the reactor by pumping the tritiated lithium back out of the reactor and passing it through a chemical processing system that extracts the tritium from the lithium, thus providing tritium to fuel the reactor and clean lithium ready to be cycled once more through the reactor. The system of the present disclosure utilizes a porous, open-cell material that is capable of retaining liquid lithium in place on reactor vessel walls against gravity and electromagnetic forces. In addition, this material allows for the liquid lithium to be slowly pumped throughout the system in order to absorb contaminants from the plasma. During operation of the disclosed system, clean lithium is first pumped into the system to the inner surfaces of the reactor walls, where the lithium is exposed to the plasma. In that location, the clean lithium absorbs contaminants from the plasma. The contaminated lithium is then removed from the reactor, and is processed to remove the plasma contaminants from the lithium. After the contaminants are removed from the lithium, the cleaned lithium is re-circulated back into the system. Liquid lithium surfaces exposed to the plasma inside experimental tokamaks and other types of fusion energy experimental devices have been shown to help stabilize the plasma and to help the plasma maintain its high internal temperature. However, it should be noted that these reactor vessels typically are constructed to have very complex shapes as well as having many discontinuities and openings for various items, such as for instruments, vacuum pumping ports, and magnetic coils. Currently, no effective methods have been proposed for retaining liquid lithium on the inside of the reactor vessel walls that accommodates all the discontinuities and openings, and which retains the lithium against the effects of gravity and electromagnetic forces. The present disclosure teaches a method which can accommodate discontinuities, and which keeps slowly flowing liquid lithium in place on reactor walls regardless of the orientation of the reactor wall surface, and the effects of gravity and electromagnetic forces. To date, experiments with liquid lithium adjoining fusion plasmas have been more focused on the effect of lithium on the plasma than on how to build a liquid lithium wall. Five types of ad hoc approaches have been used to facilitate liquid lithium-hydrogen plasma interaction experiments. These five approaches are: (1) pools of liquid lithium placed in trays at the bottom of the toroidal reactor vessel, (2) metal screens wetted with liquid lithium that are placed horizontally at the bottom of the vessel, (3) a band placed about the mid-plane of the reactor vessel has liquid lithium flowing down its inner surface from the top of the band to the bottom of the band, (4) confining the plasma in spherical and cylindrical reactor vessels that are physically rotated so as to cause the liquid lithium to continually recoat the inner surface walls of the vessel from a pool at the bottom of the vessel, and (5) coating part of the vertical portion of the inside of the reactor vessel with a porous metal, which may be deposited for example by flame or plasma spraying, and pumping molten lithium through the porous metal. The first two listed approaches have limitations of only producing lithium surfaces for a small area in the bottom of the reactor. The third approach only coats a band about the middle of the reactor, and requires high flow rates to keep the surface of the band coated. High flow rates increases the pumping power required to operate the reactor, which subtracts from any energy the reactor might produce. The fourth approach is not being easily being employed by a toroidal vessel, which has the most effectively shaped magnetic fields for containing plasmas. Continually rotating the walls of a toroidal vessel is impossible because of the rigid materials used for the construction of these vessels. In addition, the fourth approach requires portions of the inside of the reactor vessel to constantly move, which interferes with the placement and the use of other devices that must be present within the vessel wall, such as vacuum pumping ports, sensors, and magnetic coils. The fifth approach has the drawbacks of having little control over where the lithium flows, the fact that porous metal is a high-Z material, and the fact that the use of large areas of porous metal precludes having any control over electrical current flows on the inner surface of the reactor. The present disclosure employs tiles manufactured from high-temperature, open-cell sponge-like material (e.g., the Boeing Rigid Insulation (BRI) material) to retain liquid lithium in place against gravity and electromagnetic forces, and to allow for the liquid lithium to be slowly pumped throughout the system in order to remove contaminants from the plasma. There are multiple advantages to this approach. A first advantage is that the tiles can be manufactured to be small in size so that the inside of the toroidal vessel can be tiled with a mosaic of liquid lithium filled tiles despite the complex shape of the inside of the reactor vessel. A second advantage is that the material of the tiles (e.g., a porous ceramic material with open cells) is resistant to the high temperatures to which the tiles will be exposed to when the plasma is present inside of the reactor vessel. A third advantage is that the material of the tiles (e.g., porous a ceramic material with open cells) is resistant to the corrosive effects of lithium. A fourth advantage is that the construction of the tiles can be tailored to produce pore sizes and/or open channels that are optimal to the retention and flow of liquid lithium. In addition, a fifth advantage is that, if plasma disruptions cause the plasma to impact the tiles so intensely that the outer surface of lithium boils away, the high permeability of the tiles will allow more lithium to wick to the surface of the tile. A sixth advantage is that, in the event that some of the tile itself is removed by a plasma impact, the depth of the tile will allow for the tile to continue to function and, thus, several plasma impacts on a tile can be tolerated before the tile would need to be replaced. A seventh advantage is that, in the event that part of a tile is ablated by the plasma, the materials that the tile is manufactured from are mostly of low nuclear weight elements, which will have a less adverse effect on the plasma than materials of high weight metals. An eighth advantage is that, in the event that a portion of a tile is ablated, the portion of the tile that is ablated will simply be an empty space filled with liquid lithium. As such, it is evident that the use by the disclosed system of tiles, which are manufactured from a high-temperature, porous material, to retain liquid lithium on the reactor vessel walls has many beneficial advantages. In fusion energy experiments, electric coils producing modulated magnetic fields, that are installed facing the plasma, have been shown to be helpful in controlling instabilities in the plasma. However, it should be noted that fusion power reactors will have internal environments so severe that placing electrical coils near the plasma is likely impractical. As alluded to above, molten lithium has been shown in experiments to be one material that is able to face the plasma. In one or more embodiments of the present disclosure, instead of employing electrical coils facing the plasma, an electric charge is applied to the liquid lithium, which faces the plasma, in order to aid in controlling instabilities in the plasma. In some embodiments, electric coils are installed behind the structures containing the liquid lithium to work in conjunction with the electrically charged liquid lithium to help in controlling instabilities in the plasma. In the following description, numerous details are set forth in order to provide a more thorough description of the system. It will be apparent, however, to one skilled in the art, that the disclosed system may be practiced without these specific details. In the other instances, well known features have not been described in detail so as not to unnecessarily obscure the system. FIG. 1 is an illustration of the interior of a fusion power reactor 100, in accordance with at least one embodiment of the present disclosure. In this figure, it can be seen that the fusion power reactor 100 is of a torus shape. It should be noted that the system of the present disclosure can be used with various different types and shapes of fusion power reactors. The first wall of the fusion power reactor 100 is lined with small tiles 110 that are each manufactured from a high temperature-tolerant, porous material. These small tiles 110 allow for liquid lithium to coat the surface of the walls of the reactor vessel 100. The liquid lithium helps to stabilize the plasma in the reactor vessel 100, and helps the plasma maintain its high internal temperature. FIG. 2 shows a top view of a single tile 200 for maintaining liquid lithium on the surface area of the internal walls of a reactor chamber, in accordance with at least one embodiment of the present disclosure. The tile 200, which is manufactured from a high-temperature-resistant, porous material with open cells, is installed onto the reactor vessel wall 240. In this figure, the tile 200 is shown to include an input plenum 260 and an output plenum 280. Both the input plenum 260 and the output plenum 280 are a single hollow piece of non-porous material (e.g., a metal). During operation of the system, clean liquid lithium is inputted into the tile 200 through the input plenum 260. The liquid lithium is flowed into the input plenum 260 of the tile 200 via pressure being applied at the input plenum 260 and/or a vacuum being present at the output plenum 280. Various types of pumps may be employed by the system for applying pressure at the input plenum 260 of the tile 200 including, but not limited to, a propeller pump, a centrifugal pump, and a piston pump. The clean liquid lithium circulates within the interior network of open cells or channels throughout the body 250 of the tile 200. The clean liquid lithium seeps through the open cells of tile 200 to reach the porous external surface 220 of the tile 200 that faces the interior cavity of the reactor vessel, which contains the hot, tenuous plasma 230. The direction of the flow of the liquid lithium within the body 250 of the tile 200 is denoted by arrow 270. The clean liquid lithium that lies on the porous external surface 220 of the tile 200 absorbs contaminants from the plasma 230. This newly contaminated liquid lithium is then removed from the tile 200 via the output plenum 280. After the contaminated liquid lithium is removed from the tile 200, the liquid lithium is processed to remove the contaminants from the liquid lithium. The resulting cleaned liquid lithium is then re-circulated back into the system. It should be noted that in alternative embodiments, the tile 200 may not specifically include an input plenum 260 and/or an output plenum 280 as is depicted in FIG. 2, but rather may have at least one open cell or channel in its interior for the liquid lithium to be inputted into the tile 200 and/or to be outputted from the tile 200. FIG. 3 illustrates a top view of a configuration 300 of four of the tiles 310 of FIG. 2 that are installed next to one another, in accordance with at least one embodiment of the present disclosure. In this figure, it is shown that the tiles 310 are able to be installed adjacent to one another along the curved surface of the reactor vessel wall 330. When the tiles 310 are installed in this configuration, the porous external surface 340 of the tiles 310 that faces the interior cavity of the reactor vessel containing the plasma 320 is shown to form a curved surface area. FIG. 4 depicts a cross-sectional side view of a tile 410 for maintaining liquid lithium on the surface area of the internal walls of a reactor chamber that has a uniform porosity, in accordance with at least one embodiment of the present disclosure. In this figure, the tile 410 is shown to have an input plenum 430 and an output plenum 440. The tile 410 is also depicted to be manufactured to have a uniform porosity 420. In addition, the direction of the flow of the liquid lithium within the body of the tile 410 is denoted by arrow 450 in this figure. FIG. 5 illustrates a cross-sectional top view of the tile 410 of FIG. 4, in accordance with at least one embodiment of the present disclosure. This figure simply shows another cross-sectional view of the tile 410, which has a uniform porosity 420. In addition, it should be noted that, in some embodiments, the borders of the side areas 510, 520 of the tile 410 are manufactured from the same non-porous material that is used to manufacture the input plenum 430 and the output plenum 440. FIG. 6 shows a cross-sectional top view of a tile 610 for maintaining liquid lithium on the surface area of the internal walls of a reactor chamber that has a non-uniform porosity, in accordance with at least one embodiment of the present disclosure. In this figure, the tile 610 is shown to have an input plenum 630 and an output plenum 640. The tile 610 is illustrated to be manufactured to have a non-uniform porosity 620. In this figure, the porosity of the body of the tile 610 is shown to gradually lessen from the external surface 660 of the tile 610 that faces the plasma 670 to the input and output plenums 630, 640. Also in this figure, arrow 650 illustrates the direction of the flow of the liquid lithium within the body of the tile 610. FIG. 7 depicts a cross-sectional side view of a tile 710 for maintaining liquid lithium on the surface area of the internal walls of a reactor chamber, where an electric charge is applied to the liquid lithium, and the tile 710 includes channels 705 for the electrically charged liquid lithium to flow, in accordance with at least one embodiment of the present disclosure. The tile of FIG. 7 differs from the basic tile construction shown in FIGS. 2 and 4 in that the foam region containing the flowing lithium is divided into many narrow channels 705 that are exposed to the plasma 770. In the plasma-facing region, the lithium channels 705 are electrically isolated from each other. At the input to and the output from the plasma-facing region, the channels 705 are physically connected together, which makes them electrically and hydraulically in parallel. In alternative embodiments, the channels 705 can have electrically isolated “return” channels that are built deeper into the tile 710, which allows for the plasma-facing channels 705 to be electrically and hydraulically in series. The basic tile construction of FIG. 4 consists of a monolithic foam tile 410 that can be manufactured from an electrically conductive (e.g. metal) foam or an electrically insulating (e.g. ceramic) foam through which the liquid lithium will flow. For the tile 710 of FIG. 7, the plasma-facing portion of the tile 710 is divided into channels 705 through which the liquid lithium will flow, and which are electrically isolated from each other. There are several ways that the electrically isolated channels 705 can be manufactured in the tile 710. A first way is that the tile 710 surface is manufactured to have regions of ceramic or metal foam strips 705 through which the lithium will flow. These regions 705 are hydraulically and electrically separated by solid ceramic strips 720, which are able to electrically insulate the lithium channels 705 from each other and which exclude the electrically conductive lithium from their interiors. A second way that the electrically isolated channels 705 can be manufactured in the tile 710 is that, similar to the first way, the tile 710 surface is manufactured to have regions of ceramic or metal foam strips 705 through which the lithium will flow. These regions 705 are separated by ceramic foam strips 720 that have been treated to prevent the penetration of lithium into them. The treatment must itself be electrically insulating. By preventing the penetration of lithium into the treated region, and by maintaining the insulating properties of the ceramic, the treatment allows the strips 720 to electrically and hydraulically insulate the lithium channels 705 from each other. A third way that the electrically isolated channels 705 can be manufactured in the tile 710 is that the tile 710 surface is manufactured completely from ceramic foam. In the foam, narrow strips 720 are be treated with a lithium-blocking, nonconductive coating, which will prevent liquid lithium from wetting and, thus, penetrating those regions. Liquid lithium will flow freely in the untreated strips 705 between the treated strips 720. The reason for confining the lithium to thin electrically and hydraulically isolated strips 705 is to produce an array of parallel “wires” of liquid lithium on the surface of the tile 710. The wires 705 will have a voltage applied across them so that the lithium channels 705 will carry electric currents in the well defined regions and directions. This channel configuration allows for the currents in the lithium to produce a controlled magnetic field in the plasma 770 adjacent to the tile 710. In FIG. 7, the tile 710 is shown to be attached to some form of plenums 780, 790 or plumbing that provide the input 730 and output 740 of the lithium of the tile 710. The input plumbing 780 and output plumbing 790 are part of an electrical circuit because they are each connected to a voltage source 760. As such, the input plenum 780 and the output plenum 790 are manufactured to be electrically insulated from each other and their surroundings. During operation, the liquid lithium, which has been thermally conditioned, is collected in an electrically insulated reservoir (not shown). The lithium is then pumped from the reservoir either by a pump (not shown) into the input plumbing 780 from the reservoir or by pressurizing the gas in the space above the lithium in the reservoir. The initial portion 750 of the input plumbing 780 must be electrically insulated from the outside world, or manufactured from an insulating material, such as ceramic tubing. At some point in the hydraulic path to the tile 710, a section 755 of the input plumbing 780 is manufactured of a conductive material, such as metal, and is electrically connected to an electrical power supply 760 (e.g., a variable voltage supply 760 as shown in this figure). The electrical power supply 760 will supply the electric current input for the lithium. From the electrical start point, the final portion 765 of the input plumbing 780 is insulated, and will carry the liquid lithium to the inside of the tile 710. After flowing into the tile 710, the liquid lithium will flow in the parallel channels 705 of the tile 710. The direction of the flow of the liquid lithium within the channels 705 is denoted by arrow 740 in this figure. The liquid lithium will then enter into the return plumbing 790. That plumbing 790 will mostly be insulated (i.e. regions 775 and 795), but will have one conductive section 785, which will be connected to the return path of the electric circuit (i.e. connected to the other end of the electrical power supply 760). If the electrical return point is at facility ground potential, the lithium can then flow on to any heat exchangers or filtering/cleaning processes that will be needed before the lithium can re-enter into the input reservoir (not shown). The lithium re-entering the input reservoir will need to enter by some means, such as dripping, so that the entering lithium, which will be at electrical ground potential, will not form an electrically conductive path between the entrance to the reservoir at ground potential and the liquid lithium in the bottom of the reservoir, which will be at the electrical potential of the input to the tile 710. A pneumatic process that inserts insulating barriers (i.e. bubbles) in lithium flowing in an insulating pipe could also possibly provide electrical isolation between the incoming and outgoing lithium similar to that provided by dripping, as long as the lithium does not wet the walls of the insulating pipe. If the electrical return point is not at facility ground potential, such as what would be the case if the lithium electrical circuit operates from a voltage above ground at the input and a voltage below ground at the return, then the lithium leaving the electrical return point must also go through some process, such as dripping, which will electrically isolate the returning lithium from the heat exchangers and filtering or cleaning equipment. In this system, after conditioning, the lithium, which will be at facility ground potential must go through electrical isolation again before reaching the bottom of the input reservoir, just as it did in the previous approach. Returning to the lithium-containing tile 710 itself, it should be noted that this particular tile 710 design has three key features. The first key feature of this design is that because of the geometry of the lithium-filled regions 705 (i.e. the channels 705) and the lithium-free regions 720, electric currents impressed on the lithium circuit are forced to all flow either in parallel or anti-parallel to the flow of the lithium itself, depending upon the polarity of the impressed voltage. The second key feature of this design is that by placing a series of tiles 710 in a closed ring on the surface of the plasma-facing wall of a reactor, the sum of the local magnetic fields produced by all the electrical currents in the tiles 710 will produce a large net magnetic field, which can be used to manipulate the plasma 770. The third key feature of this design is that if the tiles 710 are placed between a plasma control electromagnet (e.g., an electromagnetic coil) and the plasma 770 (refer to FIGS. 10, 11, and 12), the surface electrical current paths in the tiles 710 are aligned with the currents in the plasma control magnet, and the surface electrical currents in the tiles 710 and the electrical currents in the magnet are in parallel, then the electrical currents in the liquid lithium will enhance the magnetic field produced by the plasma control magnet; whereas, without the controlled, externally driven electrical currents in the liquid lithium, the conductive liquid lithium would support local electrical currents that would respond to, but oppose, changes in the current flow in the plasma control magnet. FIG. 8 depicts a side view of the tile 710 of FIG. 7 illustrating the channels 705, in accordance with at least one embodiment of the present disclosure. In this figure, the channels 705 are shown to be isolated by thin strips 720. The liquid lithium flows in the channels 705 in a direction that is denoted by arrow 740. FIG. 9 illustrates a top view of a configuration 900 of three rows of four of the tiles 710 of FIG. 7 that are installed next to one another, in accordance with at least one embodiment of the present disclosure. In this figure, it is shown that the tiles 710 are able to be installed adjacent to one another in a row along the curved surface of the reactor vessel wall. When the tiles 710 are installed in this configuration, the porous external surface of the tiles 710 that faces the interior cavity of the reactor vessel containing the plasma 770 is shown to form a curved surface area 910. In one or more embodiments of the present disclosure, multiple rows of the tiles 710 may be installed on the surface area of the internal walls of the reactor chamber. The tiles 710 may be connected hydraulically and electrically in series, as is shown for each row of tiles 710 in FIG. 9, in which the output plumbing of one tile 710 connects to the input plumbing of another tile 710, and one end of each row of the tiles 710 is shown to be connected to a positive terminal of a power supply (not shown) via input plumbing 780, and the opposite end of each row of the tiles 710 is shown to be connected to a negative terminal of the power supply (not shown) via return plumbing 790. It should be noted that in alternative embodiments, each of the tiles 710 in a row may be connected to a separate dedicated power supply for that particular tile 710. How many tiles could be connected in series would depend upon a trade-off among the pressures needed to pump the lithium, the voltages needed to drive the needed electrical current through the lithium, and the ease or difficulty of access to the lithium tiles 710 for plumbing and electrical connections. FIG. 10 is an illustration of the interior of a toroidal fusion power reactor 100 that includes two magnetic coils 1010 mounted behind the tiles 710 of FIG. 7, in accordance with at least one embodiment of the present disclosure. As previously mentioned, electric coils 1010 installed behind tiles 710 containing electrically charged liquid lithium work in conjunction with the electrically charged liquid lithium to help in controlling instabilities in the plasma. The electric coils 1010 may be manufactured from various different materials including, but not limited to, copper alloy materials. In this figure, the tiles 1020 are shown to not have a magnetic coil 1010 mounted behind them. FIG. 11 shows a close-up view of one of the magnetic coils 1010 of FIG. 10 depicted along with a number of tiles 710, 1020, in accordance with at least one embodiment of the present disclosure. In this figure, the magnetic coil 1010 is shown to be installed behind a number of tiles 710 with channels 705 that contain electrically charged liquid lithium. The direction of the electrical current flow of the magnetic coil 1010 is denoted by arrows 1030. In this figure, the tiles 1020 that do not have the magnetic coil 1010 installed behind them are shown to be the type of tile 410 of FIG. 4, which does not include channels 705. However, it should be noted that in other embodiments, the tiles 1020 that do not have a magnetic coil 1010 installed behind them may be the type of tile 410 of FIG. 4, which does not include channels 705, and/or the type of tile 710 of FIG. 7, which does include channels 705. FIG. 12 shows a cross-sectional top view of the magnetic coil 1010 of FIG. 11, in accordance with at least one embodiment of the present disclosure. In this figure, the magnetic coil 1010 is shown to be installed in between a number of tiles 710, which include channels 705, and the internal wall 1210 (e.g., a steel alloy wall) of the reactor chamber. The direction of the electric current of the liquid lithium within the channels 705 of the tiles 710 is denoted by arrows 1220, and the direction of the current flow of the magnetic coil 1010 is denoted by arrows 1030. Although certain illustrative embodiments and methods have been disclosed herein, it can be apparent from the foregoing disclosure to those skilled in the art that variations and modifications of such embodiments and methods can be made without departing from the true spirit and scope of the art disclosed. Many other examples of the art disclosed exist, each differing from others in matters of detail only. Accordingly, it is intended that the art disclosed shall be limited only to the extent required by the appended claims and the rules and principles of applicable law. |
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description | The present invention relates to a nuclear reactor fuel integrity monitor which monitors a nuclear reactor fuel state. For example, a pressurized water reactor (PWR) or a boiling water reactor (BWR) uses light water as a nuclear reactor coolant and a neutron moderator. In a normal operation mode of the nuclear reactor, the nuclear reactor coolant cools a cladding pipe of nuclear reactor fuel so as to decrease the temperature of the cladding pipe. In order to check the integrity of the nuclear reactor fuel when operating the nuclear reactor in the normal mode, a nuclear reactor coolant and a gas dissolved in the nuclear reactor coolant are periodically sampled from the nuclear reactor. Then, a concentration of a specific radionuclide which may be discharged from the nuclear reactor fuel to the nuclear reactor coolant and the dissolved gas is monitored. For example, Patent Literature 1 discloses a radioactive gas measuring device which extracts the gas dissolved in the nuclear reactor coolant, seals the gas in a sample chamber, and monitors a radiation level of a radioactive gas. Patent Literature 1: Japanese Patent Application Laid-open No. 2001-235546 However, in the operation in which the gas dissolved in the nuclear reactor coolant is extracted, is sealed in the sample chamber, and is measured by the radioactive gas measuring device, some time is necessary for extracting the dissolved gas. Further, the operation of sealing the dissolved gas in the sample chamber is an operation which needs an operator's labor. The invention is made in view of such circumstances, and it is an object of the invention to provide a nuclear reactor fuel integrity monitor capable of periodically sampling a coolant or a dissolved gas of a nuclear reactor from a sampling point without an operator's labor and measuring a concentration of a specific radionuclide included in the coolant or the dissolved gas of the nuclear reactor. In order to solve the above-described problems and attain the object, according to an aspect of the invention, there is provided a nuclear reactor fuel integrity monitor including: a γ-ray detector which detects γ-ray of a specific radionuclide of a subject measurement medium of a nuclear reactor; a sample container which retains the subject measurement medium therein and surrounds the circumference of the γ-ray detector; and a measurement control device which performs a control so that a predetermined amount of the subject measurement medium is introduced into the sample container and calculates a concentration of the specific radionuclide from γ-ray data per each unit time detected by the γ-ray detector and a volume of the subject measurement medium introduced into the sample container. Since the sample container is formed in a shape in which the subject measurement medium surrounds the γ-ray detector, it is possible to increase the volume of the subject measurement medium which may be detected by the γ-ray detector. As a result, since the amount of the γ-ray which may be detected by the γ-ray detector 31 increases, there is no need to perform the operation of concentrating the subject measurement medium so as to improve the detection efficiency of the γ-ray detector 31. Further, since the supply of the subject measurement medium is controlled so that a predetermined amount of the subject measurement medium is introduced into the sample container, it is possible to periodically measure the γ-ray of the subject measurement medium introduced into the sample container without the operator's labor. Further, according to the invention, it is possible to calculate the concentration for each specific radionuclide. As a desirable aspect of the invention, the sample container may be formed in a hollow shape and be wound on the γ-ray detector so as to surround the γ-ray detector in a spiral shape. Since the sample container is formed in a hollow shape, it is possible to efficiently discharge the subject measurement medium inside the sample container without leaving the subject measurement medium therein. Further, since the sample container is wound on the γ-ray detector so as to surround the γ-ray detector in a spiral shape, it is possible to increase the volume of the subject measurement medium which may be detected by the γ-ray detector. As a result, it is possible to ensure the detection output of the γ-ray detector. As a desirable aspect of the invention, the subject measurement medium may be a gas and the specific radionuclide may be radioxenon. When the radioxenon is discharged from the nuclear reactor fuel, it is possible to generate an alarm at an early timing. Advantageously, in the nuclear reactor fuel integrity monitor, the sample container includes a recess portion and the γ-ray detector is disposed in the recess portion. Accordingly, the subject measurement medium surrounds the circumference of the γ-ray detector and the amount of the γ-ray which is received by the γ-ray detector from the subject measurement medium increases. Advantageously, in the nuclear reactor fuel integrity monitor, the subject measurement medium is a coolant of the nuclear reactor and the specific radionuclide is radioactive iodine. When the radioactive iodine is discharged from the nuclear reactor fuel, it is possible to generate an alarm at an early timing. Advantageously, in the nuclear reactor fuel integrity monitor, an inlet side of the sample container is provided with an activated alumina column, and the coolant passing through the activated alumina column is introduced into the sample container. Even when the coolant of the nuclear reactor is periodically sampled without the operator's labor, the radioactive fluorine as the hindering nuclide is not supplied to the γ-ray detector while being mixed with the radioactive iodine. Then, the radioactive iodine is measured by the γ-ray detector without the interference of the radioactive fluorine. According to another aspect of the present invention, a nuclear reactor fuel integrity monitor includes: a first γ-ray detector which detects γ-ray of a first specific radionuclide contained in a gas as a subject measurement medium of a nuclear reactor; a first sample container which retains the gas inside a hollow portion thereof and is wound on the first γ-ray detector so as to surround the first γ-ray detector in a spiral shape; a second γ-ray detector which detects γ-ray of a second specific radionuclide contained in a coolant as a subject measurement medium of the nuclear reactor; a second sample container which retains the coolant therein and includes a recess portion for disposing the second γ-ray detector therein; and a measurement control device which performs a control so that a predetermined amount of the gas is introduced into the first sample container and a predetermined amount of the coolant is introduced into the second sample container, and calculates concentrations of the first and second specific radionuclides from γ-ray data per unit time detected by the first and second γ-ray detectors and volumes of the subject measurement mediums introduced into the first and second sample containers. According to the invention, it is possible to monitor the specific radionuclide of the nuclear reactor fuel by both the gas and the cooling water. As a result, it is possible to recognize the nuclear reactor fuel state over again. According to the invention, the coolant or the dissolved gas of the nuclear reactor is periodically sampled without the operator's labor, and hence the concentration of the specific radionuclide included in the coolant or the dissolved gas may be measured. A mode for carrying out the invention (an embodiment) will be described in detail by referring to the drawings. The invention is not limited to the content described in the embodiment below. Further, the constituents described below include a constituent that may be easily supposed by the person skilled in the art and a constituent that has substantially the same configuration. Further, the constituents described blow may be appropriately combined with one another. FIG. 1 is a schematic diagram illustrating a nuclear plant. In the embodiment, a nuclear plant 1 is a nuclear generation facility. A nuclear reactor 2 which constitutes the nuclear plant 1 is a pressurized water reactor. In the nuclear plant 1, the nuclear reactor 2, a steam generator 3, a pressurizer 4, a nuclear reactor coolant pump 5, and a regeneration heat exchanger 11 are arranged inside a containment 1W. Further, a turbine 8, a condenser 9, and a power generator 10 are arranged outside the containment 1W. In the nuclear reactor 2, a nuclear reactor fuel 2C is disposed inside a pressure container. Further, a nuclear reactor coolant (which corresponds to cooling water, for example, light water) C1 is charged into the pressure container. The nuclear reactor coolant pump 5 is connected to the nuclear reactor 2 by a first nuclear reactor coolant supply passageway 6A, and the nuclear reactor 2 is connected to the steam generator 3 by a second nuclear reactor coolant supply passageway 6B. Further, the steam generator 3 is connected to the nuclear reactor coolant pump 5 by a nuclear reactor coolant collecting passageway 6C. The nuclear reactor coolant C1 which is ejected from the nuclear reactor coolant pump 5 passes through the first nuclear reactor coolant supply passageway 6A so as to be supplied into the pressure container of the nuclear reactor 2. Then, the nuclear reactor coolant C1 is heated by the thermal energy which is generated by the nuclear fission reaction of the nuclear reactor fuel 2C disposed inside the pressure container. The heated nuclear reactor coolant C1 passes through the second nuclear reactor coolant supply passageway 6B so as to be supplied to the steam generator 3. Then, the nuclear reactor coolant C1 passes through a heat transfer pipe 3T of the steam generator 3, flows out of the steam generator 3, passes through the nuclear reactor coolant collecting passageway 6C, returns to the nuclear reactor coolant pump 5, and is ejected again from the first nuclear reactor coolant supply passageway 6A into the pressure container of the nuclear reactor 2. The steam generator 3 includes a plurality of the heat transfer pipes 3T, and a secondary coolant C2 outside the heat transfer pipe 3T is heated and boiled by the nuclear reactor coolant C1 which flows inside the heat transfer pipe 3T, so that pressurized hot steam of the secondary coolant C2 is generated. The steam generator 3 is connected to the turbine 8 by a steam supply passageway 7S, and the condenser 9 is connected to the steam generator 3 by a secondary coolant collecting passageway 7R. Accordingly, the pressurized hot steam of the secondary coolant C2 which is generated by the steam generator 3 passes through the steam supply passageway 7S so as to be supplied to the turbine 8, thereby driving the turbine 8. Then, power is generated by the power generator 10 which is connected to the driving shaft of the turbine 8. The secondary coolant C2 which is used to drive the turbine 8 is liquefied by the condenser 9, and passes through the secondary coolant collecting passageway 7R so as to be sent to the steam generator 3 again. The nuclear reactor 2 is a pressurized water reactor, and the pressurizer 4 is connected to the second nuclear reactor coolant supply passageway 6B. Then, the pressurizer 4 pressurizes the nuclear reactor coolant C1 inside the second nuclear reactor coolant supply passageway 6B. With such a structure, the nuclear reactor coolant C1 is not boiled even when the coolant is heated by the thermal energy generated by the nuclear fission reaction of the nuclear reactor fuel 2C, and circulates the nuclear reactor 2 and the cooling system thereof in a liquid state. Here, the cooling system of the nuclear reactor 2 is a system which includes the nuclear reactor coolant pump 5, the first nuclear reactor coolant supply passageway 6A, the second nuclear reactor coolant supply passageway 6B, the steam generator 3, and the nuclear reactor coolant collecting passageway 6C and through which the nuclear reactor coolant C1 flows. In order to eliminate impurities included in the nuclear reactor coolant C1, a desalination tower 16 is installed. The desalination tower 16 includes a first desalination tower 16A and a second desalination tower 16B, and is installed outside the containment 1W. The first desalination tower 16A is a coolant hot-bed desalination tower, and the second desalination tower 16B is a coolant cation desalination tower. The nuclear reactor coolant C1 which is extracted from the inlet side (the upstream side) of the nuclear reactor coolant pump 5 is supplied from the cooling system of the nuclear reactor 2 to the desalination tower 16 so that a desalination process is performed thereon, and the nuclear reactor coolant C1 returns to the outlet side (the downstream side) of the nuclear reactor coolant pump 5. A desalination process system of the nuclear reactor coolant C1 includes a nuclear reactor coolant extracting passageway 13A, the regeneration heat exchanger 11, a nuclear reactor coolant passageway 13B, a non-regeneration heat exchanger 12, a nuclear reactor coolant passageway 13C, the desalination tower 16, a nuclear reactor coolant passageway 13D, a volume control tank 14, and nuclear reactor coolant returning passageways 13E and 13F. The nuclear reactor coolant extracting passageway 13A connects the nuclear reactor coolant collecting passageway 6C constituting the cooling system of the nuclear reactor 2 to the regeneration heat exchanger 11. The regeneration heat exchanger 11 is connected to the non-regeneration heat exchanger 12 by the nuclear reactor coolant passageway 13B, and the non-regeneration heat exchanger 12 is connected to the desalination tower 16 by the nuclear reactor coolant passageway 13C. The desalination tower 16 is connected to the volume control tank 14 by the nuclear reactor coolant passageway 13D, and the volume control tank 14 is connected to the regeneration heat exchanger 11 by the nuclear reactor coolant returning passageway 13E. Further, the regeneration heat exchanger 11 is connected to the first nuclear reactor coolant supply passageway 6A by the nuclear reactor coolant returning passageway 13F. The nuclear reactor coolant returning passageway 13E is provided with a charging pump 15. The nuclear reactor coolant C1 is extracted from the nuclear reactor coolant extracting passageway 13A, that is, the inlet side (the upstream side) of the nuclear reactor coolant pump 5. The nuclear reactor coolant C1 which is extracted from the cooling system of the nuclear reactor 2 is guided to the regeneration heat exchanger 11, and passes through the nuclear reactor coolant passageway 13B, the non-regeneration heat exchanger 12, and the nuclear reactor coolant passageway 13C so as to be guided to the desalination tower 16, where the desalination process is performed. The nuclear reactor coolant C1 subjected to the desalination process passes through the nuclear reactor coolant passageway 13D and is temporarily accumulated in the volume control tank 14. Then, the nuclear reactor coolant is sent to the regeneration heat exchanger 11 by the charging pump 15 which is installed in the nuclear reactor coolant returning passageway 13E. The nuclear reactor coolant C1 which passes by the regeneration heat exchanger 11 passes through the nuclear reactor coolant returning passageway 13F so as to be returned to the first nuclear reactor coolant supply passageway 6A, that is, the outlet side (the downstream side) of the nuclear reactor coolant pump 5. The nuclear reactor coolant C1 inside the volume control tank 14 is separated into a liquid phase and a gas phase, and a gas phase portion 14a including a gas G exists along with the nuclear reactor coolant C1. Then, the volume control tank 14 is provided with a gas phase sampling point 21 which may sample a part of the gas G of the gas phase portion 14a as the dissolved gas. The installation of the gas phase sampling point 21 is not limited to the volume control tank 14 as long as there is a tank which has the nuclear reactor coolant C1 therein and in which the nuclear reactor coolant exists while being separated into a liquid phase and a gas phase. The nuclear reactor coolant passageway 13C which constitutes the desalination process system of the nuclear reactor coolant C1 is provided with a nuclear reactor coolant sampling point 25 which may sample a part of the nuclear reactor coolant C1. Furthermore, the installation of the nuclear reactor coolant sampling point 25 is not limited to the nuclear reactor coolant passageway 13C. FIG. 2 is a schematic diagram illustrating an example of the nuclear reactor fuel integrity monitor according to the first embodiment. FIG. 3 is a schematic diagram illustrating a measurement control device. FIG. 4 is a flowchart illustrating a procedure of monitoring the nuclear reactor fuel state. As illustrated in FIG. 2, a nuclear reactor fuel integrity monitor 100 includes a radioactive noble gas detecting device 30, a measurement control device 80, opening and closing valves 41 and 42, and valve control units 43 and 44. The radioactive noble gas detecting device 30 is disposed between the gas phase sampling point 21 and an exhaust process connection point 45 and is connected to the gas phase sampling point 21 and the exhaust process connection point 45 through a gas passageway 24A, a gas passageway 24C, and a gas passageway 46. The gas G which is supplied from the gas phase sampling point 21 is branched from the gas passageway 24A into the gas passageway 24C and a gas passageway 24B by a branch point 23. The gas G which is branched to the gas passageway 24B is supplied to an analysis device 22. The radioactive noble gas detecting device 30 includes a γ-ray detector 31, a sample loop 32 as a sample container, and a lead shield 33. Radioactive argon (Ar-41), radioactive krypton (Kr-85, Kr-85 m, Kr-87), and radioxenon (Xe-133, Xe-135) may be dissolved as a radioactive noble gas in the nuclear reactor coolant C1. In order to monitor the nuclear reactor fuel state, it is important to monitor the concentration of radioxenon (Xe-133, Xe-135). The radioactive noble gas detecting device 30 may detect radioxenon (Xe-133, Xe-135). As the γ-ray detector 31, for example, a NaI scintillation detector or a Ge-semiconductor detector is used. The NaI scintillation detector and the Ge-semiconductor detector may obtain a γ-ray spectrum. Since the detection efficiency and the maintenance workability of the NaI scintillation detector are better than those of the Ge-semiconductor detector, it is desirable to use the former as the γ-ray detector 31. The resolution of the NaI scintillation detector is about 80 keV at FWHM (Full width at half maximum). Therefore, NaI scintillation detector may discriminate and measure the γ-ray of 81.0 keV emitted from Xe-133 and the γ-ray of 249.7 keV emitted from Xe-135. Since the count rates of the respective energy regions of Xe-133 and Xe-135 are discriminated and calculated, both radionuclides of Xe-133 and Xe-135 may be quantitatively measured. Furthermore, in the NaI scintillation detector, it is desirable that the periphery of the NaI scintillation detector be maintained at a constant temperature from the viewpoint of preventing noise. When the resolution of the γ-ray detector 31 is increased and a radioactive noble gas other than both radionuclides of Xe-133 and Xe-135 is detected, the Ge-semiconductor detector is desirable as the γ-ray detector 31 compared to the NaI scintillation detector. However, since the Ge-semiconductor detector has about 10% of the γ-ray sensitivity of the NaI scintillation detector, it is desirable to increase the volume of the gas G as the subject measurement medium when using the Ge-semiconductor detector as the γ-ray detector 31. The sample loop 32 is a sample container which retains the gas G as a subject measurement medium therein. The sample loop 32 is formed in a tubular shape, and may retain the gas G as the subject measurement medium inside the hollow portion. Since the sample loop 32 is formed in a hollow tubular shape, the gas G may be easily purged. For this reason, the replacement may be easily performed by supplying a new subject measurement medium into the sample loop 32. The sample loop 32 is disposed so as to surround the circumference of the γ-ray detector 31. For example, as illustrated in FIG. 2, the sample loop 32 is wound on the circumference of the γ-ray detector 31 in a spiral shape, and surrounds the circumference of the γ-ray detector 31. The gas G as the subject measurement medium surrounds the circumference of the γ-ray detector 31 which is held by the sample loop 32. The internal volume of the sample loop 32 and the number of times in which the sample loop is wound on the circumference of the γ-ray detector 31 in a spiral shape are set so as to match the concentration of the radionuclide included in the gas G as the subject measurement medium. For example, when the γ-ray detector 31 is the NaI scintillation detector, the internal volume of the sample loop 32 is about 20 ml. Further, when the sample loop 32 is degraded or the internal volume is changed, the sample loop 32 needs to be changed. For this reason, it is desirable that the sample loop 32 be separated from the circumference of the γ-ray detector 31 for the replacement. The lead shield 33 is a lead plate. The lead shield 33 is thickened as much as possible and is disposed so as to surround at least the γ-ray detector 31 and the sample loop 32. The lead shield 33 prevents the γ-ray from reaching the γ-ray detector 31 from the outside of the radioactive noble gas detecting device 30. The gas passageway 24A, the gas passageway 24C, and the gas passageway 46 are formed in a tubular shape, so that the gas G may pass therethrough. The opening and closing valve 41 is connected to the gas passageway 24C, and may adjust the amount of the gas G which is supplied from the gas passageway 24C to the radioactive noble gas detecting device 30. The opening and closing valve 42 is connected to the gas passageway 46, and may adjust the amount of the gas G which is discharged from the radioactive noble gas detecting device 30 to the gas passageway 46. The valve control units 43 and 44 control the opening and closing degrees of the opening and closing valves 41 and 42. As illustrated in FIG. 2, the gas G which is supplied from the gas phase sampling point 21 is supplied to the gas passageway 24A. The gas passageway 24A supplies the gas G to the branch point 23. The branch point 23 distributes the gas G to the gas passageway 24B and the gas passageway 24C. The radioactive noble gas detecting device 30 may receive the gas G supplied from the gas passageway 24C through the opening and closing valve 41. Further, when the opening and closing valves 41 and 42 are opened, the gas G supplied from the gas passageway 24C is extruded, so that the gas inside the radioactive noble gas detecting device 30 is discharged to the gas passageway 46 through the opening and closing valve 42. The exhaust process connection point 45 is connected to a facility that appropriately treats the gas G supplied from the gas passageway 46. The γ-ray detector 31 of the radioactive noble gas detecting device 30 is connected to the measurement control device 80. Then, the γ-ray detector 31 is electrically connected to the measurement control device 80 so as to transmit and receive the measurement data is or the instruction signal. The measurement control device 80 is electrically connected to the valve control unit 43 and the valve control unit 44 so that the instruction signal id may be transmitted to the valve control unit 43 which controls the opening and closing degree of the opening and closing valve 41 and the valve control unit 44 which controls the opening and closing degree of the opening and closing valve 42. Further, the gas G which is supplied from the gas passageway 24B moves to the analysis device 22. For example, the analysis device 22 detects the concentration and the like of hydrogen or oxygen included in the gas G. The nuclear reactor fuel integrity monitor 100 of the first embodiment commonly shares the gas phase sampling point 21 and the gas passageway 24A with the analysis device 22. As illustrated in FIG. 3, the measurement control device 80 includes an input processing circuit 81, an input port 82, a process unit 90, a storage unit 94, an output port 83, an output processing circuit 84, and a display device 85. The measurement control device further includes an input device 86 such as a keyboard if necessary. The process unit 90 includes, for example, a CPU (Central Processing Unit) 91, a RAM 92, and a ROM 93. The process unit 90, the storage unit 94, the input port 82, and the output port 83 are connected to one another through a bus 87, a bus 88, and a bus 89. By the bus 87, the bus 88, and the bus 89, the CPU 91 of the process unit 90 may exchange control data with or transmit an instruction to any one of the storage unit 94, the input port 82, and the output port 83. The input processing circuit 81 is connected to the input port 82. The input processing circuit 81 is connected with the measurement data is from the γ-ray detector 31. Then, the measurement data is which is output from the γ-ray detector 31 of the radioactive noble gas detecting device 30 is converted into a signal which may be used by the process unit 90 through a noise filter or an A/D converter provided in the input processing circuit 81, and is transmitted to the process unit 90 through the input port 82. Accordingly, the process unit 90 may acquire information necessary for calculating the concentration of the radionuclide. The output port 83 is connected with the output processing circuit 84. The output processing circuit 84 is connected with the display device 85 or the external output terminal. The output processing circuit 84 includes a display device control circuit, a valve control signal circuit for the opening and closing valve, a signal amplifying circuit, and the like. The output processing circuit 84 outputs the concentration of the radionuclide calculated by the process unit 90 as a display signal for displaying the concentration on the display device 85 or outputs the concentration as the instruction signal id transmitted to the γ-ray detector 31 and the valve control units 43 and 44. As the display device 85, for example, a liquid crystal display panel, a CRT (Cathode Ray Tube), or the like may be used. The storage unit 94 stores a computer program including a procedure of monitoring the nuclear reactor fuel state, a database of radionuclide concentration measurement data, and the like. Here, the storage unit 94 may include a volatile memory such as a RAM (Random Access Memory), a non-volatile memory such as a flash memory, and a hard disk or the combination of these. The computer program may execute a procedure of monitoring the nuclear reactor fuel state by the combination with the computer program which is previously stored in the process unit 90. Further, the measurement control device 80 may execute the procedure of monitoring the nuclear reactor fuel state by using exclusive hardware instead of the computer program. Further, the procedure of monitoring the nuclear reactor fuel state may be realized by executing the prepared program through a computer system such as a personal computer, a workstation, or a plant control computer. Further, the program is stored in a storage device such as a hard disk or a storage medium such as a flexible disk (FD), a ROM, a CD-ROM, a MO, a DVD, and a flash memory which may be read by the computer, and may be executed by reading out the program from the storage medium through the computer. Furthermore, the “computer system” mentioned herein includes hardware such as an OS or a peripheral device. Further, the “storage medium which may be read out by the computer” includes the case of dynamically storing the program for a short time as in a communication line used when transmitting the program through a network such as an internet or a communication network such as a phone line or the case of storing the program for a predetermined time as in a volatile memory inside a computer system which becomes a server or a client in that case. Further, the program may realize a part of the above-described function and further realize the above-described function by the combination with the program previously stored in the computer system. Next, the procedure of monitoring the nuclear reactor fuel state will be described by referring to FIGS. 2, 3, and 4. First, the CPU 91 of the process unit 90 included in the measurement control device 80 illustrated in FIGS. 2 and 3 receives the measurement request input from the input device 86 through the input processing circuit 81 and the input port 82, and temporarily stores the measurement request in the RAM 92 or the storage unit 94 (step S101). Alternatively, the CPU 91 stores the measurement request which is repeated every predetermined time in the RAM 92 or the storage unit 94 in advance. The CPU 91 generates a valve control instruction signal to be transmitted to the valve control units 43 and 44 which control the opening and closing degrees of the opening and closing valves 41 and 42 by using the measurement request as a trigger. Next, the CPU 91 outputs the valve opening instruction signal from the output signal processing circuit 84 to the valve control units 43 and 44 through the output port 83. The valve control units 43 and 44 which receive the valve opening instruction signal open the opening and closing valves 41 and 42 so as to supply the gas G as the subject measurement medium from the gas passageway 24C into the sample loop 32 and the gas G remaining inside the sample loop 32 is extruded by the supplied gas G so as to be entirely discharged. Furthermore, the exhaust may be performed by using a three-way valve in the opening and closing valve 41. Next, the valve control unit 44 closes the opening and closing valve 42 based on the valve closing instruction signal of the CPU 91. Next, the valve control unit 43 closes the opening and closing valve 41 based on the valve closing instruction signal of the CPU 91 so as to seal the gas G inside the sample loop 32. Then, since the gas G which remains inside the sample loop 32 is extruded by the gas G supplied from the gas passageway 24C so as to be entirely discharged, the gas which remains inside the sample loop 32 changes to the gas G of the new subject measurement medium (step S102). Next, the CPU 91 outputs the instruction signal id from the output signal processing circuit 84 to the radioactive noble gas detecting device 30 through the output port 83. The γ-ray detector 31 of the radioactive noble gas detecting device 30 which receives the instruction signal id starts the measurement. When the γ-ray detector 31 is, for example, the NaI scintillation detector, the measurement time of the γ-ray detector 31 is from 600 seconds to 1000 seconds. Then, the measurement data is of the γ-ray detector 31 is input to the measurement control device 80 (step S103). The measurement control device 80 discriminates and calculates the count rates of the respective energy regions of Xe-133 and Xe-135 from the input measurement data is of the γ-ray detector 31. The CPU 91 calculates the concentrations of the respective radionuclides of Xe-133 and Xe-135 from the data of the internal volume of the sample loop 32 stored in advance in the RAM 92 or the storage unit 94 and the count rates of the respective energy regions of Xe-133 and Xe-135. Specifically, the values which are obtained by dividing the number of signals per each unit time in the respective energy regions of Xe-133 and Xe-135 by the internal volume of the sample loop 32 become the concentrations of the respective radionuclides. The CPU 91 stores the concentrations of the respective radionuclides of Xe-133 and Xe-135 in the RAM 92 or the storage unit 94. The CPU 91 checks if there is measurement data of the precedent concentrations of the respective radionuclides of Xe-133 and Xe-135 stored in the RAM 92 or the storage unit 94. When there are the precedent concentrations of the respective radionuclides of Xe-133 and Xe-135, the CPU 91 compares the precedent concentrations of the respective radionuclides of Xe-133 and Xe-135 with the current concentrations of the respective radionuclides of Xe-133 and Xe-135. When the measurement data of the precedent concentrations of the respective radionuclides of Xe-133 and Xe-135 is not stored in the RAM 92 or the storage unit 94, the procedure of monitoring the nuclear reactor fuel state returns to step S101. When the concentration change rates of the respective radionuclides of Xe-133 and Xe-135 are within, for example, 50%/week, the procedure of monitoring the nuclear reactor fuel state returns to step S101 as a result in which the concentrations of the respective radionuclides of Xe-133 and Xe-135 are not changed (No in step S104). When the concentration change rates of the respective radionuclides of Xe-133 and Xe-135 exceed, for example, 50%/week, the procedure of monitoring the nuclear reactor fuel state proceeds to step S105 as a result in which the concentrations of the respective radionuclides of Xe-133 and Xe-135 are changed (Yes in step S104). Then, the CPU 91 outputs an alarm display on the display device 85 (step S105). Subsequently, when the repeated measurement is needed, the CPU 91 disaffirms the measurement end determination (No in step S106), and the procedure returns to the measurement request step S101 so as to continue the measurement. When the repeated measurement is not needed, the CPU 91 affirms the measurement end determination so as to end the measurement (Yes in step S106). The nuclear reactor fuel integrity monitor 100 of the first embodiment includes the γ-ray detector 31 which detects the γ-ray of the specific radionuclide of the subject measurement medium of the nuclear reactor, the sample loop 32 which is the sample container formed in a shape of causing the subject measurement medium to surround the γ-ray detector 31, and the measurement control device 80 which controls the opening and closing valves 41 and 42 so that a predetermined amount of the gas G as the subject measurement medium is introduced into the sample loop 32 and calculates the concentration of the specific radionuclide from the γ-ray data per each unit time detected by the γ-ray detector 31 and the volume of the gas G as the subject measurement medium introduced into the sample loop 32. Accordingly, the dissolved gas included in the nuclear reactor coolant C1 of the nuclear reactor is periodically sampled as the gas G from the gas phase sampling point 21 without the operator's labor, so that the concentration of the specific radionuclide included in the gas G may be measured. Since the sample loop 32 is formed in a shape in which the gas G as the subject measurement medium surrounds the γ-ray detector 31, it is possible to increase the volume of the subject measurement medium which may be detected by the γ-ray detector 31. Since the amount of the γ-ray which may be detected by the γ-ray detector 31 increases, there is no need to perform an operation which concentrates the subject measurement medium so as to improve the detection efficiency of the γ-ray detector 31. Further, since the opening and closing valves 41 and 42 are controlled so that a predetermined amount of the subject measurement medium is introduced into the sample container, it is possible to periodically measure the γ-ray of the subject measurement medium which is introduced into the sample loop 32 without needing the operator's labor. Further, the nuclear reactor fuel integrity monitor 100 of the first embodiment may calculate the concentration for each specific radionuclide. Further, the sample loop 32 as the sample container is formed in a hollow shape, and is wound on the γ-ray detector 31 in a spiral shape, so that the gas G inside the sample loop 32 surrounds the γ-ray detector. When the sample loop 32 is formed in a hollow shape and the new subject measurement medium flows thereinto, the subject measurement medium may be replaced without leaving the old gas of the sample loop 32. For this reason, when the new gas G as the subject measurement medium enters the sample loop 32, the old gas is extruded so as to be replaced by the new gas. Since the subject measurement medium is the gas G and the specific radionuclide is radioxenon, it is possible to generate an alarm at an early timing when radioxenon is discharged from the nuclear reactor fuel 2C. The nuclear reactor fuel integrity monitor 100 of the first embodiment may continuously measure the concentration of the specific radionuclide included in the gas G every predetermined time when periodically performing the measurement request on the radioactive noble gas detecting device 30. FIG. 5 is a schematic diagram illustrating an example of a nuclear reactor fuel integrity monitor according to a second embodiment. Furthermore, the same reference signs will be given to the same constituents of the above-described embodiment, and the description thereof will not be repeated. As illustrated in FIG. 5, a nuclear reactor fuel integrity monitor 200 includes a radioactive iodine detecting device 50, the measurement control device 80, opening and closing valves 71 and 72, valve control units 73 and 74, an activated alumina column 61 as a pre-treatment facility, and a degassing device 62. The radioactive iodine detecting device 50 is disposed between the nuclear reactor coolant sampling point 25 and a drainage treatment connection point 75, and is connected to the nuclear reactor coolant sampling point 25 and the drainage treatment connection point 75 through a nuclear reactor coolant passageway 26A, the activated alumina column 61, a nuclear reactor coolant passageway 26B, the degassing device 62, a nuclear reactor coolant passageway 26C, and a nuclear reactor coolant passageway 76. The radioactive iodine detecting device 50 includes at least a γ-ray detector 51, a sample container 52, and a lead shield 53. Further, the radioactive iodine detecting device 50 includes an electronic cooling device 54. Furthermore, the electronic cooling device 54 may be provided if necessary, and is not essentially needed. The nuclear reactor coolant C1 may include radioactive iodine (I-131, I-132, I-133, I-134, I-135), radioactive fluorine (F-18), and radioactive argon (Ar-41) dissolved therein as the radionuclide. In order to monitor the nuclear reactor fuel state, it is important to monitor the concentration of radioactive iodine (I-131, I-133, I-135) in radioactive iodine (I-131, I-132, I-133, I-134, I-135). The radioactive iodine detecting device 50 may detect radioactive iodine (I-131, I-132, I-133, I-134, I-135). As the γ-ray detector 51, for example, the NaI scintillation detector or the Ge-semiconductor detector is used. Since the resolution is more important than the detection efficiency or the maintenance workability, it is desirable to use the Ge-semiconductor detector as the γ-ray detector 51 compared to the NaI scintillation detector. As illustrated in FIG. 5, the sample container 52 includes a sample container recess portion 52a into which the γ-ray detector 51 is inserted. The nuclear reactor coolant C1 as the subject measurement medium which is retained by the sample container 52 surrounds the circumference of the γ-ray detector 51. Since the nuclear reactor coolant C1 as the subject measurement medium surrounds the circumference of the γ-ray detector 51, the amount of the γ-ray from the nuclear reactor coolant C1 as the subject measurement medium to the γ-ray detector 51 increases. For example, in an existing measurement method of concentrating and collecting radioactive iodine by an anion exchange filter so that the radioactive iodine adheres to the γ-ray detector 51, the nuclear reactor coolant C1 is used by 10 ml to 50 ml. The sample container 52 of the second embodiment retains the nuclear reactor coolant C1 at the volume of 100 ml to 200 ml. Then, since the subject measurement medium surrounds the circumference of the γ-ray detector 51, the γ-ray detector 51 may detect the γ-ray which is caused by the radioactive iodine from the nuclear reactor coolant C1 in the sample container 52 substantially at the same sensitivity as that of the existing measurement method of concentrating and collecting the radioactive iodine by the anion exchange filter so that the radioactive iodine adheres to the γ-ray detector 51. The sample container 52 includes a volume adjusting portion 52b which presses the liquid level of the nuclear reactor coolant C1 as the subject measurement medium upward. Since the volume adjusting portion 52b is provided, bubbles of an extra gas are eliminated from the circumference of the γ-ray detector 51. As a result, the volume of the nuclear reactor coolant C1 which exists in the circumference of the γ-ray detector 51 becomes constant. Further, the nuclear reactor fuel integrity monitor 200 prepares a plurality of sample containers 52 of which the volumes are changed in accordance with the radioactive concentration of the nuclear reactor coolant C1 as the subject measurement medium, so that the sample containers 52 having appropriate volumes may be attached to the circumference of the γ-ray detector 51. Since the sample containers 52 may be replaced, the inside of the sample containers 52 may be cleaned during the replacement. The lead shield 53 is a lead plate. The lead shield 53 is disposed so as to surround at least the γ-ray detector 51 and the sample container 52. It is desirable that the lead shield 53 be thickened as much as possible so as to prevent the γ-ray from reaching the γ-ray detector 51 from the outside of the radioactive iodine detecting device 50. When the γ-ray detector 51 is the Ge-semiconductor detector, the electronic cooling device 54 is connected to the Ge-semiconductor detector so as to reduce the current leaking to the Ge-semiconductor detector or the noise. The electronic cooling device 54 cools the Ge-semiconductor detector to, for example, the temperature substantially the same as that of the liquid nitrogen. The Ge-semiconductor detector may be cooled by using the liquid nitrogen instead of the electronic cooling device 54. As illustrated in FIG. 5, in the radioactive iodine detecting device 50, the nuclear reactor coolant C1 which is sampled from the nuclear reactor coolant sampling point 25 passes through the nuclear reactor coolant passageway 26A, the activated alumina column 61 as the pre-treatment facility, the nuclear reactor coolant passageway 26B, the degassing device 62 as the pre-treatment facility, the nuclear reactor coolant passageway 26C, and the opening and closing valve 71. The radioactive iodine detecting device 50 may discharge the nuclear reactor coolant C1 to the drainage treatment connection point 75 through the opening and closing valve 72 and the nuclear reactor coolant passageway 76. The drainage treatment connection point 75 is connected to the facility that may appropriately treats the nuclear reactor coolant C1 supplied from the nuclear reactor coolant passageway 76. Further, the γ-ray detector 51 of the radioactive iodine detecting device 50 is connected to the measurement control device 80. Then, the measurement control device 80 receives the measurement data is of the γ-ray detector 51. Further, the instruction signal id of the measurement control device 80 may be input to the γ-ray detector 51. The measurement control device 80 is connected to the valve control unit 73 and the valve control unit 74 so that the instruction signal id may be transmitted to the valve control unit 73 which controls the opening and closing degree of the opening and closing valve 71 and the valve control unit 74 which controls the opening and closing degree of the opening and closing valve 72. The measurement control device 80 is the same as that of the first embodiment. The activated alumina column 61 as the pre-treatment facility is a column which is filled with activated alumina, and the nuclear reactor coolant C1 which is injected from the nuclear reactor coolant sampling point 25 may pass therethrough. Radioactive fluorine (F-18) included in the nuclear reactor coolant C1 becomes the hindering nuclide in which radioactive iodine (I-131, I-132, I-133, I-134, I-135) is analyzed by the γ-ray detector 51. In general, the nuclear reactor coolant C1 passes through the anion exchange filter, and radioactive fluorine and radioactive iodine are concentrated and collected in the anion exchange filter. Then, since the radioactive fluorine is a short-half-life radionuclide, the collapsing of the radioactive fluorine is awaited and the radioactive iodine is analyzed by the γ-ray detector. The detection sensitivity of the γ-ray measurement is greatly affected by the positional relation between the γ-ray detector and the sample of the measurement subject. Since the radioactive iodine collected on the anion exchange filter may be measured while adhering to the γ-ray detector, the detection efficiency is high. However, the automation without the operator's labor is difficult from the viewpoint of the cost and the mechanism. In a general method, some time is needed for the collapsing of the radioactive fluorine and the concentration and the collection are manually performed. For this reason, it is difficult to periodically sample the nuclear reactor coolant C1 of the nuclear reactor from the nuclear reactor coolant sampling point 25 without the operator's labor and to measure the concentration of the specific radionuclide included in the sampled nuclear reactor coolant C1. In the activated alumina column 61 of the second embodiment, when the nuclear reactor coolant C1 passes therethrough, the radioactive fluorine is absorbed onto the activated alumina and the nuclear reactor coolant C1 from which the radioactive fluorine is eliminated is discharged to the degassing device 62. For this reason, in the nuclear reactor fuel integrity monitor 200 of the second embodiment, even when the nuclear reactor coolant C1 of the nuclear reactor is periodically sampled from the nuclear reactor coolant sampling point 25 without the operator's labor, the radioactive fluorine as the hindering nuclide is not supplied to the radioactive iodine detecting device 50. Furthermore, the activated alumina column 61 may be periodically replaced. Further, radioactive argon (Ar-41) which is included in the nuclear reactor coolant C1 becomes the hindering nuclide in which radioactive iodine (I-131, I-132, I-133, I-134, I-135) is analyzed by the γ-ray detector 51 as in the radioactive fluorine (F-18). The radioactive argon (Ar-41) is not eliminated from the activated alumina column 61. Therefore, the nuclear reactor coolant C1 from the activated alumina column 61 passes through the degassing device 62. The degassing device 62 eliminates the radioactive argon (Ar-41), and discharges the nuclear reactor coolant C1 from which the radioactive argon (Ar-41) is eliminated to the radioactive iodine detecting device 50. Further, the degassing device 62 eliminates the gas which is dissolved in the nuclear reactor coolant C1 other than the radioactive argon (Ar-41). In the second embodiment, the nuclear reactor coolant C1 is pre-treated by the activated alumina column 61 and the degassing device 62 in this order, so that the radioactive fluorine (F-18) and the radioactive argon (Ar-41) are eliminated. Furthermore, the sequence of arranging the activated alumina column 61 and the degassing device 62 may be reversed. Further, the nuclear reactor coolant C1 includes radioactive sodium (Na-24), and in the case of the hindering nuclide in which the radioactive iodine is analyzed by the γ-ray detector 51, a cation exchange column may be further provided as the pre-treatment. Next, the procedure of monitoring the nuclear reactor fuel state will be described by referring to FIGS. 3, 4, and 5. First, the CPU 91 of the process unit 90 which is included in the measurement control device 80 illustrated in FIGS. 3 and 5 receives the measurement request input from the input device 86 through the input processing circuit 81 and the input port 82, and temporarily stores the measurement request in the RAM 92 or the storage unit 94 (step S101). Alternatively, the CPU 91 stores the measurement request which is repeated every predetermined time in the RAM 92 or the storage unit 94 in advance. The CPU 91 converts the measurement request into the instruction signal id to be transmitted to the valve control units 73 and 74 which control the opening and closing degrees of the opening and closing valves 71 and 72 (step S101). Next, the CPU 91 outputs the valve opening instruction signal from the output signal processing circuit 84 to the valve control units 73 and 74 through the output port 83. The valve control units 73 and 74 which receive the valve opening instruction signal open the opening and closing valves 71 and 72 so as to supply the nuclear reactor coolant C1 as the subject measurement medium from the nuclear reactor coolant passageway 26C into the sample container 52 and to discharge the nuclear reactor coolant C1 remaining inside the sample container 52 by the supplied nuclear reactor coolant C1. Next, the valve control unit 74 closes the opening and closing valve 72 based on the valve closing instruction signal of the CPU 91. Next, the valve control unit 73 closes the opening and closing valve 71 based on the valve closing instruction signal of the CPU 91, and seals the nuclear reactor coolant C1 inside the sample container 52. Then, the nuclear reactor coolant C1 remaining inside the sample container 52 is replaced by the nuclear reactor coolant C1 of the new subject measurement medium by the nuclear reactor coolant C1 which is supplied from the nuclear reactor coolant passageway 26C (step S102). Since there is a possibility that the nuclear reactor coolant C1 may include a short-half-life radionuclide, it is desirable that the decay time be, for example, about 80 minutes. Next, the CPU 91 outputs the instruction signal id from the output signal processing circuit 84 to the radioactive iodine detecting device 50 through the output port 83. The γ-ray detector 51 of the radioactive iodine detecting device 50 which receives the instruction signal id starts the measurement. When the γ-ray detector 51 is, for example, the Ge-semiconductor detector, the measurement time of the γ-ray detector 51 is 15 minutes. Then, the measurement data is of the γ-ray detector 51 is input to the measurement control device 80 (step S103). The measurement control device 80 discriminates and calculates the count rates of the respective energy regions of I-131, I-132, I-133, I-134, and I-135 from the input measurement data is of the γ-ray detector 51. The CPU 91 calculates the concentrations of the respective radionuclides of I-131, I-132, I-133, I-134, and I-135 from the data of the internal volume of the sample container 52 stored in the RAM 92 or the storage unit 94 in advance and the count rates of the respective energy regions of I-131, I-132, I-133, I-134, and I-135. Specifically, the value which is obtained by dividing the number of signals per unit time in each of the energy regions of I-131, I-132, I-133, I-134, and I-135 by the internal volume of the sample container 52 becomes the concentration of each of radionuclides. The CPU 91 stores the concentrations of the respective radionuclides of I-131, I-132, I-133, I-134, and I-135 in the RAM 92 or the storage unit 94. The CPU 91 checks if there is measurement data of the precedent concentrations of the respective radionuclides of I-131, I-133, and I-135 stored in the RAM 92 or the storage unit 94. When there is the measurement data of the precedent concentrations of the respective radionuclides of I-131, I-133, and I-135, the CPU 91 compares the precedent concentrations of the respective radionuclides of I-131, I-133, and I-135 with the current concentrations of the respective radionuclides of I-131, I-133, and I-135. When the measurement data of the precedent concentrations of the respective radionuclides of I-131, I-133, and I-135 is not stored in the RAM 92 or the storage unit 94, the procedure of monitoring the nuclear reactor fuel state returns to step S101. When the concentration change rate of each radionuclide of radioactive iodine (I-131, I-133, I-135) is within, for example, 50%/week, the procedure of monitoring the nuclear reactor fuel state returns to step S101 as a result in which the concentration of each radionuclide of radioactive iodine (I-131, I-133, I-135) is not changed (No in step S104). When the concentration change rate of each radionuclide of radioactive iodine (I-131, I-133, I-135) exceeds, for example, 50%/week, the procedure of monitoring the nuclear reactor fuel state proceeds to the next step as a result in which the concentration of each radionuclide of radioactive iodine (I-131, I-133, I-135) is changed (Yes in step S104). Then, the CPU 91 outputs an alarm display on the display device 85 (step S105). Subsequently, when the repeated measurement is needed, the CPU 91 disaffirms the measurement end determination (No in step S106), and the procedure returns to the measurement request step S101 so as to continue the measurement. When the repeated measurement is not needed, the CPU 91 affirms the measurement end determination so as to end the measurement (Yes in step S106). The nuclear reactor fuel integrity monitor 200 of the second embodiment includes the γ-ray detector 51 which detects the γ-ray of the specific radionuclide of the subject measurement medium of the nuclear reactor, the sample container 52 which is formed in a shape of causing the subject measurement medium to surround the γ-ray detector 51, and the measurement control device 80 which controls the opening and closing valves 71 and 72 so that a predetermined amount of the nuclear reactor coolant C1 as the subject measurement medium is introduced into the sample container 52 and calculates the concentration of the specific radionuclide from the γ-ray data per unit time detected by the γ-ray detector 51 and the volume of the nuclear reactor coolant C1 as the subject measurement medium introduced into the sample container 52. Accordingly, the nuclear reactor coolant C1 of the nuclear reactor is periodically sampled from the nuclear reactor coolant sampling point 25 without the operator's labor, so that the concentration of the specific radionuclide included in the nuclear reactor coolant C1 may be measured. Since the sample container 52 is formed in a shape in which the γ-ray detector 51 is surrounded by the nuclear reactor coolant C1 as the subject measurement medium, it is possible to increase the volume of the subject measurement medium which may be detected by the γ-ray detector 51. Since the amount of the γ-ray which may be detected by the γ-ray detector 31 increases, there is no need to perform the operation of concentrating the subject measurement medium so as to improve the detection efficiency of the γ-ray detector 31. Further, since the opening and closing valves 71 and 72 are controlled so that a predetermined amount of the subject measurement medium is introduced into the sample container 52, it is possible to periodically measure the γ-ray of the subject measurement medium introduced into the sample container 52 without the operator's labor. Further, the nuclear reactor fuel integrity monitor 200 of the second embodiment may calculate the concentration for each specific radionuclide. Since the sample container 52 includes the sample container recess portion 52a and the γ-ray detector 51 is inserted into the sample container recess portion 52a, the nuclear reactor coolant C1 as the subject measurement medium surrounds the circumference of the γ-ray detector 51, and the amount of the γ-ray which is received by the γ-ray detector 51 from the nuclear reactor coolant C1 as the subject measurement medium increases. Further, when the radioactive iodine is discharged from the nuclear reactor fuel 2C, it is possible to generate an alarm at an early timing. In the nuclear reactor fuel integrity monitor 200 of the second embodiment, when the measurement request of the radioactive iodine detecting device 50 is periodically performed, the concentration of the specific radionuclide included in the nuclear reactor coolant C1 is continuously measured every predetermined time. Further, since the nuclear reactor coolant C1 as the subject measurement medium passes through the activated alumina column 61 so as to be introduced into the sample container 52, even when the nuclear reactor coolant C1 of the nuclear reactor is periodically sampled from the nuclear reactor coolant sampling point 25 without the operator's labor, the radioactive fluorine as the hindering nuclide is not supplied to the radioactive iodine detecting device 50. Then, the radioactive iodine is measured by the γ-ray detector 51 without the interference of the radioactive fluorine. Both the nuclear reactor fuel integrity monitor 100 of the first embodiment and the nuclear reactor fuel integrity monitor 200 of the second embodiment may be connected to the nuclear plant 1. In this case, the nuclear reactor fuel integrity monitor 100 includes a first γ-ray detector which detects a γ-ray of a first specific radionuclide and a first sample container, and the nuclear reactor fuel integrity monitor 200 includes a second γ-ray detector which detects a γ-ray of a second specific radionuclide and a second sample container. Further, the measurement control device 80 performs a control so that a predetermined amount of the gas is introduced into the first sample container and a predetermined amount of the coolant is introduced into the second sample container. Also, the measurement control device calculates the concentrations of the first and second specific radionuclides from the γ-ray data per unit time detected by the first and second γ-ray detectors and the volume of the subject measurement medium introduced into the first and second sample containers. Since both the nuclear reactor fuel integrity monitors 100 and 200 are provided, the nuclear reactor fuel state may be monitored based on the measurement results of both the gas G and the nuclear reactor coolant C1. 1 NUCLEAR PLANT 1W CONTAINMENT 2 NUCLEAR REACTOR 2C NUCLEAR REACTOR FUEL 3 STEAM GENERATOR 3T HEAT TRANSFER PIPE 4 PRESSURIZER 5 NUCLEAR REACTOR COOLANT PUMP 13A NUCLEAR REACTOR COOLANT EXTRACTING PASSAGEWAY 13B, 13C, 13D NUCLEAR REACTOR COOLANT PASSAGEWAY 13E, 13F NUCLEAR REACTOR COOLANT RETURNING PASSAGEWAY 14 VOLUME CONTROL TANK 14a GAS PHASE PORTION 15 CHARGING PUMP 16 DESALINATION TOWER 21 GAS PHASE SAMPLING POINT 22 ANALYSIS DEVICE 23 BRANCH POINT 24A, 24B, 24C GAS PASSAGEWAY 25 NUCLEAR REACTOR COOLANT SAMPLING POINT 26A, 26B, 26C NUCLEAR REACTOR COOLANT PASSAGEWAY 30 RADIOACTIVE NOBLE GAS DETECTING DEVICE 31 γ-RAY DETECTOR 32 SAMPLE LOOP 33 LEAD SHIELD 41, 42 OPENING AND CLOSING VALVE 43, 44 VALVE CONTROL UNIT 45 EXHAUST PROCESS CONNECTION POINT 46 GAS PASSAGEWAY 50 RADIOACTIVE IODINE DETECTING DEVICE 51 γ-RAY DETECTOR 52 SAMPLE CONTAINER 52a SAMPLE CONTAINER RECESS PORTION 52b VOLUME ADJUSTING PORTION 53 LEAD SHIELD 54 ELECTRONIC COOLING DEVICE 61 ACTIVATED ALUMINA COLUMN 62 DEGASSING DEVICE 71, 72 OPENING AND CLOSING VALVE 73, 74 VALVE CONTROL UNIT 75 DRAINAGE TREATMENT CONNECTION POINT 76 NUCLEAR REACTOR COOLANT PASSAGEWAY 80 MEASUREMENT CONTROL DEVICE 90 PROCESS UNIT 91 CPU 92 RAM 93 ROM 94 STORAGE UNIT 100, 200 NUCLEAR REACTOR FUEL INTEGRITY MONITOR |
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039792560 | description | DETAILED DESCRIPTION OF THE INVENTION As shown in FIG. 1, a pair of neutron detectors 11 and 13 are positioned in reactor 10 to measure the neutron flux therein. The outputs from detectors 11 and 13 are coupled to failure detection circuit 15 which acts to detect abnormal values of neutron flux in the reactor and develop control signals in response to abnormal values of neutron flux. The control signals are coupled to other circuits and mechanisms (for example, control rod actuator 18) to shut down or otherwise change the operation of the reactor. The failure detection circuit 15 may include circuit redundance, voting schemes, automatic self-checking and other features to minimize the chance of reactor shutdown because of a failure in the monitoring circuitry when the reactor was in a safe operating condition. None of the safety systems which have previously been used have provided for satisfactory continuous monitoring of the safety instrumentation over the entire operating range of the reactor, including startup, intermediate power or full power. In the continuous reactor monitoring system of this invention it is assumed that the entire reactor behaves as a unit; that is, the signals from radiation detectors (or other instruments) located at different positions in the reactor will track to a satisfactory degree over the operating range of the reactor. Thus if a ratio is taken between any pair of radiation detector outputs, this ratio will remain substantially constant over the operating range of the reactor. Referring again to FIG. 1, the outputs from neutron detectors 11 and 13 are coupled to the division circuit 17 through amplifier circuits 14 and 16. Division circuit 17 develops an output signal which is proportional to the ratio of the signals from detectors 11 and 13. This signal from division circuit 17 is compared to reference signals in comparator 19 to develop an alarm signal when the ratio signal is outside a given range. The alarm signal is coupled to logic and control circuit 20 where it can be utilized as desired. It can, for example, be used to actuate an alarm signal or circuit or provide a display indicating that a safety channel has malfunctioned. The continuous safety monitor of this invention, for example, could require that a signal proportional to the ratio i.sub.1 /i.sub.2 of the output currents of two neutron detectors be within a narrow range of values over the operating range of a reactor. Signals proportional to log i.sub.1 and log i.sub.2 will already be available since logarithmic safety channels are in existence on most reactors. A signal proportional to log (i.sub.1 /i.sub.2) is easily obtained by taking the difference between two logarithmic channels. The logarithm of the ratio serves as well as the ratio since, for small variations of the ratio about unity: EQU log (i.sub.1 /i.sub.2) .apprxeq. K.sub.10 (i.sub.1 /i.sub.2 - 1) where K.sub.10 is the inverse of the natural logarithm of 10 needed for conversion to common logarithms which are convenient in the operation of the continuous safety monitor. In FIG. 2 there is shown a continuous safety monitor using the logarithmic safety channels of a reactor. A pair of neutron flux detectors 22 and 24 are positioned within the reactor 21 and measure the flux of the reactor at different points. The output signal i.sub.2 from detector 22 is coupled to a log amplifier 25 and the output signal i.sub.1 from detector 24 is coupled to a log amplifier 27. The output signals from log amplifiers 27 and 25 are: EQU V.sub.1 ' = -(V.sub.d1 ' log i.sub.1 + V.sub.10 ') (1) EQU v.sub.2 ' = (v.sub.d2 ' log i.sub.2 + V.sub.20 ') (2) where V.sub.d1 ' and V.sub.d2 ' are the average volts per decade, constants of amplifiers 27 and 25 and V.sub.10 ' and V.sub.20 ' are the dc offset voltages of the amplifiers. Signal V.sub.2 ' is combined with a reference voltage from reference voltage source 29 and the combined signal is coupled to offset adjustment amplifier 30. The output signal from amplifier 30 is: EQU V.sub.2 = (V.sub.d2 log i.sub.2 + V.sub.20) (3) the output signal V.sub.1 ' from amplifier 27 is coupled to buffer amplifier 32 where it is amplified to develop the signal: EQU V.sub.1 = -(V.sub.d log i.sub.1 + V.sub.10) (4) the polarities of the amplifiers 25, 27, 30 and 32 are such that the output signals have the desired polarities. The gain of amplifier 30 is set by adjusting the impedance of the feedback loop, represented by variable resistor 28, so that the volts per decade constant V.sub.d2 is substantially equal to V.sub.d. Thus equation (3) becomes EQU V.sub.2 = (V.sub.d log i.sub.2 + V.sub.20) (5) signals V.sub.1 and V.sub.2 are combined and amplified in summing amplifier 33 to develop an output signal V.sub.3 : EQU v.sub.3 = -g(v.sub.1 + v.sub.2) (6) where G is the gain of amplifier 33 and is established by the variable impedance 34 in the feedback loop of amplifier 33. From equations (3) and (5): EQU V.sub.3 = -G(-V.sub.d log i.sub.1 - V.sub.10 + V.sub.d log i.sub.2 + V.sub.20) (7) EQU v.sub.3 = g(v.sub.d log i.sub.1 /i.sub.2) + G(V.sub.10 - V.sub.20) (8) EQU v.sub.10 - v.sub.20 = d (9) EQU v.sub.3 = gv.sub.d log i.sub.1 /i.sub.2 + GD (10) thus the output signal V.sub.3 coupled to comparator 35 is proportional to log (i.sub.1 /i.sub.2) plus an offset signal GD. Signal V.sub.3 is compared in comparator 35 with an upper limit voltage and a lower limit voltage and an alarm signal is developed if the voltage V.sub.3 falls outside of this range of voltage. The voltage range can be adjusted as desired to provide the desired monitoring and safety. The alarm signal from comparator 35 is coupled to the logic and control circuit 36 where it can be used to sound an alarm or to actuate a display as desired. The offset voltage GD is used to make the system failsafe against any failure which causes V.sub.3 to go to zero. For example, if i.sub.1 and i.sub.2 were substantially equal, log (i.sub.1 /i.sub.2) would be zero and comparator 35 would be set to monitor voltages centered about zero. A short circuit or other failure at the output of amplifier 33 which would develop a zero output signal would indicate that the safety channels were performing satisfactorily when, in fact, there might have been a failure in one of the channels. By proper selection of the offset voltage a failure in the continuous safety monitoring system which causes V.sub.3 to go to zero would be detected. The offset voltage GD is also chosen so that malfunction which causes amplifier 33 to go to either positive or negative saturation would also be detected. GD is set by adjusting the magnitude of the reference voltage from the reference voltage source 29 and by adjusting the gain of amplifier 33. An example of the values which were used in a prototype circuit GD was set at +5V. The upper and lower voltage levels detected by comparator 35 were set at +8V and +2V, which gave a range of 6V centered about the offset voltage GD. Under ideal circumstances the ratio i.sub.1 /i.sub.2 would remain constant over a wide range of currents. However, in practice a certain amount of variation can be expected. For example, variation could be caused by instrument error or neutron flux differences at different locations of the neutron flux detectors. Thus, V.sub.3 must be outside of a particular voltage range before an alarm is given. However, this voltage range is constant over a wide reactor operating range while in the case of a system which used the algebraic difference between the signals from the detectors the voltage range would have to increase at large reactor power levels to values which would make the safety monitoring useless. Thus the continuous safety monitoring system can operate continuously over a wide range of reactor operation while prior art systems cannot operate either continuously or over the reactor operating range or both. While the continuous reactor monitoring system of this invention has been described in conjunction with the measurement of neutron flux, other reactor parameters, as for example coolant temperature, could be measured. |
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