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046876006 | claims | 1. Treating process for the separation of coated nuclear fuel particles from a graphitic matrix in which the nuclear fuel particles are embedded; comprising removing the graphite encasing the nuclear fuel particles through comminuting of said graphite by a brush, said brush and said graphite being moved towards each other at a pressure so as to comminute said graphite while concurrently isolating the nuclear fuel particles with their coating or the nuclear fuel particles themselves in an undestroyed, intact condition; and separating the nuclear fuel particles contained in the brushed material from the comminuted graphite. 2. Treating process as claimed in claim 1, comprising exposing said separated nuclear fuel particles in a crushing means. 3. Treating process as claimed in claim 2, said crushing means comprising a jaw crusher. 4. Treating process as claimed in claim 2 or 3, said crushing means comprising a roll crusher. 5. Treating process as claimed in claim 2, comprising conducting graphite residues containing nuclear fuel particles which have not been conveyed away during the brushing step to said crushing means. 6. Treating process as claimed in claim 2, comprising chemically dissolving the nuclear fuel particles comminuted in said crushing means subsequent to sifting of the ground material and separation of graphite. 7. Treating process as claimed in claim 1, comprising conditioning and storing the separated graphite. |
description | This application is a continuation of and claims priority to U.S. patent application Ser. No. 10/767,723 entitled METHOD AND SYSTEM FOR AUTOMATICALLY SCANNING AND IMAGING THE CONTENTS OF A MOVING TARGET filed Jan. 30, 2004 now U.S. Pat. No. 7,039,159, which is incorporated herein by reference in its entirety. 1. Field of the Invention The invention relates generally to the field of imaging a target and more particularly to the field of imaging the contents of a moving target. 2. Description of the Related Art In this time of increased security concerns, authorities are continually looking for ways to improve national security through imaging technology. Additionally, law enforcement continues to battle drug, stolen goods and people trafficking both at the borders and within the borders of the United States through nonintrusive x-ray and gamma-ray imaging. There is a balance that must be struck between the desire to check the contents of vehicles for illegal and/or potentially hazardous materials and the desire to protect the drivers of the vehicles and to minimize the impact of the investigation on the flow of commerce. Currently available vehicle and cargo imaging systems, particularly those directed towards the imaging of moving targets, i.e., trucks, etc., utilize stop-and-go procedures that require manual control of the scan process. For example, current systems require the driver of the moving target to (1) stop the vehicle in a scanning zone, (2) certain systems require the driver to exit the vehicle and go a safe distance from the scanning zone to avoid potential exposure to the imaging radiation and, (3) certain systems require the driver, or some third party, to manually initiate the scanning of the vehicle. Further, many of the systems that are currently available for such imaging, utilize a high power x-ray source for the imaging radiation. By way of specific example, a particular known vehicle and cargo imaging system and process is described as follows. The driver approaches the first of 3 traffic signals. The first signal “enter” is green when there is no vehicle sensed between opposing source and detector towers defining the scan area, and red when there is a vehicle in the scan area. When the “enter” signal turns green the driver approaches a driver arm and the second traffic signal, which is red at this point. The “enter” of the first traffic signal also turns red, prohibiting any other vehicles from entering the scan area. Sensors detect the presence of the vehicle and a flashing yellow light on the driver arm engages. This prompts the driver to press the driver pushbutton located on a panel outside of the vehicle before the driver arm. This pushbutton sends a signal to an operator console notifying the operator that the driver is ready for his vehicle to be scanned. The operator presses the blinking “scan” button on the operator console and the shutters to the scanning source are opened. At this time, the second “scan” traffic light turns green and the driver proceeds through the scan area. As the vehicle exits the scan area, sensors detect the lack of a vehicle and automatically close the shutters to the scanning source. At this point, the first “enter” traffic signal turns green for another vehicle to proceed to the driver arm. Lastly, the “Exit” button on the operator panel lights up and the operator can depress the button to change the third traffic signal from red to green, thus allowing the vehicle to completely exit the area. This whole process takes approximately 20 seconds for a nominal scan. Summary of the Problem Referring to the “Description of the Related Art,” there is a need for a system and method to automatically scan and image moving vehicles in an optimally efficient and unobtrusive manner so as to minimize the effect on the flow of commerce and protect the drivers and third parties from exposure to the scanning radiation. Summary of the Solution The present invention describes a system and method for automatically scanning a target vehicle according to at least the following embodiments. According to a first embodiment of the present invention, described herein is an automated target inspection system for inspecting a moving target. The system includes a scanning zone that comprises a radiation source and a radiation source detector. The system further includes a first sensor component for automatically sensing when a first portion of the moving target has passed through the scanning zone and a second portion of the moving target is about to enter the scanning zone, wherein the first sensor component sends a signal to the automated target inspection system to initiate a scan of the second portion upon sensing that the second portion of the target is about to enter the scanning zone. Additionally, a shutter, triggered by a signal from the first sensor component, allows radiation from the radiation source to pass through the scanning zone in the direction of the radiation detector when the second portion of the moving target is passing through the scanning zone and closes off the radiation when the second portion of the moving target is no longer within the scanning zone. According to a second embodiment of the present invention, described herein is a method for automatically inspecting a moving target with an automated target inspection system. The method includes (1) sensing when a first portion of the moving target has passed through a scanning zone and a second portion of the moving target is about to enter the scanning zone; (2) sending a signal to the automated target inspection system to initiate a scan of the second portion upon sensing that the second portion of the target is about to enter the scanning zone; (3) opening a shutter to allowing radiation from a radiation source to pass through the scanning zone in the direction of a radiation detector when the second portion of the moving target is passing through the scanning zone; and (4) closing the shutter to shut off the radiation when the second portion of the moving target is no longer within the scanning zone. According to a third embodiment of the present invention, described herein is a system for automatically inspecting a moving target. The system includes means for sensing when a first portion of the moving target has passed through a scanning zone and a second portion of the moving target is about to enter the scanning zone; means for sending a signal to the automated target inspection system to initiate a scan of the second portion upon sensing that the second portion of the target is about to enter the scanning zone; means for opening a shutter to allowing radiation from a radiation source to pass through the scanning zone in the direction of a radiation detector when the second portion of the moving target is passing through the scanning zone; and means for closing the shutter to shut off the radiation when the second portion of the moving target is no longer within the scanning zone. In addition to the description set forth explicitly below, numerous details and descriptions for various aspects of the preferred embodiments are set forth in the following United States patents and patent applications which are incorporated herein by reference in their entireties: U.S. Pat. No. 6,255,654 for DENSITY DETECTION USING DISCRETE PHOTON COUNTING; U.S. Pat. No. 6,507,025 for DENSITY DETECTION USING REAL TIME DISCRETE PHOTON COUNTING FOR FAST MOVING TARGETS; U.S. Pat. No. 6,552,346 for DENSITY DETECTION USING DISCRETE PHOTON COUNTING; U.S. patent application Ser. No. 09/925,009, for DENSITY DETECTION USING REAL TIME DISCRETE PHOTON COUNTING FOR FAST MOVING TARGETS, filed Aug. 9, 2001; and U.S. patent application Ser. No. 10/717,632 for DENSITY DETECTION USING REAL TIME DISCRETE PHOTON COUNTING FOR FAST MOVING TARGETS, filed Nov. 21, 2003. Referring to FIGS. 1a and 1b, a preferred embodiment of the present invention provides a non-stop drive-through scanning system 10 for imaging the contents of moving target vehicles, e.g., 15a, 15b, 15c, etc. (referred to herein individually as 15). The preferred embodiments of the present invention facilitate back-to-back scanning of moving vehicles 15a, 15b, 15c, etc., without the need to stop the vehicle and initiate scanning manually, thus facilitating increased rate of the flow of commerce. The system of FIGS. 1a and 1b allows the driver of a vehicle, e.g., van, truck, train, etc. to enter the scanning zone 20 without the need to stop or exit from the target vehicle 15. The scanning zone 20 is defined by the space between opposing source and detector towers 25, 30. Details regarding various embodiments of the source and detector towers are described in the patents and applications listed above which are incorporated herein by reference. By way of example, the source tower 25 may include a radiation source such as a 3.7×1010 Bq shuttered source of Cs-137 gamma-rays, i.e., 662 keV gamma-ray energy. In an alternative embodiment, a Co-60 source may be used. A suitable source is readily available as Model Nos. SH-F-2 and SH-F-3 from Ohmart Corporation of Ohio. The radiation source may include a collimator that provides desired vertical and lateral opening specifications. The radiation source provides gamma-rays that are only moderately attenuated by steel walls typically found in tanker trucks or railroad cars. Yet such rays are sufficiently attenuated by contraband packages to make them easily detectable by measuring the penetration of the gamma-rays emitted from the source and deriving relative material densities therefrom. In addition, there is negligible scattering of the gamma-ray energy from the tanker walls or cargo, much less than would occur if a high-powered x-ray source was utilized. An exemplary detector tower 30 includes a detector array that employs a plurality of high efficiency gamma-ray detectors, e.g., between twenty and sixty, e.g., forty-eight, detectors arranged in a vertical column. The detectors make it possible to scan the target vehicle with a very low intensity gamma-ray field. In order to facilitate the use of very low intensity gamma-radiation, high efficiency detectors are used, such as are available as Part No. 1.5M1.5M1.5, NaI (TI) (sodium iodide crystal, thallium activated) (with R2060 photomultiplier tube) from BICRON of Ohio. Such gamma-ray detectors are scintillation counter-type detectors and are 3.8 cm in diameter, 3.8 cm high and mounted on a 3.8 cm photo-multiplier tube (PMT). Alternatively, 1.125″ (2.858 cm) square detectors may be used with the number of detectors used in the detector configuration ranging between 40 and 180, depending on desired resolution requirements. Referring to FIG. 2, in a specific embodiment, the detector array includes a plurality of staggered NaI/PMT square photon detector elements 100. Each individual detector is 1.125″ (2.858 cm) square and has a pitch P smaller than the diameter (d) of the staggered detector elements 100. Two (2) vertical rows R of staggered detector elements 100 are employed, instead of a single row of detectors. The two (2) vertical rows R are vertically staggered from each other. The pitch P between two (2) closest adjacent such staggered detector elements 100 may preferably be about 0.7″, when employing staggered detector elements 100 having a 1.125″ (2.858 cm) diameter, thereby yielding a count rate of about 95,000 counts/second for each staggered detector element for D=16 feet and for a 1.0 Curie Cs-137 source. By way of example, this pitch P results in a vertical resolution, Rvert or vertical grid unit of about 0.23″ when the radiation source is a distance D of 16 feet from the staggered detector element and the radiation source is a distance z of 8 feet (2.4 meters) from a center of the moving target vehicle wherein Rvert=PZ/D. The staggered detectors are staggered from each other in a vertical direction, yet their surfaces of each vertical row all lie in a same plane, thereby avoiding shadowing from any other staggered detector while enabling a smaller pitch P. Referring to FIGS. 3a and 3b, at least one start/stop sensor 35 is located prior to the scanning zone and is used to determine when the cab or driver/passenger area 40 (hereafter “cab”) has cleared the scanning zone and the payload 45 of the target vehicle 15 is entering the scanning zone. Once the start/stop sensor 35 senses the payload 45 has entered, or is about to enter, the scanning zone, the non-stop system initiates an automatic scan. The start/stop sensor 35 may include at least one of optical, electrical, pressure, video technology or the like for determining the start/stop points and automatic scan initiations as described above. More particularly, and by way of example, sensor technologies may be employed to count axles and/or measure the space between the cab 40 and the payload 45. In FIG. 3a, the at least one start/stop sensor 35 is located prior to the scanning zone and is above the level of the driving surface. In an alternative embodiment shown in FIG. 3b, the start/stop sensor 35 is located within or very near to the driving surface. In this embodiment, the sensor may be outside of the driving lane or actually in the driving lane such that the target vehicle is driven directly over the start/stop sensor 35. One skilled in the art can appreciate the various sensor configurations that would be considered to be within the scope of the present invention. Referring to FIGS. 4a–4c, in an embodiment of the present invention the automatic initiation of a scan includes, among other features, controlling a shutter assembly, including a fast shutter mechanism 200 located in the source tower that allows at least a first fast shutter to open and allow radiation from the radiation source to exit through beam aperture 210 in a sufficient time, e.g., on the order of a few tens of milliseconds, between detection of the cab and the payload by the start/stop sensor, so as to allow for a complete payload scan, i.e., including the beginning edge of the payload, while the target vehicle is moving, i.e., at speeds of up to 10 miles per hour, with normal operating range of between 5–10 mph. By way of particular example, the system and method of the present invention facilitates a fast shutter opening time on the order of 50 milliseconds or less, preferably 40 milliseconds, to allow the shutter to fully open after the cab passes the start/stop sensor when a target vehicle driving at 7 MPH proceeds through the non-stop system. Further, based on the data from start/stop sensor, the fast shutter mechanism is controlled so as to close the at least one shutter at the end of the payload, before a second target vehicle enters the scanning zone. The time for the fast shutter to close is, for example, on the order of no less than 100 milliseconds, preferably on the order of 350 milliseconds. The latter closing time reflects a desired traffic pattern of vehicles allowing for at least 15 feet of separation between the payload trailing edge and the leading edge of the next target vehicle. The fast shutter mechanism utilizes at least one linear sliding shielding block 220 driven by an electromagnetic pulse actuated by a solenoid assembly 230. When this shutter is closed, the shutter shielding block 220 operates to attenuate the radiation source. i.e., gamma source, to within acceptable exposure levels. In both the open and closed position, the configuration of the shutter shielding block 220 provides shielding of off-axis radiation exiting from the source assembly. In a preferred embodiment, the shutter shielding block comprises tungsten, with a thickness of approximately 2.4 inches (6.1 cm). An exemplary solenoid assembly 230 includes two electromechanical solenoids 235a and 235b operating together and control by solenoid drive electronics 237 to pull the shutter to the open position. The solenoids 235a and 235b are sized to open the shutter and allow radiation from the radiation source to exit through beam aperture 210 independently in the case of the failure of the other. This design, though resulting in a slightly slower shutter opening time, avoids system shut down altogether. And as described further below, shutter diagnostics will report the solenoid failure, i.e., the slower than normal opening time to the shutter mechanism control system and alert operators to the need for repair. Referring to FIG. 4c, a return spring 240 is used to close the fast shutter 210 in response to the start/stop sensor sensing the end of the payload. The return spring 240 is sized to provide enough energy to close the shutter in less than approximately 350 milliseconds, yet not provide excess force required to open the shutter 210. Additionally, as shown in FIGS. 3a and 3b, the non-stop system may include at least one radiation sensor 50, such as at least one rate meter, for sensing the levels of radiation, i.e., gamma or x-ray, outside of the confines of the scanning zone 20 as a protective measure. An alarm or other indication mechanism is triggered by at least one radiation sensor 50 to alert drivers, passengers, and other 3rd parties to the potential for radiation exposure in the area around each scanning zone 20. The shutter assembly described herein with respect to the preferred embodiments of the present invention also includes a failsafe operating design, wherein the at least one shutter fails in a closed condition, thus ensuring safety from possible radiation exposure in the event of, for example, a system power failure, a system error, or shutter failure. Further, the control system for the shutter assembly includes safety interlocks. For example, a key switch is used to control access to the shutter controls. The primary shutter must be opened by an operator through a pushbutton and the secondary, i.e., fast, shutter, must be enabled through a pushbutton. The portal system must be functional and enabled for operation. Emergency stops de-energize both the primary and secondary shutters directly in the event of an operator initiated emergency stop condition. The control system indicates shutter status to an operator through open/close status sensors. In order to maintain desired operation conditions, shutter diagnostics are incorporated into the control system and shutter assembly to report shutter failure status by providing error signals for the following failure conditions: shutter opening time exceeds preset value; shutter closing time exceeds preset value; shutter in transit, i.e., not in closed or open position; shutter current exceeds preset value; and shutter temperature exceeds preset temperature. The latter is measured as the output of a solenoid field effect transistor switch temperature sensor. At least part of the source radiation from the radiation tower is detected by the detector array within the detector tower for imaging the contents of the target vehicle through a process of discrete photon counting. This process is described in detail in each of the patents and applications that have been incorporated herein by reference. By way of example, showing the detectors from the detector array are coupled through, for example, 16-channel processing units, RS-485 line drivers, and an RS-485 interface card to a computer, wherein the computer processes discrete photon count information received from the detectors and causes a display device to display an image of the contents of a target vehicle, in response thereto. In this particular example, the detectors are coupled in groups of 16 to 16-channel data processing circuits. Preferably, twenty (20) groups of detectors are used. In practice, the number of detectors used is variable depending on the height of the vehicles to be inspected and the resolution, i.e., number of pixels, in the image desired. In a preferred embodiment, 320 detectors are used. The data processing circuits, of which there are preferably twenty (2), are each coupled to an RS-485 line driver, which is coupled to an RS-485 interface. The RS-485 interface is embodied on a circuit card located within a computer system. A suitable RS-485 interface is available as Model No. 516-485, Part No. 3054 from Seal Level Systems, Inc., and from numerous other vendors under respective model/part number designations. Each of the radiation detectors is coupled to a preamplifier within the 16-channel data processing circuits. Each preamplifier is coupled to an amplifier, which is in turn coupled to a discriminator. Each discriminator is coupled to a pulse generator, which generates an electrical pulse for each photon received into the radiation detector coupled thereto. The pulse generators within each of the 16-channel data processing circuits is coupled to a line driver. Each of the 16-channel data processing circuits includes its own line driver. The line drivers operate under the programmatic control of a firmware operating system. In operation, the preamplifiers, and amplifiers function in a conventional manner to amplify signals generated by the detectors. Outputs of the amplifiers are passed along to the discriminators, which impose a noise threshold on the amplified signal. Waveforms within the amplified signal that exceed the noise threshold are passed along to the pulse generator, which generates a pulse for each waveform within the amplified signal corresponding to a received gamma-ray or x-ray photon. The line driver passes the pulses generated by each of the pulse generators within a particular 16-channel data processing circuit along to the computer system via the RS-485 interface. The computer system operates programmatically under the control of a software system. The computer system receives detector pulses from each of the 16-channel data processors, in response to the detection of individual photons by the detectors. The software system processes the incoming detector pulses, evaluating their relative amplitudes, i.e., energies, and generates a radio graphic image-like display output signal in response thereto. The radio graphic, image-like display output signal is coupled to the graphical display device and is used by the graphical display device to generate a graphical representation of the densities within the vehicle under inspection. In summary, the system described herein is utilized to generate a graphical representation, i.e., a “picture”, of the densities of the contents of the target vehicle. Advantageously, this allows for easy visual interpretation of the results of the scanning of the target vehicle under inspection, as opposed to interpreting more subtle indications of the densities within the vehicle under inspection as may be required in prior art systems. The preferred imaging software system causes the display of a reference image simultaneous with the image generated in response to the target vehicle under inspection, so that an operator of the present embodiment can easily make a visual comparison between what a vehicle of the same type or having the same type of cargo being inspected should “look like”, and what the vehicle and cargo under inspection actually “looks like”. Such side-by-side inspection further simplifies the detection of contraband using the present embodiment. The embodiments and descriptions set forth herein are intended to be exemplary and not inclusive. One skilled in the art recognizes the numerous variations and equivalent components that may be used in accordance with the described invention and this fall within the scope thereof. |
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description | This application is a continuation of U.S. Nonprovisional application Ser. No. 17/097,915 filed Nov. 13, 2020, which is a continuation of U.S. Nonprovisional application Ser. No. 16/713,843 filed Dec. 13, 2019, which claims the benefit of and priority to U.S. provisional application Ser. No. 62/779,822 filed Dec. 14, 2018, the disclosures of which are incorporated by reference herein in the their entirety. None. In embodiments, the present disclosure relates generally to the field of radiation shielding and shielding of hadrons such as protons, neutrons, pions, and heavy ions associated with hadron therapy and with applications to shielding of photons in radio therapy. In embodiments, the present disclosure relates generally to the field of radiation shielding, where optimization of shielding material independent from structure may be beneficial, including but not limited to radiation therapy, nuclear power, scientific research, and industrial accelerators Particle generation and acceleration facilities are used in many applications, such as for scientific research, power generation, and industrial non-destructive inspections and medical treatment. Radiation in the form of photon (x-ray and gamma ray) and electron beams have been used for diagnostic, therapeutic, targeting, industrial, aerospace and research purposes for many years. Energy levels employed for these purposes range from the low KeV levels (5 KeV to 250 KeV) up to 25 MeV, with 10 MeV to 25 MeV photon and electron beams representing the highest energies typically employed in radiation therapy today. Since these radiation types and energy levels have historically represented the overwhelming majority of all such uses, the vaults built to contain this radiation have historically employed materials, means and methods most suited to the combination of physics challenges which are unique to those types of radiation and the energy and intensity levels being so employed. Given that set of physics challenges, the goals were relatively simple: stop or contain the electrons and photons and/or any other forms of secondary ionizing radiation produced by interactions of the primary radiation sources. High energy electron beams as well as any secondary (scatter) radiation they produce are relatively easily stopped. High energy photons are much more penetrating and produce much more scatter radiation, and thus require much more substantial shielding structures (vaults). Accordingly, the physics of photon radiation, penetration and attenuation are the dominant considerations in the formulation of conventional radiation therapy shielding solutions; i.e. in the selection of materials used and in the design and construction of the containment vault. Historically, the most commonly employed solution to these physics requirements and constraints has been the concrete vault and/or the concrete block with walls and ceilings ranging from two (2) to eight (8) feet thick wherein the concrete served to satisfy the requirements of shielding while also serving as the structure, or being structurally independent. In recent years, another solution has been introduced that separates the shielding and structural components and satisfies each of these two requirements using different materials. For example, the PRO System vault and the Temporary Radiotherapy Vault (TRV), by RAD Technology Medical Systems, each use an assembly of steel modules to satisfy the structural requirements of the vault and these modules also act as vessels to contain “any sufficiently dense granular material that can be readily and locally sourced” to satisfy the shielding requirement. These existing RAD Technology solutions allow the typical radiation oncology or industrial vaults to be modular and easily transportable, but are often physically larger than a poured concrete or concrete block vault due to the use of shielding materials that are less dense than concrete. The difference in overall size (footprint) is usually not significant enough to be meaningful due to the relatively low energies. But the difference in terms of transportability, recoverability and adaptability represents a paradigm shift in the shielding industry. That said, RAD Technology's existing vaults share one common characteristic with the traditional concrete vault: they are designed and built to shield against mid-range energy photons and even lower energy secondary neutrons produced from them. Secondary neutron radiation, though, is a relatively small and therefore less consequential consideration. By adding an inch or two of borated polyethylene and maybe some additional plywood or gypsum, the small amount of secondary neutron radiation is handled: the fundamental design of the vault remains the same. In recent years, however, proton accelerators have grown in favor and popularized a new and different treatment modality: Proton Therapy. These proton accelerators operate at energies more than a full order of magnitude greater than photon and electron beam modalities, and come with a whole new set of physics challenges and a consequent need for new shielding solutions. Radiation from the production and/or use of protons, neutrons, or other heavy particles; e.g., hadrons, whether the primary beam or secondary radiation created as a byproduct of the primary beam, must be shielded to protect nearby personnel, the public, and equipment. As such, the facilities that contain this equipment must be designed and constructed to provide adequate attenuation of various radiation types, energies and intensities to prevent exposure to people and, sometimes, equipment-both inside and outside of the facility. Radiation levels both inside and outside of such facilities must also comply with appropriate federal and state regulations. Proton and other heavy ion accelerator facilities are generally made of concrete walls, ceilings and floors that can have thicknesses of 8 to 20 feet or more. The concrete participates in both the shielding and structure of the facility. This, however, has proven very costly in terms of time, money and real estate (size/footprint). With energies sometimes in excess of 250 MeV/nucleon (proton or neutron) accelerating the more massive proton and heavy ion particles (such as carbon ions), the shielding physics challenges are not only more substantial, but fundamentally different from conventional radiation therapy. The dominant concern of this new challenge is neutron penetration. Protons and neutrons are over 1800 times more massive than electrons and the accelerating energies of these new particle beam accelerators can be more than 10 times greater than the highest energies traditionally employed in photon and electron beam modalities. Like gamma radiation, neutrons undergo scattering and absorption interactions with matter. These interactions form the basis for methods used to shield neutron radiation. However, unlike gamma radiation, which interacts primarily with the atomic electrons in matter, neutrons interact primarily with the atomic nuclei. Consequently, the types of materials favored for neutron shielding are quite different than the dense, high atomic number absorbers which are most effective in the attenuation of gamma radiation. In general, for fast neutrons, scattering interactions are more likely than capture interactions. Moreover, as the energy of neutrons is reduced through scattering interactions, additional neutron interactions, such as capture, increase in probability and number. Interactions of high energy protons (or heavy ions) with objects or components within the accelerating device, in the air, inside the patient, with other objects in the room, and even with the shielding walls themselves, cause secondary, or scatter, radiation. This also occurs with the traditional photon and electron beam modalities. However, unlike with the photon and electron modalities, the more massive hadronic particles at these higher energies undergo different interactions and produce significant levels of neutron radiation covering a wide spectrum of energies, ranging from near zero up to the beam energy. Each different energy particle undergoes different primary reactions with different reaction probabilities. The protons are essentially fully absorbed in the patient, while the secondary particles produced, photons and most importantly the neutrons-penetrate to the shielding barriers and become the primary shielding challenge. This broad spectrum, high-energy, high-fluence neutron radiation challenge requires a fundamentally different shielding approach. In addition, a significant challenge of this new radiation environment is “activation” wherein the traditional shielding material-concrete-becomes radioactive due to prolonged exposure to very high energy radiation. Some components of this “activated” concrete take years, and even decades, to decay to safe levels and thereby can represent both an immediate and a long-term safety hazard. Traditional hadron and radiation facilities have numerous disadvantages from a shielding standpoint. Traditional shielding walls generally consist of a concrete mixture and are formed in place through a continuous pour operation which leads to scheduling difficulties and a great deal of lost time, which translates to lost market opportunity (revenue). The requisite use of extremely thick concrete walls adds to the hadron beam facility's already large cost and footprint, and decreases the amount of usable space, both within the facility and on the property itself. Moreover, it does not allow for easy repair or modification of the resulting structure. Decommissioning and removal of the structure at the end of its useful life is complicated by the need to remove and properly dispose of radioactive material in the shielding barrier. In traditional concrete shielding vaults, some of the concrete barrier material becomes radioactively activated as a result of long term bombardment by large, high energy particles. Having a significant radioactive half-life, that material must either be left in place, secured and isolated from human interaction, or broken down and disposed of in accordance with applicable laws and regulations at significant expense of labor, time and money. In addition, concrete is inhomogeneous, which can lead to inconsistent shielding density or other property variations in the shielding walls and deterioration over time, resulting in incomplete capture and/or slowing of radiative particles. The use of concrete can also necessitate embedding, within the poured structure, multiple conduits and ducts, which can be large in number and must be, by construct, complicated in path to ensure no voids through the shielding. Because the shielding walls are structural in a conventional poured concrete center, reinforcing bar (rebar) material is also embedded in the concrete walls to increases the tensile strength of the structure. Conduit paths must not only be circuitous to avoid creating shielding voids, but must also be managed within a rebar grid which is costly and time-consuming to design and place. The shielding solution here presented is non-structural, and therefore no such rebar grid is required. Moreover, conduits can be placed in modules prior to being brought to the site, again reducing total on-site construction time for complicated designs. Unlike poured concrete, should future system changes or upgrades require modifications to or expansions of the conduits or ducts, or should there be problematic issues discovered with an existing layout, the removable fill design solution here presented would allow for modifications to any and all penetrations through the shielding. In embodiments, the present disclosure addresses the challenges identified herein including, but not limited to (a) removing the need for the shielding to be structural; (b) allowing for easier transport of the shielding material, facilitating re-use or effective decommissioning; (c) facilitating easy installation and removal of shielding materials; (d) optimization of neutron attenuation based on a variety of fundamental process interactions; (e) reduction of long lasting (long half-life) activation of the shielding material and of decommissioning costs and difficulties. In embodiments, the present disclosure is a facility comprising: a. a device configured to generate a beam of radiative energy having an energy range of 5 MeV to 500 MeV, b. a first shielding barrier surrounding the device, wherein a thickness of the first shielding barrier is 0.5 meter to 6 meters, and wherein the first shielding barrier comprises: i. a first radiation shielding wall surrounding the device, ii. a second radiation shielding wall surrounding the first radiation shielding wall, iii. radiation shielding fill material positioned between the first radiation shielding wall and the second radiation shielding wall forming a first barrier, wherein the radiation shielding fill material comprises at least fifty percent by weight of an element having atomic number between 12 and 83, and. In embodiments, the element having atomic number from 12 to 83 is selected from the group consisting of iron, lead, tungsten and titanium. In yet another embodiment, the radiation shielding fill material comprises at least fifty percent by weight of at least one of magnetite and hematite. In another embodiment, the radiation shielding fill material is granular. In another embodiment, the energy range of the beam is selected from the group consisting of 5 MeV to 70 MeV, 5 MeV to 250 MeV, and 5 MeV to 300 MeV. In yet other embodiments, at least one of the first radiation shielding wall and the second radiation shielding wall comprises panels mounted onto a structural exoskeleton. In yet another embodiment, at least one of the first radiation shielding wall and the second radiation shielding wall is steel. In another embodiment, the facility further comprises a second shielding barrier, wherein the second shielding barrier comprises: a third radiation shielding wall surrounding the second radiation shielding wall of the first shielding barrier; and second radiation shielding fill material is positioned between the second radiation shielding wall and the third radiation shielding wall of the second shielding barrier, wherein the second radiation shielding fill material comprises at least 25 percent by weight of an element having atomic number from 1 to 8, and wherein a thickness of the second shielding barrier is 0.5 meter to 6 meters. In an embodiment, the third radiation shielding wall comprises panels mounted onto a structural exoskeleton. In another embodiment, the third radiation shielding wall is steel. In yet another embodiment, the element having atomic number between 1 and 8 is selected from the group consisting of hydrogen, carbon, oxygen and boron. In an embodiment, the second radiation shielding fill material comprises at least one of borax, gypsum, colemanite, a plastic composite material, or lime. In an embodiment, the beam of radiative energy comprises at least one of: particles or photons. In an embodiment, the particles are hadrons. In an embodiment, the hadrons comprise at least one of protons, neutrons, pions, deuterons, heavier ions (having A>2), or any combination thereof In yet another embodiment, the present disclosure is a facility comprising: a. a plurality of electronic devices, b. a first shielding barrier surrounding the plurality of electronic devices, wherein a thickness of the first shielding barrier is 0.5 meter to 6 meters, and wherein the first shielding barrier comprises: i. a first radiation shielding wall surrounding the plurality of electronic devices, ii. a second radiation shielding wall surrounding the first radiation shielding wall, iii. radiation shielding fill material positioned between the first radiation shielding wall wherein the radiation shielding fill material comprises at least fifty percent by weight of an element having atomic number from 12 to 83. In yet another embodiment, the element having atomic number between 12 and 83 is selected from the group consisting of iron, lead, tungsten and titanium. In embodiments, radiation shielding fill material comprises at least fifty percent by weight of at least one of magnetite and hematite. In embodiments, the radiation shielding fill material is granular. In an embodiment, at least one of the first radiation shielding wall and the second radiation shielding wall comprises panels mounted onto a structural exoskeleton. In another embodiment, at least one of the first radiation shielding wall and the second radiation shielding wall is steel. In another embodiment, the facility comprises a second shielding barrier, wherein the second shielding barrier comprises: a third radiation shielding wall surrounded by the first radiation shielding wall of the first shielding barrier, and a second radiation shielding fill material positioned between the first radiation shielding wall of the first shielding barrier and the third radiation shielding wall of the second shielding barrier, wherein the second radiation shielding fill material comprises at least 25 percent by weight of an element having atomic number from 1 to 8, and wherein a thickness of the second shielding barrier is 0.5 meter to 6 meters. In embodiments, the third radiation shielding wall comprises panels mounted onto a structural exoskeleton. In another embodiment, the third radiation shielding wall is steel. In yet other embodiments, the element having atomic number from 1 to 8 is selected from the group consisting of hydrogen, carbon, oxygen and boron. In embodiments, the second radiation shielding fill material comprises at least one of borax, gypsum, colemanite, a plastic composite material, or lime. In some embodiments, the first shielding barrier is structural. In some embodiments, the first shielding barrier is non-structural. In some embodiments, the second shielding barrier is structural. In some embodiments, the second shielding barrier is non-structural. In some embodiments, there may be additional shielding barriers. For example, there may be three, four, five, six, seven, eight, and so on, shielding barriers. Some or all of these shielding barriers may be structural. Some or all of these shielding barriers may be non-structural. The figures constitute a part of this specification and include illustrative embodiments of the present disclosure and illustrate various objects and features thereof. Further, the figures are not necessarily to scale, some features may be exaggerated to show details of particular components. In addition, any measurements, specifications and the like shown in the figures are intended to be illustrative, and not restrictive. Therefore, specific structural and functional details disclosed herein are not to be interpreted as limiting, but merely as a representative basis for teaching one skilled in the art to variously employ the present disclosure. Among those benefits and improvements that have been disclosed, other objects and advantages of this disclosure will become apparent from the following description taken in conjunction with the accompanying figures. Detailed embodiments of the present disclosure are disclosed herein; however, it is to be understood that the disclosed embodiments are merely illustrative of the disclosure that may be embodied in various forms. In addition, each of the examples given in connection with the various embodiments of the disclosure which are intended to be illustrative, and not restrictive. Throughout the specification and claims, the following terms take the meanings explicitly associated herein, unless the context clearly dictates otherwise. The phrases “in one embodiment” and “in some embodiments” as used herein do not necessarily refer to the same embodiment(s), though it may. Furthermore, the phrases “in another embodiment” and “in some other embodiments” as used herein do not necessarily refer to a different embodiment, although it may. Thus, as described below, various embodiments of the disclosure may be readily combined, without departing from the scope or spirit of the disclosure. In addition, as used herein, the term “or” is an inclusive “or” operator, and is equivalent to the term “and/or,” unless the context clearly dictates otherwise. The term “based on” is not exclusive and allows for being based on additional factors not described, unless the context clearly dictates otherwise. In addition, throughout the specification, the meaning of “a,” “an,” and “the” include plural references. The meaning of “in” includes “in” and “on.” The following disclosure is used, at least in part, to support the embodiments detailed herein. In embodiments, the present disclosure addresses: (1) hadron beam applications such as proton and heavier ion therapy, and other applications such as power generation where neutron shielding is of primary concern; (2) the use of modular shielding specifically as a method to facilitate optimal shielding material choice and design, such as presented here for broad spectrum neutron attenuation; (3) the use of non-structural, iron-ore (or other) materials that are nonetheless a part of room wall composition; (4) a solution for transportable neutron shielding (as opposed to beam dump and other fixed shielding applications); and (5) the use of multiple barriers of different composition to allow for better optimization of a shielding wall. In embodiments, the present disclosure is directed to a modular approach to hadron (proton, neutron, pion, heavy ion, etc.) shielding, providing a combination of both transportability in shielding and the ability to tune the radiation shielding solution to optimize for the type of radiation (proton, neutron, pion, etc.), and for a broad and continuous spectrum of energies. For evaluating the effects of ionizing radiation on humans, the physical dose is determined by measuring the energy absorbed at a given point in a small test volume of a human tissue equivalent medium. For other forms of radiation, neutrons in particular, the biological effect is further dependent on the radiation type and energy. Just as the effects of 1 MeV neutrons are different from the effects of 200 MeV neutrons, the effects, biological and otherwise, of 200 MeV neutrons are vastly different from the effects of 200 MeV protons or 200 MeV photons. In the case of neutrons, the physical (absorbed) dose, expressed as Gray units and measured in joules/kilogram, is multiplied by an energy-dependent Conversion Coefficient, Sv(E) to yield Sievert dose, or effective dose (E). Furthermore, when the radiation energy is a distribution (a spectrum), the product of Sv(E) and fluence, f(E), must be integrated over all relevant spectral energies. For the convolution of Sv(E) and f(E), Sv(E) must be expressed as an equivalent discontinuous function, wk. The ICRP92, 2007 Publication 103 Radiation Weighting Factors, wk, for radiation type k, are given as numbers and as continuous curves for certain neutron and other particle energy bands as follows: Weighting Factors: By Particle Type and Energy Photons, electrons and muons of all energies: wk=1 “Slow” or “Thermal” Neutrons of E<1 MeV: wk=2.5+18.2 exp(−(ln(E)) 2/6) “Fast” Neutrons of E from 1 to 50 MeV: wk=5+17.2 exp(−(ln(2E)) 2/6) “High Energy Fast” Neutrons of E>50 MeV: wk=2.5+3.5 exp(−(ln(0.04E)) 2/6) Protons E>2 MeV: wk=2 Alpha particles, fission fragments and heavy nuclei of all energies: wk=20 (maximum) Damage to electronics is different from damage to humans, but it also follows an energy-dependent spectrum with a neutron damage peak typically at about 1 MeV which is clearly different from the above, where the higher energy ranges have the largest wk (weighting) values. Secondary neutron radiation is the predominant shielding challenge in a proton or other hadronic beam facility such as those used in carbon ion radiotherapy, and in general for many applications involving various high energy beams (hadronic, or others). FIGS. 1a and 1b demonstrate neutron fluence distributions created from an example proton beam incident on a water phantom (simulating human tissue), or target, using two different approaches. In FIG. 1A, the spatial beam coverage directly downstream of the incident beam on target is divided into equal areas at a typical treatment room distance away. This way, the number of neutrons per area can be viewed directly as corresponding neutron fluence. In FIG. 1B, the area of each segment changes but the increment in radius remains constant. This approach allows one to evaluate to what degree the number of neutrons changes with increasing radius from the primary beam direction. Both approaches, however, result in the same fluence behavior as a function of radius. Radiation source energy, as well as production geometry, may also be considered in shielding applications. The average neutron energy and fluence can vary with changes in incident beam angle but the maximum energy of the neutron that results from, for example, a 230 MeV proton beam at 0 degrees (perpendicular to the barrier) may be up to the incident proton energy minus the binding energy required to release neutrons from any material in the beam path. As the neutron travels through a shielding barrier, it interacts with the shielding material and the energy of the neutron decreases with each interaction by an amount dependent on the type and severity of interaction. Via these interactions, the neutron energies can decrease to ˜eV levels, 6 or more orders of magnitude less than the highest eV energies. This creates a broad spectrum of energies, covering a range of weighting factors (wk) as noted above. Moreover, different beam currents may be utilized for different situations. In a radiation oncology setting, this is typically mandated by the dose prescribed for the patient for a given treatment. However, this fluence can also be energy-dependent as is the case with the energy degrader systems deployed in cyclotron type accelerators. There are various types of interactions which play a role in neutron attenuation, including, but not limited to, ionization and nuclear fragmentation. Ionization describes the removal of a charged particle from a neutral atom. Nuclear fragmentation processes are where larger nuclei fragment into smaller nuclei. In some embodiments, the present disclosure is directed to a facility configured to perform “non-destructive testing.” As used herein, the term “non-destructive testing” refers to techniques for evaluating the properties of a material, component, or system without causing damage to the material, component, or system. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 350 kV to 1.5 MeV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 350 kV to 1 MeV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a particle beam having an energy range of 350 kV to 500 kV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a particle beam having an energy range of 350 kV to 400 kV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 400 kV to 1.5 MeV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 500 kV to 1.5 MeV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 1 MeV to 1.5 MeV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 400 kV to 500 MeV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 400 kV to 1 MeV. In some embodiments, the facility configured to perform non-destructive testing includes a device configured to generate a beam having an energy range of 500 kV to 1 MeV. In embodiments, the present disclosure, among other things, facilitates optimization of solutions ranging from absorption of slow (thermal) neutrons (<1 MeV) to moderation of fast and high energy fast neutrons (1 MeV up to the beam energy). In some embodiments, the facility includes a particle beam having an energy range of 5 MeV to 500 MeV located within the first and/or second barriers. In some embodiments, the energy range of the beam or radiation source located within the facility is 5 MeV to 400 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 5 MeV to 300 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 5 MeV to 250 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 5 MeV to 150 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 5 MeV to 100 MeV. In, the energy range of the beam or radiation source located within the facility is 5 MeV to 75 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 5 MeV to 50 MeV. In some embodiments, the facility includes a beam or radiation source having an energy range of 50 MeV to 500 MeV located within the first and/or second barriers. In some embodiments, the energy range of the beam or radiation source located within the facility is 100 MeV to 500 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 150 MeV to 500 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 250 MeV to 500 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 300 MeV to 500 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 400 MeV to 500 MeV. In some embodiments, the energy range of the beam or radiation source located within the facility is 1 MeV to 5 MeV. In some embodiments the energy range of the beam or radiation source located within the facility is not limited. For instance, in some embodiments, the energy can be as low as 1 keV. In some embodiments, the energy can exceed 100 GeV. In embodiments, the present disclosure provides a shielding solution that is modular and transportable. This is achieved by separating the shielding component of the resulting shielding facility (vault) from its structural component. In other words, the structural goals are achieved using one set of materials and methods while the shielding goals are met using a different set of materials and methods. In embodiments, the present disclosure adopts attenuating materials previously discounted and disregarded due to their absence of structural properties. This fact is here leveraged in particular to allow for broad energy spectrum absorption, but also encompasses other desirable benefits. There are multiple and sometimes conflicting properties determining the desirability and effectiveness of different shielding materials such as, but not limited to, low cost, availability, homogeneity, non-solubility, high density or high atomic number, low atomic number, minimal neutron regeneration, high neutron capture cross section, compactability, ease of use, low toxicity, and low radiation activation potential. In embodiments, the present disclosure relates to hadron beam production and generation, cosmic rays, and any radiation facility structure wherein the shielding is not a structural element of the facility structure and allows for the use of a variety of granular shielding materials. In embodiments, the first barrier radiation shielding fill material comprises element(s) having an adequate interaction cross-section (a measure of interaction probability which may be measured in barn units) to optimize the shielding performance of the barrier. In embodiments, the radiation shielding fill material may be determine based, at least in part, on the data shown in Table 1 below. TABLE 1Neutron Cross SectionsElasticInelasticCaptureElementΔ E (MeV)Δσ (barn)Δ E (MeV)Δσ (barn)Δ E (MeV)Δσ (barn)Magnetite168O0.0001-2149.2−241.0−21 2.74-2344.3−26−6.2−230.0001-203.9−28−5.4−265626Fe0.0001-2244.05−23−5.4−21 0.85-20.38.4−24−1.45−220.0001-201.15−21−7.0−27Colemanite11H 0.001-2422.0−21−3.9−2410−6−205.3−24−2.7−27105B10−6 −2341.2−234.4−2010−6−2344.4−24−1.0−20 0.01-208.2−30−2.7−25168O0.0001-2149.2−241.0−21 2.74-2344.3−26−6.2−230.0001-203.9−28−5.4−264020Ca 0.001-2327.4−22−2.4−23 0.1-2391.5−29−1.3−22 0.001-206.3−27−8.8−23Concrete11H 0.001-2422.0−21−3.9−2410−6−205.3−24−2.7−27105B10−6 −2341.2−234.4−2010−6−2344.4−24−1.0−20 0.01-208.2−30−2.7−25168O0.0001-2149.2−241.0−21 2.74-2344.3−26−6.2−230.0001-203.9−28 −5.4−262713Al 0.001-2321.6−23−2.4−21 1.0-2326.9−24−9.8−23 0.001-204.3−27−9.1−232814Si 0.001-2321.7−231.3−211.275-2232.6−25−1.2−2210−6−203.2−27 −6.7−234020Ca 0.001-2327.4−22−2.4−23 0.1-2391.5−29−1.3−22 0.001-206.3−27−8.8−23 Table 1 (above) provides the range of cross sections of interest for shielding for proton therapy cancer treatments for different types of energy absorption mechanisms (elastic and inelastic scattering, and capture reactions). Here, the relatively high capture cross sections for low MeV neutrons in Boron are evident. It is also instructive to look at the elastic scattering cross section range for hydrogen in concrete. Here, the cross section is high for the low energy end of the spectrum, but comparably small for the high energy neutrons. In embodiments, the present disclosure highlights the optimization of neutron shielding over a broad spectrum of energies. This approach facilitates not only all requisite human protection, but also reduces damage to electronic components where, for example, single event effects (SEEs) and upsets (SEUs) can cause equipment malfunction in treatment rooms, or—in other applications—large warehouse-type computer server facilities or strategic ground-based electronics. SEEs can be an issue even in low dose areas and are caused largely by hadrons such as protons or thermal neutrons. Without a structural requirement on it, or even a “self-supporting structural integrity” requirement (such as with concrete block), the radiation shielding fill material can be optimized for maximum full energy spectrum neutron absorption, and predominantly for higher energy neutrons through a focus on nucleus fragmentation. Neutrons of different energies are stopped, absorbed or otherwise mitigated by different neutron termination processes. In some embodiments, the present disclosure represents a shielding solution that focuses and capitalizes on nuclear fragmentation (also known as “spallation”), as opposed to the current industry-standard dependence on ionization processes associated with concrete walls. In embodiments, the present disclosure is configured to provide shielding barriers that increase attenuation levels in the 1 MeV range to provide an application specific radiation barrier for electronic equipment. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 83 (hereinafter “a high-Z element”) or a multi-barrier or dual barrier comprising both material having a high-Z element(s) and material having elements with an atomic number from 1 to 8 (hereinafter “a low-Z element”). The role for this can be seen, for example, in a proton therapy facility, where the ˜1 MeV neutrons are the dominant concern for radiation damage to electronics, while the quality factor (Q), the multiple of a measured dose, employed in consideration of dose to humans is higher for the ˜200 MeV neutrons. The large number of transmitted low energy (“slow”, or “thermal”) neutrons generated in the last few inches of a treatment room shielding wall do not contribute significantly to the transmitted dose to employees or general population in the center—and so they are typically ignored in concrete and other standard shielding approaches. However, with a binary barrier using embodiments of the present disclosure detailed herein, the low energy neutrons can be absorbed as well in a second barrier to protect also electronics. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 70. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 65. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 60. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 50. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 40. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 30. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 25. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 20. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 12 to 15. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 15 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 20 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 25 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 30 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 40 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 50 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 60 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 65 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 70 to 83. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 15 to 70. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 20 to 65. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 25 to 60. In embodiments, the present disclosure is a single barrier comprising a material having element(s) with an atomic number from 30 to 50. In embodiments, the present disclosure is a single-barrier or multi-barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 1 to 8 (hereinafter “a low-Z element”). In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 1 to 7. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 1 to 6. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 1 to 5. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 1 to 4. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 1 to 3. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 1 to 2. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 2 to 8. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 3 to 8. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 4 to 8. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 5 to 8. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 6 to 8. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 7 to 8. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 2 to 7. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 3 to 6. In embodiments, the present disclosure is a multi-barrier or dual barrier comprising both material having a high-Z element(s) in any range detailed herein and material having elements with an atomic number from 4 to 5. In embodiments, the present disclosure herein described can fulfill decommissioning requirements because it provides for a way to more easily extract the shielding material from the walls by it being a loose granular fill material, and because there is potentially less material that is susceptible to long term activation. Moreover, because the potentially radioactive shielding material to be removed could be chosen to have a substantially faster decay time (shorter half-life) measured in seconds, days or weeks rather than years or decades, and because it is not a structural part of the building, there is greater overall safety during the decommissioning process. With the design presented herein, unlike in conventional concrete shielded structures, the overall structure may remain intact and safe for workers while the shielding material is removed. In embodiments, the present disclosure provides a new approach to the construction of hadron beam facilities in which the facility is constructed with an inner and outer exoskeleton that provides the structure of the building. Between the inner and outer exoskeleton is a series of containers, vessels, or voids formed between inner and outer walls comprising, or mounted on, the exoskeleton. These voids are filled with a radiation shielding fill material that is non-structural. As used herein, the term “non-structural” means non-load bearing; not even capable of being self-supporting as in the case of concrete blocks. Thus, a material that is “non-structural” does not solidify or provide structure or support of any kind. Because the radiation shielding fill material is non-structural, unlike concrete which is structural, the composition of the radiation shielding fill material can be selected primarily for its radiation shielding capabilities and its mechanism of shielding without regard to any structural considerations or requirements. In embodiments of the present disclosure, the radiation shielding fill material is positioned between a first radiation shielding wall and a second radiation shielding wall forming a first barrier. In some embodiments, the radiation shielding fill material includes material with high-Z elements and/or other materials that rely on nuclear fragmentation as the predominant method of attenuation. Non-limiting examples of radiation shielding fill material high-Z elements include iron, lead, tungsten and titanium. In some embodiments, the radiation shielding fill material includes magnetite, hematite, goethite, limonite or siderite. In embodiments, the radiation shielding fill material is in the form of an aggregate and thus, is a granular material. In embodiments of the present disclosure, the radiation shielding fill material comprises at least fifty percent by weight of at least one high-Z element. In embodiments of the present disclosure, the radiation shielding fill material comprises at least sixty percent by weight of at least one high-Z element. In embodiments of the present disclosure, the radiation shielding fill material comprises at least seventy percent by weight of at least one high-Z element. In embodiments of the present disclosure, the radiation shielding fill material comprises at least eighty percent by weight of at least one high-Z element. In embodiments of the present disclosure, the radiation shielding fill material comprises at least ninety percent by weight of at least one high-Z element. In embodiments of the present disclosure, the radiation shielding fill material comprises at least 95 percent by weight of at least one high-Z element. In embodiments of the present disclosure, the radiation shielding fill material comprises at least fifty percent by weight of iron, lead, tungsten, titanium, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material comprises at least sixty percent by weight of iron, lead, tungsten, titanium, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material comprises at least seventy percent by weight of iron, lead, tungsten, titanium, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material comprises at least eighty percent by weight of iron, lead, tungsten, titanium, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material comprises at least ninety percent by weight of iron, lead, tungsten, titanium, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material comprises at least 95 percent by weight of iron, lead, tungsten, titanium, or combinations thereof In embodiments, the selection of the high-Z element(s) for the radiation shielding is based, at least in part, on the nuclear binding energy. Iron, in its various forms (isotopes), is the most abundant element on earth while nickel is the twenty second most abundant element in the earth's crust and not very accessible or cheap. Of all nuclides, iron has the lowest mass per nucleon and highest nuclear binding energy (8.8 MeV per nucleon in 56Fe, the most common iron isotope at 91.75% natural abundance), rendering it one of the most tightly bound nuclei, exceeded only by 58Fe (0.28% natural abundance) and the rare 62Ni (3.6% natural abundance). We here employ these facts for shielding. Iron-ore materials have the largest binding energy of all readily available shielding materials. This means that more energy is needed (expended), on average, to knock a neutron free from an iron nucleus than from other nuclei and, therefore, these materials absorb substantial energy—making iron an optimal, while also available, shielding material—in the fragmentation processes being herein leveraged by some embodiments of the present disclosure. Iron-ore materials enhance the natural “Faraday cage” environment of the steel modules which contain them. This is important to applications where electromagnetic fields may cause background noise or interference with signals of interest, for instance, in sensitive research laboratory equipment or in medical applications such as Magnetic Resonance Imaging (MRI). Faraday cages are used specifically to protect sensitive electronic equipment from external radio frequency interference (RFI), or to enclose devices that produce RFI, such as cellular and radio transmitters, to prevent their radio waves from interfering with other nearby equipment. They are also used to protect people and equipment against electric currents such as electrostatic discharges. Emergency radio communications typically found at medical facilities could also be subject to interference. In some embodiments, a thickness of the first barrier is 0.5 meters to 10 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 9 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 8 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 7 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 6 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 5 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 4 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 3 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 2 meters. In some embodiments, a thickness of the first barrier is 0.5 meters to 1 meters. In some embodiments, a thickness of the first barrier is 1 meters to 10 meters. In some embodiments, a thickness of the first barrier is 2 meters to 10 meters. In some embodiments, a thickness of the first barrier is 3 meters to 10 meters. In some embodiments, a thickness of the first barrier is 4 meters to 10 meters. In some embodiments, a thickness of the first barrier is 5 meters to 10 meters. In some embodiments, a thickness of the first barrier is 6 meters to 10 meters. In some embodiments, a thickness of the first barrier is 7 meters to 10 meters. In some embodiments, a thickness of the first barrier is 8 meters to 10 meters. In some embodiments, a thickness of the first barrier is 9 meters to 10 meters. In some embodiments, a thickness of the first barrier is 2 meters to 9 meters. In some embodiments, a thickness of the first barrier is 3 meters to 8 meters. In some embodiments, a thickness of the first barrier is 4 meters to 7 meters. In some embodiments, a thickness of the first barrier is 5 meters to 6 meters. In some embodiments, the first barrier or the second barrier comprises a plurality of sensors. In other embodiments, the sensors are configured to detect when the shielding material in the first barrier should be removed. In embodiments, the sensors are configured to detect when the shielding material in the first barrier has been activated. In embodiments, the sensors are timers configured to determine when to remove the shielding material in the first barrier. In embodiments, the sensors are calibrated to measure radiation produced within the enclosed vault. In embodiments, a second barrier of a different shielding material is utilized. Here, high energy fast neutrons are stopped or slowed by reactions within a high density (for instance material with high-Z element(s)), but these reactions cause the creation of lower energy fast and/or slow or thermal neutrons. For the latter, high density materials do not necessarily provide the optimal shielding, as different reactions are dominant in different energy ranges. To optimally absorb this lower energy radiation, secondary inner barriers that include at least one low-Z element may be deployed. Such a second inner barrier may be provided, for instance, within a treatment room to protect electronics. Alternatively, such a second outer barrier may be provided, for instance, external to the treatment room wall to provide additional protection for employees. In embodiments, a multi-barrier option may also be deployed wherein for example the high-density material is encased on both sides by a material having low-Z elements as above to accomplish both interior and exterior low energy shielding optimization. This approach could be used additionally for cases of, for example, side-by-side treatment rooms where either interior or exterior shielding is needed, but the interior of one room is the exterior of the neighboring room. In embodiments of the present disclosure, the radiation shielding fill material is positioned between a second radiation shielding wall and a third radiation shielding wall forming the second barrier. In some embodiments, the radiation shielding fill material includes material with low-Z elements. Non-limiting examples of radiation shielding fill material low-Z elements include hydrogen, carbon, oxygen and boron. In some embodiments, the radiation shielding fill material includes at least one of borax, gypsum, colemanite, a plastic composite material, or lime. In embodiments, the radiation shielding fill material is in the form of an aggregate and thus, is a granular material. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least fifty percent by weight of at least one low-Z element. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least sixty percent by weight of at least one low-Z element, the radiation shielding fill material forming the second barrier comprises at least seventy percent by weight of at least one low-Z element. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least eighty percent by weight of at least one low-Z element. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least ninety percent by weight of at least one low-Z element. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least 95 percent by weight of at least one low-Z element. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least fifty percent by weight of hydrogen, carbon, oxygen, boron, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least sixty percent by weight of hydrogen, carbon, oxygen, boron, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least seventy percent by weight of hydrogen, carbon, oxygen, boron, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least eighty percent by weight of hydrogen, carbon, oxygen, boron, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least ninety percent by weight of hydrogen, carbon, oxygen, boron, or combinations thereof. In embodiments of the present disclosure, the radiation shielding fill material forming the second barrier comprises at least 95 percent by weight of hydrogen, carbon, oxygen, boron, or combinations thereof In some embodiments, a thickness of the second barrier is 0.5 meters to 10 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 9 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 8 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 7 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 6 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 5 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 4 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 3 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 2 meters. In some embodiments, a thickness of the second barrier is 0.5 meters to 1 meters. In some embodiments, a thickness of the second barrier is 1 meters to 10 meters. In some embodiments, a thickness of the second barrier is 2 meters to 10 meters. In some embodiments, a thickness of the second barrier is 3 meters to 10 meters. In some embodiments, a thickness of the second barrier is 4 meters to 10 meters. In some embodiments, a thickness of the second barrier is 5 meters to 10 meters. In some embodiments, a thickness of the second barrier is 6 meters to 10 meters. In some embodiments, a thickness of the second barrier is 7 meters to 10 meters. In some embodiments, a thickness of the second barrier is 8 meters to 10 meters. In some embodiments, a thickness of the second barrier is 9 meters to 10 meters. In some embodiments, a thickness of the second barrier is 2 meters to 9 meters. In some embodiments, a thickness of the second barrier is 3 meters to 8 meters. In some embodiments, a thickness of the second barrier is 4 meters to 7 meters. In some embodiments, a thickness of the second barrier is 5 meters to 6 meters. In some embodiments, the first barrier comprises material having low-Z elements and the second barrier comprises material having high-Z elements. In other words, in some embodiments, the first barrier is configured consistent with the configuration of the second barrier detailed herein and the second barrier is configured consistent with the configuration of the first barrier as detailed herein. In embodiments, at least one of the first and/or second barrier comprises a combination of material having low-Z elements and material having high-Z elements. In embodiments, the facility may include third, fourth, fifth, sixth, seventh or more barriers having material and thicknesses detailed herein with respect to the first and/or second barriers depending on the requirements of the facility. In embodiments, any of the barriers (first, second, third, fourth or more) may be formed of a plurality of sections. In embodiments, the plurality of sections of each barrier may be configured to allow for removal of a portion of the radiation fill material forming the barrier. In embodiments, the barrier may be comprised of individual modular sections that may be combined to form the first and/or second barriers. In embodiments, each of the individual modular sections may be removed after use and replaced with a modular section filled with unused radiation shielding fill material. In embodiments, one or more of the individual modular sections may include a sensor as detailed herein for indicating when the radiation barrier fill material in the section requires replacement. In embodiments, certain materials can be used as sensors to determine a dose of radiation. For instance, plastic turns yellow in the presence of radiation and also darkens at a certain level. In embodiments, the present disclosure includes a shielding wall containing an optimized radiation shielding fill material that does not need to be as thick as a shielding wall made from non-optimized materials such as concrete to achieve the same level of radiation shielding. In embodiments, a shielding wall of a proton beam facility having shielding walls filled with material comprising high-Z elements as detailed herein can be reduced in thickness by 5% to 25% as compared to a concrete or concrete block shielding wall while providing the same or better shielding capability. In some embodiments, the radiation shielding fill material includes a series of voids that are filled with different radiation shielding materials so as to provide different barriers of shielding in certain directions, which can serve to provide more specifically tailored radiation shielding capabilities and/or size efficiencies. FIG. 2 shows the relative distribution of processes that contribute to final termination of motion of neutrons traversing a binary shielding wall/barrier composed of Magnetite and Colemanite aggregates (left, identified as a “binary barrier”) according to an embodiment of the present disclosure as compared to a prior art barrier composed of poured concrete (right). The numbers were obtained from a GEANT4 Monte Carlo simulation, where the neutrons were produced in a water target simulating a patient in a proton radiotherapy treatment room. As used herein, a “GEANT4 Monte Carlo simulation” is developed to determine transmitted neutron dose as the basis for the barrier neutron attenuation performance, Geant4 is a publicly-available (see http://geant4.web.cern.ch) “toolkit” for the simulation of the passage of particles through matter. Its areas of application include high energy, nuclear and accelerator physics, as well as studies in medical and space science. The three main reference papers for Geant4 are published in Nuclear Instruments and Methods in Physics Research A 506 (2003) 250-303, IEEE Transactions on Nuclear Science 53 No. 1 (2006) 270-278 and Nuclear Instruments and Methods in Physics Research A 835 (2016) 186-225. FIGS. 3 and 4 and Table 2 present examples of different materials studied for binary and non-binary wall composition. This study is for a 3 m total binary barrier thickness, with alpha=the ratio of thicknesses of a first barrier (A) element to a second subsequent barrier (B) element. Hence, alpha=infinity is a non-composite, single material 3 m wall composed of material A. TABLE 2A257∞Barrier CompositionTransmitted Sievert Dose (mSv/year)Concrete———2.404Magnetite + Colemanite0.3480.3180.1970.178Hematite + Colemanite0.2950.2600.2570.263Magnetite + Gypsum0.2210.1890.1830.178 FIG. 5 illustrates unshielded neutron fluence angular distributions directly downstream of a 230 MeV proton beam incident on a water target (simulated proton radiotherapy patient). The processes listed in the FIG. 2 are the possible interactions evaluated by the simulation within the shielding barrier, and they are based both on the type of radiated particle (the primary particle) and on the secondary particles with which they interact. FIG. 2, however, was generated exclusively for the secondary neutron spectrum produced from a 230 MeV proton beam incident on a water target (simulated human), which comprises about 91% of the shielding challenge in a proton therapy center. This modeling of a 230 MeV proton beam incident on a water target (simulated patient) within a typical concrete barrier reveals that the dominant neutron motion termination process of a concrete barrier is ionization, with electronic ionization constituting approximately 60% and hadronic ionization constituting approximately 10% of the total neutron termination processes. Nuclear fragmentation only accounts for about 16% of the total termination processes in a concrete barrier. This contrasts with the design presented in embodiments of the present disclosure that relies most heavily on nuclear fragmentation. Nuclear fragmentation absorbs more energy and is thus a more efficient method that allows for a thinner and more transportable barrier. We note again here that this element of transportability and the need for increased efficiency; i.e. a smaller footprint, are additional motivations for separating the structural and shielding components of the solution. Both the electromagnetic and radiation shielding properties of the proposed technology are multi-directional. In other words, a person standing outside of a radiation therapy treatment room can be shielded from the radiation produced therein by a shielding barrier/wall, or electronics in the treatment room could be shielded from radiation occurring as a result of interactions inside the shielding barriers/walls (secondary, or scatter, radiation) by a strategically-chosen material barrier on the interior wall, and/or electronic components in the room could be shielded from electromagnetic signals or other radiation generated outside of the room. In a multi-material composition barrier approach, as another example, a wall between adjacent treatment rooms could provide shielding to both rooms. Though this is true as well for concrete, the approach presented here provides more efficient shielding (translating to reduced barrier thickness and lower cost) across a broader energy spectrum with the added benefit of efficiently shielding against high-energy, high-fluence, neutron radiation not found in concrete vaults designed and constructed to contain the less energetic photon and electron beams. In another example, sensitive electronics, for example, could be placed in a smaller shielding room inside a larger, unprotected, facility or in a facility where radiation was being produced. In all the above applications, it should be noted that the dual or multi barrier approach allows for multiple materials to be employed in different barriers, once again providing a broader spectrum and optimization of attenuation. While Iron-ore materials may be used for one barrier, for example, less dense materials may be used for another to optimize low energy neutron absorption. FIGS. 3 and 4 compare, for example, the performance of a conventional concrete wall and a modular, transportable binary barrier wall composed of varying, relative amounts of Magnetite (MR2) and Colemanite (CR2) according to an embodiment of the present disclosure. Here, the ratio α=LA/LB, i.e. the ratio of thickness of the first barrier encountered by the neutrons (A) to the second (B). α corresponding to infinity, then, is a pure Magnetite barrier. The safety-requisite limitation of 2 mSv/year transmitted Sievert dose (“TSD”) typically determines the minimum allowable wall thickness. In this example, the circle size is proportional to the dose of transmitted neutrons in each case; i.e. the TSD. In all cases, the modular transportable wall, leveraging and optimizing the neutron absorption process of nuclear fragmentation, is a superior approach. The results presented in the figure come from a GEANT4 Monte Carlo simulation, and were scaled to a somewhat aggressive annual clinical use dose of a proton therapy machine (corresponding to 5×1015 protons/year). As compared to a structural concrete shielding wall relying on ionization as the predominant neutron termination process, the predominant neutron termination process of a shielding wall primarily composed of (a) high-Z element(s) according to the present disclosure is nuclear fragmentation. As herein shown, by selecting and leveraging the more efficient attenuating mechanism of nuclear fragmentation as the predominant neutron termination process, we achieve the greatest radiation absorption and demonstrate an improved, more efficient, shielding barrier. Thus, as shown in FIGS. 3 and 4, the thickness of a radiation shielding fill material barrier is less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 5% to 25% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 5% to 20% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 5% to 15% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 5% to 10% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 10% to 25% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 15% to 25% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 20% to 25% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. In embodiments, the thickness of a radiation shielding fill material barrier is 5%, 10%, 15%, 20% or 25% less than a thickness of a concrete wall to achieve the same Transmitted Sievert Dose. FIGS. 5, 6a, 6b and 6c depict a GEANT4 ray-trace of a beam (in color black) incident on a water target cylinder (simulating a patient) producing secondary neutron rays and other particles emanating from the target, passing through a binary barrier according to an embodiment of the present disclosure, and finally through a simulated detector volume. As shown in the figures, very few neutrons penetrate the first portion of the barrier, an observation that led us to investigate what was the dominant absorption mechanism at work in the primary barrier. FIG. 7 illustrates a multi-story modular proton therapy facility 700 according to an embodiment of the present disclosure. The facility includes a plurality of modules 701 configured to be used together to form the facility. In embodiments, one or more of the plurality of modules 701 are filled, at least in part, by shielding fill material (not shown). FIG. 8 shows an exploded view of the modular proton therapy facility 700 shown in FIG. 7. In some embodiments, a top set of the plurality of modules 701 are a binary layer system having one set of modules (not shown) disposed below another set of modules (not shown), each having the same or differing thicknesses as determined by site specific design parameters. FIGS. 9 and 10 illustrate side elevation views in full section of a non-limiting example of a multi-story modular proton therapy facility 900 similar to the facility 700 shown in FIG. 7. The figures include an optional internal barrier wall 902 positioned between the outer walls 903. FIG. 10 further illustrates the corridors for gaining access to the high radiation areas on each of the lower three (3) levels. FIG. 11 illustrates a plan view of the bottom set of modules 701 (containing the inner barrier 1104 shielding material) which are part of the top level of a non-limiting example of a multi-story modular proton therapy facility 1100 (and 700). The depicted facility is constructed with two barriers of shielding material (i.e. an inner barrier 1104 and an outer barrier 1105), indicated by the two different shaded areas above and surrounding exemplary treatment room (shown in 1206 of FIG. 12). The top set of modules making up this top level (not shown) would contain the same shielding as the outer barrier 1105. In some embodiments, a removable core 1106 may allow removal of shielding material through the roof for easy access to key components for installation, removal and/or repair. In some embodiments, the interior space of the facility of the present disclosure can be divided into multiple interior rooms that can be arranged to accommodate people and/or equipment in need of shielding. For example, in some embodiments, people and/or sensitive electronics can be in interior rooms of the facility and shielded from external radiation. Alternatively, in other embodiments, radiation emitting sources can be in interior rooms of a facility and people outside the facility can be shielded by the shielding walls from radiation produced by the primary and secondary radiation emitting sources inside the facility. FIG. 12 illustrates a plan view of the lower levels of a non-limiting example of a multi-story modular proton therapy facility 1200. FIG. 12 includes an inner barrier 1204, an outer barrier 1205, and an entrance maze (corridor) and treatment room (indicated by the white space) having a proton delivery device 1206 therein. In one form of the present disclosure, a hadron beam facility is constructed from a series of pre-fabricated modules that are constructed off site, shipped to the site, and then assembled together at the build site to form the structural exoskeleton of the hadron beam facility vault as well as all necessary non-shielding spaces (clinical, mechanical, etc.). The shielding modules are preferably prefabricated with the desired interior structures of the building, using conventional modular construction techniques. However, specific to the unique radiation shielding needs of a hadron beam facility, each shielding module has an exterior structural frame, typically steel, comprised of various panels. Some sides of each module are composed of metal walls (“panels”) while other sides are left open. The panels on the various modules are oriented such that when the modules are assembled together, the various panels align with the panels in the modules above or below and optionally with the modules to either side so as to create relatively continuous inner and outer walls that frame out void spaces. These void spaces are subsequently filled with the selected radiation shielding material. The structural frames of the various modules, once connected together, combine to form the inner and outer exoskeleton of the building, and the panels comprising or mounted to the modules combine to form the inner and outer walls that establish the void spaces that contain the radiation shielding fill material. There can be intermediate walls between the inner and outer walls constructed in the same fashion such that there are multiple void spaces that may be filled with different types of shielding material. The modules also contain the interior finishes of the corresponding functional spaces of the facility, such as the waiting room, the control room, the treatment room containing the patient table and gantry for the proton therapy device (for example), etc. Details of building a radiotherapy facility in this modular fashion with a single barrier of granular shielding material is described more fully in U.S. Pat. No. 6,973,758 to Zeik et al. and U.S. Pat. No. 9,027,297 to Lefkus, et al., incorporated herein by reference, and this approach can be applied to create a hadron beam facility by appropriate modification of the interior spaces and the shielding wall arrangements, number of walls and consequent number of void spaces and shielding materials, thicknesses and materials for the desired configuration of the hadron beam facility. In one refinement, the shielding wall can be created with distinct compartments that can be separately filled with different radiation shielding fill materials. These distinct compartments can serve a number of purposes. For example, by creating distinct compartments through the thickness of the shielding wall, a layered wall can be created with an inner barrier (inner meaning closest to the radiation source) having one type of fill material optimized for one type of attenuating interaction and an outer later (outer meaning farther from the radiation source) optimized for another type of attenuating interaction. For example, the inner barrier may serve to slow high energy neutrons to lower energy states while the outer barrier may serve to absorb the slower, lower energy, neutrons. Additional barriers can be created in similar fashion, resulting in a two, three, four or more shielding barriers. As explained above, the radiation shielding fill material for each barrier is non-structural, and thus a wide range of materials are possible. This approach creates an apparatus for broad energy spectrum shielding, leveraging in each material the dominant process of relevance for any given application (radiation type and energy range). Most semiconductor electronic components are susceptible to radiation damage. Prolonged exposure to residual ionizing radiation, such as neutrons, may destroy the electronics of the medical equipment in particle therapy facilities. Some medical facilities change charge-coupled device (CDD) cameras monthly and others purchase expensive radiation hardened equipment that can better withstand the challenging environment. To address this, one or more of the shielding barriers can be optimized to reduce the residual ionizing radiation. An example would be a secondary barrier of fill containing a hydrogen-rich material like gypsum (optimal for moderating fast neutrons), or a boron rich material like borax or colemanite (optimal for capturing slow neutrons). This method, while aimed at hadron particle therapy, is applicable to electronic components in a variety of radiation environments, even including low-level radiation environments such as large warehouse-type computer server facilities or strategic ground-based electronics where even terrestrial or cosmic rays can cause loss of security via SEEs. The particles which cause significant soft fails in electronics are neutrons, protons, and pions. Alternatively, or in addition to the creation of partitions through the thickness of the shielding wall; i.e. inner and outer barriers, lateral partitions can be created in the shielding fill material. One use of lateral partitions is to allow specific sections of the shielding wall to be removed independently of the other sections. This is particularly useful for areas that are exposed to the most radiation and have the potential to become activated. By creating distinct fill containing vessels in the potential activation area, those distinct vessels can be regularly tested and then removed and disposed of should they become activated, without needing to dismantle the entire wall of which they are a part. In cases where it may be easier to remove the activated sections in large blocks/sections, a grout can be introduced into the fill material to cause it to solidify into the most manageable size, which facilitates the most economical means of removal, transportation and disposal. Fluid conduits can be embedded in the sections to facilitate the introduction of the grout. Radiation sensors may also be embedded in different sections of the shielding wall. The radiation sensors can detect the level of radiation reaching each wall section and can also be used to determine if a particular section has become activated and needs to be removed. The loose aggregate method suggested here lends itself to this type of apparatus, as it allows for the instrumentation to be accessed and removed for maintenance, upgrades, and repair. This is not possible with sensors embedded in poured concrete without conduits for cable runs to instrumentation, which cause unwanted voids in the shielding. The panels that create the innermost walls, ceiling, and floor separating the radiation shielding fill from the vault room may be made of steel or other conductive material such that they create a de facto Faraday cage around the central vault room or wherever necessary or desirable. This Faraday cage is beneficial in avoiding communication interference or introduction of noise into any circuitry of any kind in the region of the proton vault, including in the proton accelerator, its related electrical and electronic components and all other computers and electrical and electronic devices throughout and immediately neighboring the facility. Simulations of the shielding properties of a binary barrier for a proton therapy center according to the present disclosure were modeled for different wall thicknesses. The modeled barrier of the disclosure was a binary barrier with an inner barrier of magnetite (barrier A) and an outer barrier of colemanite (barrier B). Four different ratios of the thickness of the inner magnetite barrier to the thickness of the outer colemanite barrier (α=barrier A/barrier B) were modeled: 2, 5, 7 and infinity (the latter corresponding to a single barrier of magnetite and no barrier of colemanite). As compared to the modeled results for a comparably thick concrete wall, the modeled inventive barriers all substantially outperformed the concrete wall. It was found that a 3-meter thickness of the modeled barrier (including a barrier of only magnetite) would provide sufficient shielding for a 230 MeV proton beam energy to reduce the transmitted Seivert dose to well below the target of 2 mSv/year as illustrated by FIGS. 3 and 4. In embodiments, the present disclosure is designed to make it easier to remove when it has ended its useful life. Decommissioning radiation facilities involves safely removing a facility from service and eliminating or reducing any residual radioactivity to a level that permits any radiation use license to be terminated, with the property released either for unrestricted use or, at worst, under specified restricted conditions. In embodiments, the present disclosure facilitates a faster and less expensive decommissioning, as any radioactive material could either be retracted from the vessels via suction or hardened into them and subsequently removed in the form of manageably sized blocks. In some embodiments, the granular nature of the material would allow the separation of activated components from non-activated components. In some embodiments, at least some of the separated materials can be saved. In some embodiments, at least some of the separated materials can be stored. In some embodiments, at least some of the separated materials can be disposed of. In some embodiments, at least some of the separated materials can be sold. Any of the suitable technologies set forth and incorporated herein may be used to implement various example aspects of the disclosure as would be apparent to one of skill in the art. In one aspect of the disclosure, a process for designing and constructing a radiation shielding facility is provided. The initial step is to determine what is to be protected. For example, this may be humans, electronics, or both. Having determined the thing(s) to be shielded, one then determines the neutron energy range of interest, the radiation intensity, and the maximum dosage allowed. As noted above, these quantities are different for humans and electronics. The next step is to determine where the objects (people or equipment) to be shielded would be located in relation to the source of the radiation. The object(s) to be shielded may be on the same side as the primary radiation source, on the opposite side, or both. This determination leads to a selection of whether to use a simple (uni-directional) layered barrier approach or a bi-directional barrier approach. Next, based on the neutron energy range and direction radiation would be traversing the barrier, one would assess and determine which type of nuclear attenuation interaction most efficiently attenuates the radiation of that range and type, and then select a shielding material whose composition is leveraged toward the optimum type of nuclear attenuation interaction. The objective is to leverage the material property to increase the relative proportion of the most effective type of nuclear attenuation interactions; i.e. to maximize attenuation by selecting the most effective attenuation method(s) and using the materials that most effectively employ that (or those) method(s). Having selected the material and thus knowing its nuclear attenuation characteristics, a model is used to calculate the wall thickness needed to achieve the level of attenuation required to bring the transmitted radiation dose below the desired threshold. The process may be repeated for additional material barriers, with the design parameters being the type of shielding material (which determines its shielding characteristics), the thickness of the shielding barrier(s), and the order/arrangement of the barriers if more than one. The objective is to optimize the shielding materials based the characteristics of the entity to be shielded (human and/or electronics) and the relative location(s) of the entity or entities to be protected versus the radiation source and the barrier(s) of the shielding wall. An iterative process is contemplated in which the free variables can be one or more of (a) number of barriers; (b) material choice for each barrier; (c) material density for each barrier (as may be affected by compaction); (d) thickness of each barrier, (e) order or arrangement of each barrier if more than one, and (f) tolerable activation. While any number of materials can theoretically be chosen, it is envisioned that the materials chosen will be first based on their ability to preferentially leverage the more desirable or effective nuclear attenuation interactions, which, as described above, is a function of the chosen purpose of the shielding wall; i.e. the characteristics of the radiation being addressed as well as what is being protected/shielded. Moreover, the material selection process, in some embodiments, is directed to materials that are relatively inexpensive and/or readily available, which further restricts the scope of material choices. Thus, once the shielding challenge has been fully understood, determining the cost, availability and suitability of the available shielding materials is the reasonable next step. For example, given a scenario wherein it has been determined that a three-layered wall is the best solution and the desired properties of each layer have been established, one would first select three materials that are suitable to the task; i.e. optimized for a particular type of nuclear attenuation interaction, and that are also sufficiently available, and inexpensive. Then, having decided on the number of barriers and the material to be used in each barrier, a total wall thickness for all barriers combined is calculated, and simulations are then performed to model the radiation attenuation properties and effects using different relative thicknesses of the different barriers making up the shielding wall. The simulations can be optimized to find the most effective relative thicknesses of the different barriers for the given total wall thickness, and even the total wall thickness can be modified (and the iterative process repeated) if the simulation results so indicate. In embodiments, different total wall thicknesses may be initially selected and the process of optimizing the relative ratios of the relative thicknesses of each barrier may be repeated. In yet other embodiments, different starting materials can be selected and the process repeated to optimize wall construction parameters for different shielding materials. This method may be of most value in situations where it is desirable to minimize the building footprint, such as due to high land cost or site constraints. A higher cost shielding material may provide superior nuclear attenuation properties and results for a given shielding challenge. Thus, it may allow the overall thickness of the shielding wall to be smaller than if a less expensive shielding material were used, and the overall footprint of the facility may thereby be reduced. In such a case, the additional costs attributable to use of a higher cost shielding material can be offset by reduced land use costs and/or increased design freedom. In yet another embodiment of the present disclosure, the facility is designed to protect electronic devices or other equipment that may be negatively affected by the radiation. In the embodiment, the facility comprises a plurality of electronic devices or other equipment that may be negatively affected by radiation instead of the device configured to generate a beam. In view of the above, the fact that the shielding material does not participate in the structure of the facility and can be chosen based solely on its radiation shielding properties, as well as its cost and availability, provides new and unprecedented design freedoms. These design freedoms can be exploited according to the present disclosure to create shielding facility structures in places and at costs and at a pace of construction that were heretofore not possible. In some embodiments, optimization of the facility may be based on three key drivers. These three drivers can include, but are not limited to at least one of shielding performance, shielding space, or shielding cost. A non-limiting optimization solution driven by shielding cost, shielding space, and shielding performance is depicted in FIG. 13. An exemplary flow chart depicting how the non-limiting optimization drivers of FIG. 13 may affect the design of an exemplary shielding facility is shown in FIG. 14. In some situations, shielding performance is a primary driver for facility design. Shielding performance includes optimization for type of challenge and level of attenuation desired. The next driver is shielding space available. The shielding space available includes optimization of available physical space to achieve a solution. The third driver is the shielding cost. The shielding costs includes optimization of the cost required to achieve acceptable performance. In some embodiments, a modular approach also allows for different shielding levels in different areas; e.g. higher attenuation in areas of higher radiation exposure or of higher occupancy levels. In some situations, shielding performance is the primary driver for the facility design. Shielding performance is predicated on providing the most effective solution to attenuate neutrons and other sub atomic particles. In the following non-limiting example, there is no concern for cost. In this example, sensitive electronic equipment requires protection from neutrons and other sub atomic particles. The integrity of the electronics over time requires a Transmitted Sievert Dose (mSv/year) of 0.20 which is ten times less that what humans can safely absorb. Based on the desire to protect the equipment, the highest performing solution must be selected. Additional considerations include the amount of space available. Space is a constraint of the physical barrier. The smaller the allowable area, the more efficient or high performing the barrier must be. The performance of the barrier may be optimized by selection of materials, their purity, compaction and volume. As noted above, in this example, cost would not be a driver. In some situations, performance may have several sub-drivers which may be optimized. For instance, one may optimize shielding performance based on several factors including but not limited to photons, neutrons, protons or a host of other challenges. In some situations, shielding space is the primary driver for the facility design. Shielding space can be the driver when an existing location provides physical constraints in the allowable amount of area available. In a non-limiting example, the courtyard of a facility is chosen to place new equipment due to proximity to existing operations and/or even sensitivity to public view. Shielding space is less than 3 meters and performance is 2.00 mSv/year. The limited shielding space does not offer adequate square footage for traditional shielding methods of concrete and block and the logistics for placing concrete are difficult. Thus, the efficiency of the shielding is the primary driver. Knowing the gross available area for the barrier, the next consideration would be performance; i.e. which materials would provide adequate protection in that limited space. In some embodiments, cost is not a primary driver. In some situations, shielding space may have several sub-drivers which may be optimized. For instance, one may optimize shielding space based on several factors including but not limited to vertical or horizontal limitations or gross volume. In some situations, the cost of shielding is the primary driver for the facility design. The cost of shielding could be the driver in greenfield commercial sites. There would not be space constraints and performance would be typical. In a non-limiting example, a new facility is being built with a medical device typically used in proton therapy. The university customer is required to bring in the lowest cost solution possible. Available land is not an issue and no special attenuation is required. Several acres of open space exist for the project. Dose rate limitations are again moderate at 2.00 mSv/year. The cost of the shielding would be the primary driver with standard performance a secondary consideration. Shielding materials would be selected based on cost of acquisition, which is affected by proximity to the site. In some embodiments there is a trade-off between purity and volume. In some embodiments more volume to achieve the same space equates to higher shipping costs. Thus, the shielding space available would not be a driver. In some situations, shielding cost may have several sub-drivers which may be optimized. For instance, one may optimize shielding costs based on several factors including but not limited to at least one of up-front savings, long-term savings, or time-savings. Within the three key drivers exist opportunities for optimization within the technical calculations. Depending, at least in part, on type and energy of the radiation to be shielded, different interactions may be leveraged and balanced. In some embodiments, the optimization can be conducted using a statistical weighting algorithm. Non-limiting quantities such as material cost or barrier size may be assigned an array of values through which the optimization algorithm can re-weigh the results to determine an optimized solution. In embodiments, Bayesian optimization of the weighted calculations may be deployed via a Monte Carlo sampling technique to scan through numerous options with statistical rigor in contrast to conventional shielding algorithms. The flexibility of the methods detailed herein will allow designers through algorithms and potentially machine learning and Artificial Intelligence, to evaluate various scenarios to achieve an established goal. Using this method, the range of materials, physical space, types of radiation (photonic, atomic or sub-atomic), specific energies and/or range of energies. The values for the energies are not limited. For instance, in some embodiments the energies can be as low as 1 keV. In some embodiments, the energies can be as high as 1000 GeV. Desired performance can also be optimized using predictive analytics. These methods, in some embodiments, may achieve results significantly different than the traditional approach of standard construction which may include limited variables by simply using more volume and/or denser aggregates. In embodiments, a first step in creating a proton therapy facility is to consider the treatment room wall that is protecting radiation therapists from the radiation being used to treat a patient lying on a bed inside an adjacent treatment room. The neutron energy for this application will range from near zero MeV up to the beam energy minus the binding energy of the shielding material(s). A maximum allowable Transmitted Sievert Dose for a radiation therapist is 2 mSv/Yr (the “Threshold Transmitted Sievert Dose”). Therapists work outside of the treatment room while beam is being delivered, so the design objective must consider neutrons coming from the room during beam delivery through the barrier and into areas where the therapist(s) could be working. (Protons are quickly and easily stopped and are not a factor beyond the fact that they spawn neutrons prior to being stopped.) In this application, it has been found that optimum shielding may be achieved by leveraging nuclear fragmentation processes via an iron-ore material. As illustrated herein, reduction of the Transmitted Sievert Dose (TSD) to below the Threshold Transmitted Sievert Dose can be achieved using a single barrier of such a material. In this case, a requisite barrier thickness would be less than concrete, which is typically deployed for a combination of structural and shielding properties. Additional barriers composed of different shielding materials may be included and the relative thicknesses of the multiple barriers optimized as described above. Multiple barriers of material may be used throughout the shielding walls of the facility or only in select locations. The locations for additional shielding barriers may be selected based on the anticipated radiation spectrum hitting different areas of the shielding wall, because in a particle therapy facility, the radiation spectrum is not uniform in all directions. The locations for additional shielding barriers may further be selected based on who or what is on the other side of that barrier, such as sensitive electronics or an un-controlled high occupancy waiting room. Thicker shielding, for instance, can be placed in the areas directly opposite the beam direction (which may form a vertically oriented circular “band” around a gantry which rotates a full 360 degrees). Additional barriers may be added and/or optimized based on the location of electronics within the treatment room. For this optimization, backscatter radiation (the radiation that is scattered back into the room after high energy neutrons (also called secondary, or scatter, radiation) have entered the shielding wall), is modeled and interior barriers of shielding material are selected to attenuate the radiation that would otherwise scatter back into the room and damage the electronics. Having selected shielding materials, iterative modeling of the combined radiation shielding characteristics is performed as explained above to find the necessary thicknesses of the different barriers to achieve the design parameter (i.e. Threshold Transmitted Sievert Dose to therapist of less than 2 mSv/year and/or the established maximum permissible dose to equipment). Current simulations have revealed magnetite to be a desirable shielding material for this type of proton facility. Hematite has also been found to be acceptable and may be less expensive. Although exemplary embodiments and applications of the disclosure have been described herein, including as described above and shown in the included example Figures, there is no intention that the disclosure be limited to these exemplary embodiments and applications or to the manner in which the exemplary embodiments and applications operate or are described herein. Indeed, many variations and modifications to the exemplary embodiments are possible, as are applications in fields beyond medicine such as research, power or strategic facilities, as would be apparent to a person of ordinary skill in the art. The disclosure may include any device, structure, method, or functionality, as long as the resulting device, structure or method falls within the scope of one of the claims that are allowed by the patent office based on this or any related patent application. While a number of embodiments of the present disclosure have been described, it is understood that these embodiments are illustrative only, and not restrictive, and that many modifications may become apparent to those of ordinary skill in the art. Further still, the various steps may be carried out in any desired order (and any desired steps may be added and/or any desired steps may be eliminated). |
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056195486 | claims | 1. A device for measuring properties of a thin-film disposed on a layered structure using scattering X-rays, said device comprising: a source adapted to produce X-rays; a curved surface positioned between said source and said thin-film, with the surface being in fixed orientation with respect to the layered structure and adapted to focus the X-rays onto a first focal area of the layered structure with one point of said area having X-rays impinging thereon at varying angles of incidence, with a reflected X-ray being associated with each of said plurality of X-rays; a detector positioned to sense said reflected X-rays, said detector adapted to produce a signal corresponding to an angle of reflection and an intensity of each of the reflected X-rays sensed; and a processor means, connected to receive signals produced by the detector, for determining properties of the thin-film based upon a comparison of the intensity and the angle of reflection of the reflected X-rays sensed, the properties including thin-film thickness. an X-ray reflector defining a focal area proximate to the thin-film; a source directing a plurality of X-rays onto the reflector, with the reflector directing the plurality of X-rays as a bundle of monochromatic X-rays onto said thin-film at a plurality of incident angles greater than the critical angle (.psi..sub.c), with each of the monochromatic X-rays reflecting from the thin-film at an angle of reflection corresponding to an angle of incidence, the reflector being a curved monochromator defining at least one focal point proximate to the substrate, whereby the curved monochromator simultaneously directs the plurality of X-rays onto the thin-film at varying angles of incidence; a detector positioned to sense X-rays reflected from the thin-film along a plane transverse to the plane of the thin-film, the detector producing a signal corresponding to an intensity and a spatial position of the transverse X-ray in the plane, with each spatial position in the plane corresponding to an angle of reflection; and a processor means connected to receive signals produced by the detector for determining physical properties of the thin-film based upon all signals received by the detector, the physical properties including the mass per unit area of the thin-film. a source of X-rays adapted to produce X-rays; an X-ray reflector positioned in the path of the X-rays and adapted to simultaneously impinge onto at least one point of the thin-film a plurality of said X-rays at a plurality of incident angles with each of the plurality of X-rays reflecting from the thin-film at an angle of reflection, defining reflected X-rays with the angle of reflection corresponding to an angle of incidence; a detector positioned to sense said reflected X-rays reflected from the thin-film and produce a signal corresponding to the intensity and an angle of reflection of the reflected X-rays sensed; and a processor means connected to receive signals produced by the detector for determining properties of the thin-film based upon a comparison of the intensity and the angle of reflection of the reflected X-rays sensed, the properties including thin-film thickness, thin-film density and a relative smoothness between the surface of the thin-film and the smoothness of the substrate-thin-film interface. 2. The device of claim 1 wherein the reflecting surface is a curved reflector. 3. The device of claim 1 wherein the reflecting surface is a curved monochromator. 4. The device of claim 1 wherein the reflecting surface is a curved monochromator having an ellipsoidal shape. 5. The device of claim 1 wherein the reflecting surface has a cylindrical shape, defining a second focal area, with the source positioned at the second focal area. 6. The device of claim 1 wherein the reflecting surface has a toroidal shape, defining a second focal area, with the source positioned at the second focal area. 7. The device of claim 1 wherein the detector is in a fixed rotational relationship with respect to said layered structure. 8. The device of claim 5 wherein the detector is a self-scanning diode array. 9. The device of claim 5 wherein the detector is a charge coupled device. 10. The device of claim 5 wherein the detector is a multiple-wire proportional counter. 11. The device of claim 5 wherein the detector is a multiple-anode microchannel detector. 12. The device of claim 5 wherein the detector is a photostimulated storage phosphor image detector. 13. The device of claim 5 wherein the varying angles of incidence are greater than a critical angle .psi..sub.c. 14. A device for measuring the structure of a thin-film disposed on a substrate, comprising: 15. The device of claim 14 wherein the reflector has a cylindrical shape defining a second focal area, with the source positioned at the second focal area with said first focal area corresponding to a line lying in the plane of the thin-film. 16. The device of claim 14 wherein the reflector has a toroidal shape defining a second focal point, with the source positioned at the second focal point. 17. The device of claim 14 wherein the detector is positioned to sense X-rays reflected from the thin-film along a plane transverse to the plane of the thin-film, the detector further including means for producing a signal corresponding to both intensity and a spatial position of the X-ray in the plane. 18. The device of claim 16 wherein the detector is a self-scanning diode array. 19. The device of claim 16 wherein the detector is a charge coupled device. 20. The device of claim 16 wherein the detector is a self-scanning diode array. 21. The device of claim 16 wherein the detector is a charge coupled device. 22. The device of claim 16 wherein the detector is a multiple-wire proportional counter. 23. A device for measuring the properties of a thin-film disposed on a substrate, defining a substrate-thin-film interface, comprising: 24. The device of claim 23 wherein said detector and said interface are in a fixed rotational relationship with respect to each other. 25. The device of claim 23 wherein the monochromator is ellipsoidal defining first and second focal points, with the source positioned at the first focal point and the second focal point laying in the plane of the thin-film. 26. The device of claim 23 wherein the detector is positioned to sense monochromatic X-rays reflected from the thin-film along a plane transverse to the plane of the thin-film, the detector further including means for producing a signal corresponding to both intensity and a spatial position of the monochromatic X-ray in the plane. 27. The device of claim 26 wherein the detector is a self-scanning diode array. 28. The device of claim 26 wherein the detector is a charge coupled device. 29. The device of claim 26 wherein the detector is a self-scanning diode array. 30. The device of claim 26 wherein the detector is a charge coupled device. 31. The device of claim 26 wherein the detector is a multiple-wire proportional counter. |
044249039 | abstract | A process and apparatus for storing tritium, particularly tritium waste from nuclear power plants, wherein the tritium is first oxidized to HTO or T.sub.2 O and is then bound to an adsorbent having molecular sieve properties, and the tritium-containing adsorbent being enclosed by a corrosion-resistant metal container hermetic with respect to hydrogen diffusion. |
claims | 1. A chamber for exposing a workpiece to charged particles propagated along a path, the chamber comprising:a charged particle source configured to generate a stream of charged particles;a collimator configured to collimate and direct the stream of charged particles from the charged particle source along an axis of propagation;a beam digitizer configured to create temporally and spatially resolved digital flashes comprising groups including at least one charged particle by adjusting longitudinal spacing between the charged particles along the axis of propagation;a deflector configured to deflect said digital flashes transversely to the direction of propagation, the deflector in said path between the collimator and the deflector;an objective lens assembly configured to demagnify said digital flashes or one or more groups of said digital flashes, the deflector in said path between the beam digitizer and the objective lens assembly; anda workpiece stage configured to hold a workpiece, the objective lens assembly between the deflector and the workpiece stage. 2. The chamber of claim 1, wherein the chamber is adapted to direct the stream of charged particles accelerating at potentials between about 5 and 500 keV. 3. The chamber of claim 1, wherein the charged particles comprise ions. 4. The chamber of claim 3, wherein the ions are positively charged. 5. The chamber of claim 3, wherein the ions are negatively charged. 6. The chamber of claim 3, wherein the ions are singly charged. 7. The chamber of claim 3, wherein the ions are doubly charged. 8. The chamber of claim 1, wherein the charged particles comprise electrons. 9. The chamber of claim 1, wherein the charged particles comprise positrons. 10. The chamber of claim 1, wherein the charged particle source comprises a liquid metal ion source (LMIS). 11. The chamber of claim 1, wherein the charged particle source comprises a plasma ion source (PIS). 12. The chamber of claim 1, wherein the charged particle source comprises a volume plasma ion source (VPIS). 13. The chamber of claim 1, wherein the charged particle source comprises a gas field ionization source (GFIS). 14. The chamber of claim 1, wherein the charged particle source comprises a carbon nanotube field emitter. 15. The chamber of claim 1, wherein the charged particle source comprises a free electron laser. 16. The chamber of claim 1, wherein the charged particle source comprises a pulsed ablation ion source. 17. The chamber of claim 1, wherein the charged particle source comprises a magnetically confined plasma anode source (MAP). 18. The chamber of claim 1, wherein the charged particle source comprises a thermal field emission electron source (TFE). 19. The chamber of claim 1, wherein the charged particle source is configured to generate a plurality of charged particle species. 20. The chamber of claim 1, wherein the charged particle source is configured to output the stream at accelerating potentials between about 5 and 30 keV. 21. The chamber of claim 1, wherein the charged particle source is configured to provide a current density of about 1×105 A/cm2 over an approximately 10 nm spot size measured at a workpiece held be the workpiece stage. 22. The chamber of claim 1, wherein the collimator comprises a lens. 23. The chamber of claim 1, wherein the collimator comprises reflective optics. 24. The chamber of claim 1, wherein the collimator comprises a lens and reflective optics. 25. The chamber of claim 1, wherein the collimator comprises two reflective optical elements. 26. The chamber of claim 1, wherein the collimator comprises a plurality of lenses. 27. The chamber of claim 1, wherein the collimator is configured to collimate the stream of charged particles. 28. The chamber of claim 1, wherein the collimator is configured to de-magnify the stream of charged particles to a Gaussian beam having less than about 1 μm full width half max diameter spot size. 29. The chamber of claim 1, wherein the collimator is configured to de-magnify the stream of charged particles to a Gaussian beam having a spot size less than about 100 nm full width half max diameter. 30. The chamber of claim 1, wherein the collimator is configured to de-magnify the stream of charged particles to a Gaussian beam having a spot size less than about 10 nm full width half max diameter. 31. The chamber of claim 1, further comprising a mass separator including a mass separator aperture plate, the mass separator in said path between the collimator and the beam digitizer and configured to deflect selected charged particle species into the mass separator aperture plate. 32. The chamber of claim 31, wherein the mass separator comprises reflective optics. 33. The chamber of claim 31, wherein the mass separator comprises an ExB Lens. 34. The chamber of claim 31, wherein the mass separator comprises a Wein filter. 35. The chamber of claim 1, further comprising a beam blanker configured to blank or compress the stream of charged particles or to both black and compress the stream of charged particles, wherein the beam digitizer is disposed in said path between the beam blanker and the deflector. 36. The chamber of claim 1, wherein the beam digitizer comprises a beam buncher. 37. The chamber of claim 36, wherein the beam buncher is configured to apply electromagnetic radiation having a frequency between about 1 MHz and 100 GHz. 38. The chamber of claim 36, wherein the beam buncher is configured to apply electromagnetic radiation having two or more resonant frequencies between about 1 MHz and 25 GHz. 39. The chamber of claim 36, wherein the beam buncher is configured to apply electromagnetic radiation and is configured to modulate an amplitude of the electromagnetic radiation. 40. The chamber of claim 36, wherein the beam buncher is configured to apply electromagnetic radiation and is configured to modulate a frequency of the electromagnetic radiation. 41. The chamber of claim 36, wherein the beam buncher comprises a series of electrodes, the beam buncher electrodes configured to be modulated by a voltage at a radio frequency (RF). 42. The chamber of claim 36, wherein the beam buncher is configured to apply electromagnetic radiation and wherein a mean velocity of the stream of charged particles is substantially similar to the frequency of the electromagnetic radiation. 43. The chamber of claim 36, wherein the beam buncher comprises a helical coil configured to be modulated by a voltage at a radio frequency (RF). 44. The chamber of claim 36, wherein the beam buncher is configured to alter the relative velocity of the particles thereby forming groups of said digital flashes, said groups forming a digital beam. 45. The chamber of claim 36, wherein the beam buncher is configured to apply an electric field configured to longitudinally compress or converge said charged particles into said digital flashes, said flashes forming a digital beam. 46. The chamber of claim 45, wherein an average velocity of the digital beam is between about 1×104 meters/second and 3×108 meters/second. 47. The chamber of claim 1, wherein the beam digitizer comprises a blanker configured to blank the stream of charged particles. 48. The chamber of claim 1, wherein the beam digitizer is configured to modulate an on/off state of the charged particle source so as to blank the stream of charged particles. 49. The chamber of claim 1, wherein the beam digitizer is configured to apply electromagnetic radiation configured to blank the stream of charged particles. 50. The chamber of claim 1, wherein the beam digitizer comprises a plasma beat wave modulator configured to blank the stream of charged particles. 51. The chamber of claim 1, wherein the beam digitizer comprises a space charge wake field modulator configured to blank the stream of charged particles. 52. The chamber of claim 1, wherein the beam digitizer comprises a resonant absorption space charge wake field modulator configured to blank the stream of charged particles. 53. The chamber of claim 1, wherein the beam digitizer comprises a generalized phased contrast modulator configured to blank the stream of charged particles. 54. The chamber of claim 1, wherein the beam digitizer comprises a pulsed incident neutralizing beam modulator configured to blank the stream of charged particles. 55. The chamber of claim 1, wherein the beam digitizer comprises a pulsed laser beam modulator configured to blank the stream of charged particles. 56. The chamber of claim 1, wherein the beam digitizer comprises a Bradbury-Nielson Gate (BNG) particle beam modulation device configured to blank the stream of charged particles. 57. The chamber of claim 1, wherein the beam digitizer is configured to create a digital beam comprising between about 1 and 7,000,000 charged particles per digital flash. 58. The chamber of claim 1, wherein the beam digitizer is configured to create a digital beam comprising between about 1 and 100,000 charged particles per digital flash. 59. The chamber of claim 1, wherein the beam digitizer is configured to create a digital beam comprising between about 1 and 10,000 charged particles per digital flash. 60. The chamber of claim 1, wherein the beam digitizer is configured to create a digital beam comprising between about 1 and 5,000 charged particles per digital flash. 61. The chamber of claim 1, wherein the beam digitizer is configured to create periods between pairs of said digital flashes of a digital beam that are directly adjacent of between about 1 nm and 9.99 meters of digital beam travel. 62. The chamber of claim 1, wherein the beam digitizer is configured to create periods between pairs of said digital flashes of a digital beam that are directly adjacent of between about 1 nm and 1 meter of digital beam travel. 63. The chamber of claim 1, wherein the beam digitizer is configured to create periods between pairs of said digital flashes of a digital beam that are directly adjacent of between about 1 nm and 10 cm of digital beam travel. 64. The chamber of claim 1, wherein the beam digitizer is configured to create periods between pairs of said digital flashes of a digital beam that are directly adjacent of between 1 nm and 1 cm of digital beam travel. 65. The chamber of claim 1, wherein the beam digitizer is configured to create periods between pairs of said digital flashes of a digital beam that are directly adjacent of between 1 nm and 100 μm of digital beam travel. 66. The chamber of claim 1, wherein the beam digitizer is configured to create periods between pairs of said digital flashes of a digital beam that are directly adjacent of between 1 nm and 10 μm of digital beam travel. 67. The chamber of claim 1, wherein the beam digitizer is configured to create periods between pairs of said digital flashes of a digital beam that are directly adjacent of between 1 nm and 1 μm of digital beam travel. 68. The chamber of claim 1, wherein the beam digitizer is configured to create a space of between about 1 nm and 10 meters of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 69. The chamber of claim 1, wherein the beam digitizer is configured to create a space of less than about 10 meters of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 70. The chamber of claim 1, wherein the beam digitizer is configured to create a space of less than about 1 meter of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 71. The chamber of claim 1, wherein the beam digitizer is configured to create a space of less than about 1 cm of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 72. The chamber of claim 1, wherein the beam digitizer is configured to create a space of less than about 1 mm of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 73. The chamber of claim 1, wherein the beam digitizer is configured to create a space of less than about 100 nm of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 74. The chamber of claim 1, wherein the beam digitizer is configured to create a space of less than about 10 nm of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 75. The chamber of claim 1, wherein the beam digitizer is configured to create a space of less than about 1 nm of beam travel between a pair of said digital flashes of a digital beam that are directly adjacent. 76. The chamber of claim 1, wherein the beam digitizer is configured to create a space between a pair of said digital flashes that are directly adjacent of between about 2 nm and 12 mm. 77. The chamber of claim 1, wherein the beam digitizer is configured to create variable spaces between one or more pairs of said digital flashes. 78. The chamber of claim 1, wherein spacing between one or more pairs of said digital flashes is substantially the same. 79. The chamber of claim 1, wherein spacing between adjacent said digital flashes is substantially harmonic. 80. The chamber of claim 1, wherein spacing between adjacent said digital flashes is substantially random. 81. The chamber of claim 1, wherein the digital flashes form a digital beam comprised of charged particle compression waves longitudinal to the axis of propagation. 82. The chamber of claim 1, wherein at least some of the digital flashes have a three-dimensionally Gaussian geometry at the workpiece stage. 83. The chamber of claim 1, wherein at least some of the digital flashes have a trapezoidal cross-section along the axis of propagation. 84. The chamber of claim 1, wherein the digital flashes are three-dimensional and have a geometric cross-section having the shape of a rectangle, triangle, circle, or square. 85. The chamber of claim 1, wherein a current density of at least some of the digital flashes is adjustable. 86. The chamber of claim 1, wherein the period between a pair of said digital flashes that are directly adjacent is adjustable. 87. The chamber of claim 1, wherein the space between a pair of digital flashes that are directly adjacent is adjustable. 88. The chamber of claim 1, wherein an address placement of said digital flashes is adjustable. 89. The chamber of claim 1, wherein the energy of individual digital flashes is adjustable. 90. The chamber of claim 1, wherein the size of individual digital flashes is adjustable. 91. The chamber of claim 1, wherein the deflector is configured to arrange individual digital flashes or a group of said digital flashes into a three-dimensional timespace. 92. The chamber of claim 1, wherein the deflector comprises an array of collimated deflection electrodes longitudinally disposed along the axis of propagation. 93. The chamber of claim 1, wherein the deflector is configured to deflect said digital flashes substantially perpendicularly to the axis of propagation. 94. The chamber of claim 1, wherein the deflector comprises a series of deflection electrode stages disposed longitudinally along the axis of propagation, wherein each deflection electrode stage comprises one or more electrodes. 95. The chamber of claim 94, wherein each deflection electrode stage comprises two or more electrodes. 96. The chamber of claim 94, wherein each deflection electrode stage comprises three or more electrodes. 97. The chamber of claim 94, wherein each deflection electrode stage comprises four or more electrodes. 98. The chamber of claim 94, wherein longitudinal positions of the deflection electrode stages are adjustable. 99. The chamber of claim 94, wherein the deflection electrode stages are configured to synchronize deflection of said digital flashes to create a distributed pattern of individual digital flashes, a group of said digital flashes, or an adjustable virtual digital stencil comprising a plurality of said digital flashes or groups of said digital flashes. 100. The chamber of claim 94, wherein potentials of each of the deflection electrode stages are configured to be synchronized with a mean velocity of the beam. 101. The chamber of claim 94, wherein potentials of each of the deflection electrode stages are configured to be harmonically synchronized with velocities of said digital flashes or groups of said digital flashes. 102. The chamber of claim 94, wherein the deflector comprises at least two deflection electrode stages and wherein every other deflection electrode stage is configured to displace a digital flash or groups of said digital flashes towards an intended trajectory. 103. The chamber of claim 94, wherein the deflector comprises at least three deflection electrode stages and wherein every third deflection electrode stage is configured to displace a digital flash or groups of said digital flashes towards an intended trajectory. 104. The chamber of claim 94, wherein the deflector comprises at least N deflection electrode stages and wherein every Nth deflection electrode stage is configured to displace a digital flash or groups of said digital flashes towards an intended trajectory. 105. The chamber of claim 94, wherein potentials of each of the deflection electrode stages are configured to be randomly synchronized with velocities of said digital flashes or groups of said digital flashes. 106. The chamber of claim 94, wherein potentials of each of the deflection electrode stages are configured to partially displace a digital flash or groups of said digital flashes towards an intended trajectory. 107. The chamber of claim 94, wherein potentials of one deflection electrode stage is configured to substantially fully displace a digital flash or groups of said digital flashes towards an intended trajectory. 108. The chamber of claim 94, wherein potentials of a plurality of deflection electrode stages are configured to substantially fully displace each said digital flash or groups of said digital flashes towards an intended trajectory. 109. The chamber of claim 94, wherein a phase of adjacent said digital flashes of a digital beam longitudinal to the axis is configured to be substantially equal and wherein spacing between the deflection electrode stages is configured to be synchronized with the phase. 110. The chamber of claim 94, wherein a phase of adjacent said digital flashes of a digital beam longitudinal to the axis is configured to be single harmonic and wherein spacing between the deflection electrode stages is configured to be synchronized with the phase. 111. The chamber of claim 94, wherein a phase of adjacent said digital flashes of a digital beam longitudinal to the axis is configured to be multiple harmonic and wherein spacing between the deflection electrode stages is configured to be synchronized with the phase. 112. The chamber of claim 94, wherein a phase of adjacent digital flashes of a digital beam longitudinal to the axis is configured to be random and wherein spacing between the deflection electrode stages is adapted to be synchronized with the phase. 113. The chamber of claim 94, wherein the deflector further comprises a digital feedback system. 114. The chamber of claim 94, wherein a field perimeter of deflection by the deflection electrode stages is defined as the minor deflection field and wherein a size of the field is dependent on the variable energy of the beam. 115. The chamber of claim 94, wherein a field perimeter of deflection by the deflection electrode stages is defined as the minor deflection field and is displaced transverse to the axis of propagation less than about 4 mm from the center of the axis of propagation. 116. The chamber of claim 94, wherein a field perimeter of deflection by the deflection electrode stages is defined as the minor deflection field and is displaced transverse to the axis of propagation less than about 2 mm from the center of the axis of propagation. 117. The chamber of claim 94, wherein a field perimeter of deflection by the deflection electrode stages is defined as the minor deflection field and is displaced transverse to the axis of propagation less than about 1 mm from the center of the axis of propagation. 118. The chamber of claim 94, wherein a field perimeter of deflection by the deflection electrode stages is defined as the minor deflection field and is displaced transverse to the axis of propagation less than about 100 μm from the center of the axis of propagation. 119. The chamber of claim 1, wherein the objective lens assembly is configured to demagnify, focus, and deflect individual digital flashes or groups of said digital flashes to expose a workpiece held by the workpiece stage to an adjustable virtual digital stencil. 120. The chamber of claim 1, wherein the objective lens assembly comprises an electromagnetic lens. 121. The chamber of claim 1, wherein the objective lens assembly comprises a plurality of deflection plates. 122. The chamber of claim 1, wherein the objective lens assembly comprises reflective optics. 123. The chamber of claim 1, wherein the objective lens assembly comprises a combination of reflective optics and a refractive lens. 124. The chamber of claim 1, wherein the objective lens assembly comprises a combination of reflective optics and deflection electrodes. 125. The chamber of claim 1, wherein the objective lens assembly comprises a combination of deflection electrodes and a refractive lens. 126. The chamber of claim 1, wherein the objective lens assembly comprises one or more deflection electrode stages. 127. The chamber of claim 126, wherein each deflection electrode stage of the objective lens assembly comprises at least one electrode. 128. The chamber of claim 1, wherein the objective lens assembly is configured to demagnify and focus a single one of said digital flashes or a group of said digital flashes by a factor of less than about 1000×. 129. The chamber of claim 1, wherein the objective lens assembly is configured to demagnify and focus a single one of said digital flashes or a group of said digital flashes by a factor of less than about 100×. 130. The chamber of claim 1, wherein the objective lens assembly is configured to demagnify and focus a single one of said digital flashes or a group of said digital flashes by a factor of less than about 10×. 131. The chamber of claim 1, wherein a field perimeter of deflection by the objective lens assembly is defined as the major deflection field and wherein a size of the field is dependent on the variable energy of the beam. 132. The chamber of claim 1, wherein a field perimeter of deflection by the objective lens assembly is defined as the major deflection field and is displaced transverse to the axis of propagation less than about 10 mm from the center of the axis of propagation. 133. The chamber of claim 1, wherein a field perimeter of deflection by the objective lens assembly is defined as the major deflection field and is displaced transverse to the axis of propagation less than about 4 mm from the center of the axis of propagation. 134. The chamber of claim 1, wherein a field perimeter of deflection by the objective lens assembly is defined as the major deflection field and is displaced transverse to the axis of propagation less than about 2 mm from the center of the axis of propagation. 135. The chamber of claim 1, wherein a field perimeter of deflection by the objective lens assembly is defined as the major deflection field and is displaced transverse to the axis of propagation less than about 1 mm from the center of the axis of propagation. 136. The chamber of claim 1, wherein a field perimeter of deflection by the objective lens assembly is defined as the major deflection field and is displaced transverse to the axis of propagation less than about 100 μm from the center of the axis of propagation. 137. The chamber of claim 1, wherein the workpiece stage is configured to move continuously over a dimension in X, Y, and Z axes. 138. The chamber of claim 1, wherein the workpiece stage is configured to move continuously over a dimension up to 600 mm in each direction perpendicular to the axis of propagation. 139. The chamber of claim 1, wherein the workpiece stage is configured to move continuously over a dimension up to 60 mm in each direction perpendicular the axis of propagation. 140. The chamber of claim 1, wherein the workpiece stage comprises a linear drive workpiece stage. 141. The chamber of claim 1, wherein the workpiece stage comprises an air bearing workpiece stage. 142. The chamber of claim 1, wherein the workpiece stage comprises an interferometer configured to determine a location of the workpiece stage on a horizontal plane. 143. The chamber of claim 1, wherein the workpiece stage comprises an interferometer configured to determine a location of the workpiece stage on a horizontal plane and on a vertical axis. 144. The chamber of claim 1, further comprising a workpiece stage control system configured to measure and adjust x, y, and z positions and yaw, pitch, and roll of the workpiece stage. 145. The chamber of claim 144, wherein the workpiece stage control system is configured to limit a motion of the workpiece stage. 146. The chamber of claim 144, wherein the workpiece stage control system is configured to limit a velocity of the workpiece stage. 147. The chamber of claim 144, wherein the workpiece stage control system is configured to limit a position of the workpiece stage. 148. The chamber of claim 144, wherein the workpiece stage control system is configured to limit a breakaway force of the workpiece stage. 149. The chamber of claim 144, wherein the workpiece stage control system is configured to limit a height of the workpiece stage. 150. The chamber of claim 144, wherein the workpiece stage control system is configured to limit a drag of the workpiece stage. 151. The chamber of claim 1, wherein the workpiece stage is configured to perform full motion writing (FMW) including continuous movement during exposing a workpiece held by the workpiece stage. 152. The chamber of claim 1, wherein the workpiece stage is configured to move without stopping for more than 5 nanoseconds during 0.5 seconds during exposing a workpiece held by the workpiece stage. 153. The chamber of claim 1, wherein the workpiece stage is configured to compensate for horizontal positions of at least one of said digital flashes, a group of said digital flashes, or an adaptable virtual digital stencil comprising a plurality of said digital flashes or groups of said digital flashes. 154. The chamber of claim 1, wherein the workpiece stage is configured to rotate a workpiece held by the workpiece stage during exposure of said workpiece at not greater than about 40,000 rpm. 155. The chamber of claim 1, wherein the workpiece stage is configured to electro-statically clamp a workpiece. 156. The chamber of claim 1, wherein the workpiece stage is configured to modify a temperature of a workpiece held by the workpiece stage. 157. The chamber of claim 1, further comprising a workpiece alignment system. 158. The chamber of claim 1, further comprising a detection system configured to detect time of flight of the digital flashes. 159. The chamber of claim 1, further comprising a registration sensor. 160. The chamber of claim 159, wherein the registration sensor is configured to detect emissions incident from a registration mark on the workpiece stage or on a workpiece held by the workpiece stage. 161. The chamber of claim 160, wherein the emissions comprise electrons. 162. The chamber of claim 160, wherein the emissions comprise secondary ions. 163. The chamber of claim 159, wherein the registration mark includes a moiré pattern. 164. The chamber of claim 159, wherein the registration sensor is configured to detect emissions from a plurality of registration marks on the workpiece stage or on a workpiece held by the workpiece stage. 165. The chamber of claim 159, wherein the registration sensor is configured to detect and process differential signals from a plurality of registration marks on the workpiece stage or on a workpiece held by the workpiece stage. 166. The chamber of claim 159, wherein the registration sensor is configured to globally detect emissions across the workpiece stage or across a workpiece held by the workpiece stage. 167. The chamber of claim 159, wherein the registration sensor is configured to locally detect emissions across portions of the workpiece stage or across a workpiece held by the workpiece stage. 168. The chamber of claim 159, wherein the registration sensor is configured to detect backscatter emissions from a workpiece stage or from a workpiece held by the workpiece stage. 169. The chamber of claim 159, wherein the registration sensor is configured to detect emissions from an edge of the workpiece stage or an edge of a workpiece held by the workpiece stage. 170. The chamber of claim 1, further comprising a height control system configured to measure a height of the workpiece stage. 171. The chamber of claim 170, wherein the height control system comprises a laser and a detector, the detector configured to receive light emitted from the laser and reflected by the workpiece stage or by a workpiece held by the workpiece stage. 172. The chamber of claim 170, wherein the height control system comprises a laser and a plurality of detectors, the plurality of detectors configured to receive light emitted from the lasers and reflected by the workpiece stage or by a workpiece held by the workpiece stage. 173. The chamber of claim 170, wherein the height control system is configured to compensate for variation in the measured height of the workpiece stage by adjusting an elevation of the workpiece stage. 174. The chamber of claim 173, wherein the height control system is configured to compensate for height variations of less than about 1 micron. 175. The chamber of claim 171, wherein the height control system comprises electrostatic clamps and piezoelectric devices. 176. The chamber of claim 1, wherein the workpiece stage is configured to hold a semiconductor wafer. 177. The chamber of claim 1, wherein the workpiece stage is configured to hold a semiconductor device. 178. The chamber of claim 1, wherein the workpiece stage is configured to hold a photomask. 179. The chamber of claim 1, wherein the workpiece stage is configured to hold a digital media disk. 180. A workpiece processing apparatus comprising:a loadlock chamber;the chamber of claim 1; anda processing chamber. 181. The apparatus of claim 180, wherein the processing chamber is selected from a group consisting of deposition, etch, and rapid thermal anneal chambers. 182. The apparatus of claim 180, comprising a plurality of processing chambers. 183. The apparatus of claim 180, further comprising a workpiece stage control system configured to detect positional accuracy information. 184. The apparatus of claim 183, further comprising a feedback system configured to adjust a parameter the digital flashes based on said detected positional accuracy information. 185. The apparatus of claim 183, wherein the workpiece stage control system is configured to correct for at least one of coma distortion, digital beam astigmatism, digital beam pure distortion, chromatic aberration, spherical aberration, and field curvature. 186. The apparatus of claim 180, further comprising an integrated pattern data and beam deflection correction system. 187. The apparatus of claim 180, further comprising a transport module configured to move workpieces within the apparatus. 188. The apparatus of claim 180, further comprising a workpiece prealigner. 189. The apparatus of claim 188, wherein the workpiece prealigner is configured to determine overlay parameters of workpiece alignment features. 190. The apparatus of claim 189, wherein the overlay parameters comprise x and y offset. 191. The apparatus of claim 189, wherein the overlay parameters comprise rotation. 192. The apparatus of claim 180, further comprising a particle detector. 193. The apparatus of claim 180, further comprising a temperature quenching station. 194. The apparatus of claim 180, further comprising system for indexing and identifying workpieces. 195. The apparatus of claim 180, further comprising a metrology station. 196. The apparatus of claim 195, wherein the metrology station comprises a secondary ion mass spectrometer (SIMS). 197. The apparatus of claim 195, wherein the metrology station comprises a scanning electron microscope (SEM). 198. The apparatus of claim 195, wherein the metrology station comprises a two-dimensional laser scanning imager. 199. The apparatus of claim 195, wherein the metrology station comprises a three-dimensional imaging laser radar (LADAR). 200. The apparatus of claim 195, wherein the metrology station comprises a thermal imager. 201. The apparatus of claim 195, wherein the metrology station comprises a millimeter wave imager. 202. The apparatus of claim 195, wherein the metrology station comprises a workpiece imager. 203. The apparatus of claim 195, wherein the metrology station comprises a camera. 204. The apparatus of claim 195, wherein the metrology station comprises energy dispersive spectrometry (EDS). 205. The apparatus of claim 195, wherein the metrology station comprises wavelength dispersive spectrometry (WDS). 206. The apparatus of claim 180, wherein the processing chamber comprises a temperature control system including automated control hardware and software. 207. The apparatus of claim 180, wherein the processing chamber comprises a pressure control system including automated control hardware and software. 208. The apparatus of claim 207, wherein the pressure control system is configured to control partial pressures of gas species in the processing chamber. 209. The apparatus of claim 180, wherein the loadlock chamber is configured to accept a front opening unified pod (FOUP). 210. The apparatus of claim 180, wherein the loadlock chamber is configured to simultaneously handle a plurality of workpieces. 211. The apparatus of claim 180, wherein the chamber is mounted on a vibration isolation system. 212. The apparatus of claim 211, wherein the vibration isolation system includes active damping. 213. The apparatus of claim 211, wherein the vibration isolation system includes passive damping. 214. The apparatus of claim 211, wherein the vibration isolation system includes active damping and passive damping. 215. The apparatus of claim 180, configured to process a workpiece from an initial state to a substantially finished state without removing said workpiece from the apparatus. 216. The apparatus of claim 215, wherein the initial state comprises a bare substrate. 217. The apparatus of claim 215, wherein the initial state comprises a substrate including a deposited oxide layer. 218. The apparatus of claim 215, wherein the initial state comprises a substrate including a resist material upon the surface. 219. The apparatus of claim 215, wherein the initial state comprises a substrate including a combination resist material upon the surface and deposited oxide layer. 220. The apparatus of claim 215, wherein the initial state comprises a substrate prepared with interactive coating of the workpiece surface. 221. The apparatus of claim 215, wherein the initial state comprises a substrate prepared with non-interactive coating of the workpiece surface. 222. The apparatus of claim 215, wherein the substantially finished state comprises one or more process layers. 223. The apparatus of claim 215, wherein the substantially finished state comprises one or more critical layers. 224. The apparatus of claim 215, wherein the substantially finished state comprises partial exposure of said workpiece to the beam. 225. The apparatus of claim 215, wherein the substantially finished state comprises one or more devices. 226. The apparatus of claim 215, wherein the substantially finished state comprises one or more devices ready for passivation. 227. The apparatus of claim 215, wherein the substantially finished state comprises a fully completed workpiece requiring no further process layers. 228. The apparatus of claim 215, wherein the initial state comprises a semiconductor wafer prior to forming layers that are part of a particular device. 229. The apparatus of claim 215, wherein the final state comprises said semiconductor wafer after forming said particular device. 230. A chamber for exposing a workpiece to charged particles, the chamber comprising:a charged particle source configured to generate a stream of charged particles;a collimator configured to collimate and direct the stream of charged particles from the charged particle source along an axis of propagation;a beam digitizer downstream of the collimator, the beam digitizer configured to create a digital beam comprising groups of at least one charged particle by adjusting longitudinal spacing between the charged particles along the axis;a deflector downstream of the beam digitizer configured to deflect the groups of charged particles; anda workpiece stage downstream of the deflector, the workpiece stage configured to hold the workpiece. |
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041938437 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows a storage pool 1 for fuel assemblies. A separating gate 2 is removed so that the same water level exists in the fuel storage pool 1 and in the flooding canal 3. The removal and replacement of fuel assemblies is performed in a known way using a telescoping grapple tool 4 suspended from a hoist 5 which removes the fuel assembly 6 vertically from the reactor vessel 7. The test apparatus 9 is attached to the mast 8 through support 10 of the hoist making it possible to identify defective fuel rods while the fuel assembly 6 is still hanging over the reactor vessel, during the transportation to the storage pool 1 or when the fuel assemblies 6a are stored in the storage pool. The test apparatus consists of a main support 10 which is attached to the mast 8. Vertical support members 11 are attached to the main support 10 and a support plate 12 is connected at the bottom of the members 11. A carriage 13 with comb-like arranged fingers slides along rails of the support plate 12. Ultrasonic transducer heads are attached to the free ends of the fingers. The members 11 can be moved vertically by means of a telescoping device. The main support 10 can be rotated around the mast 8 to permit access to a fuel assembly from all sides. The carriage 13 with the ultrasonic transducer heads is moved in the direction of the arrow 14 in the region of the lower fuel rod ends and perpendicularly to the fuel rod axis. From FIG. 2, which schematically represents a single fuel rod 24, it can be seen that above and below the stack of fuel pellets 15 there is an upper empty space 16 and lower empty space 17, respectively. A spring 18 is located to bear on the end caps of the fuel rods and serves to support the fuel pellets and maintain them in proper position. When the reactor is shutdown, the water which leaked into the defective fuel rods collects in the empty space 17. Therefore, the testing will be performed mainly in the region of the lower fuel rod ends. FIG. 3 shows the testing device in contact with the first row of fuel rods of the fuel assembly. A fuel assembly with 5.times.5 fuel rods is shown for simplicity. The device can also be used for fuel assemblies with 17.times.17 fuel rods. The carriage 13 is provided with fingers 26 which at their free ends (26a) have ultrasonic transducer heads 22 attached by adhesive means. The distance between fuel rods of a pressurized water reactor is about 2 to 3 mm. The fingers are therefore made of thin metal strips of a cross section of 1.times.20 mm. The ultrasonic transducing head is a vibrating crystal of about 1 mm thickness obtainable from the pertinent industry. A small wavy strip spring 25 is attached, e.g. by rivets or spot welding, in a depression 27 of the finger 26 opposite of the ultrasonic head. During the insertion of the device between the fuel rods 24 the hump 19 of the spring is depressed and the fuel rod is locked between the humps 19 and 20 of the spring, so that the ultrasonic test head 22 is pressed against the opposite fuel rod. The water surrounding the fuel assemblies serves as the coupling medium between the ultrasonic transducer and the fuel rod. The ultrasonic head 22 which induces vibration is connected through electronic wiring to instrumentation and the resonance of the ultrasonic waves is conducted to an instrument (not shown) for evaluation. The signals from the various fuel rods can be compared which permits a determination of the amount of water that leaked-in or the size of the crack. The testing of a single row of fuel rods requires about 20 seconds, so that a fuel assembly consisting of 17.times.17 rods can be tested in about 6 minutes. Since electronic transducers are used the number of fuel rods in the assembly is of minor significance. A special spring 23 is used to provide the required pressure on the outer fuel elements in each row. FIG. 4 shows a single finger 26 with the ultrasonic head 22. This figure represents a section along line 4--4 of FIG. 3. It can be seen that the ultrasonic head 22 is arranged at the free end 26a of the finger. The distance "a" between the head and the beginning of the comb-like fingers is long enough to permit insertion to the last row of fuel rods 28 (FIG. 3). FIG. 5 shows an example of the guide rails for the carriage 13 on the support plate 12 which is in the form of flat slides. The movement of the carriage in the direction of the arrows 14 can be accomplished by hydraulic, pneumatic or electric drive means (not shown). A special design of the fingers and the ultrasonic transducer heads is shown on FIG. 6. The ultrasonic head 22 is inserted into the body of the carriage 13a. The ultrasonic impulses are transmitted to the tested fuel rod through an ultrasonic wave conductor of low absorption losses such as aluminum, which is used to form the finger 26. For this purpose it is necessary to provide the free end of the finger with a forty-five degree end surface 21 to obtain a ninety degree deflection of the sound waves towards the fuel element. In this arrangement, the known good sound wave conductance of aluminum is advantageously utilized. The method of this invention is not to be limited to water cooled reactors, but may be applied to reactors cooled by any fluid. |
043269206 | claims | 1. In a pool type nuclear reactor including a number of reactor components located within a vertically extending cavity which is located under a reactor deck arrangement and which is defined by a circumferential cavity wall assembly, the improvement comprising a reactor vessel separated and distinct from said deck arrangement and said cavity wall assembly for containing said reactor components, said vessel including a main body located within said cavity and an upper circumferential rim formed by a vertically exteded support flange located and interlocked between said deck arrangement and an upper section of said wall assembly, whereby to form a support means for said vessel body within said cavity and said deck arrangement above said vessel body, said support means including a first downwardly directed, circumferential shoulder extending around and forming part of said support flange, a first upwardly directed, circumferential shoulder extending around and forming part of said cavity wall assembly said first upwardly directed shoulder cooperating with said first downwardly directed shoulder for supporting said vessel within said cavity, a second downwardly directed circumferential shoulder extending around and forming part of said deck arrangement, a second upwardly directed, circumferential shoulder extending around and forming part of said support flange on the opposite side of and above said first downwardly directed shoulder but below the top of said flange, said second upwardly directed shoulder cooperating with said second downwardly directed shoulder for supporting said deck arrangement over the main body of said vessel and means for sealing between said support flange and both said deck arrangement and said wall assembly, said sealing means displaying at least a limited degree of compliance whereby to accomodate thermal displacement. 2. The improvement according to claim 1 wherein the main body of said reactor vessel and said support flange are both constructed of stainless steel. 3. The improvement according to claim 2 wherein said main body and support flange are formed as a single unit. 4. The improvement according to claim 4 including a slide plate located between at least one of said pair of cooperating shoulders in order to allow for differential thermal displacement between said one pair of shoulders. 5. The improvement according to claim 4 including a slide plate located between each pair of said cooperating shoulders. 6. The improvement according to claim 1 including shear pins extending between said deck arrangement and flange support and between the latter and said wall assembly for resisting seismic forces. 7. The improvement according to claim 6 wherein the shear pins are aligned with said arrangement, flange, and assembly so as not to restrain differential thermal expansion between these latter components. 8. The improvement according to claim 1 including a compliant ring located between said first shoulders in order to allow for differential thermal displacement. 9. In a pool type nuclear reactor including a number of reactor components located within a vertically extending cavity which is located under a reactor deck arrangement and which is defined by a circumferential cavity wall assembly, the improvement comprising a reactor vessel separate and distinct from said deck arrangement and said cavity wall assembly for containing said reactor components, said vessel including a stainless steel main body located within said cavity, and an upper circumferential rim forming a support flange located between said deck arrangement and an upper section of said cavity wall assembly, a first downwardly directed, circumferential shoulder extending around and forming part of said support flange, a first upwardly directed, circumferential shoulder extending around and forming part of said wall assembly, said upwardly directed shoulder cooperating with said downwardly directed shoulder for supporting said vessel within said cavity, a second downwardly directed, circumferential shoulder extending around and forming part of said deck arrangement, a second upwardly directed, circumferential shoulder extending around and forming part of said support flange on the opposite side of and above said first downwardly directed shoulder but below the top of said flange, said second upwardly directed shoulder cooperating with said second downwardly directed shoulder cooperating with said second downwardly directed shoulder for supporting said deck arrangement over said vessel body, means for sealing between said support flange and both said deck arrangement and said wall assembly, said sealing means displaying at least a limited degree of compliance in order to accomodate thermal displacement, and means located between said first and second shoulders in order to allow for differential thermal displacement between said shoulders, said last mentioned means including a slide plate between said first shoulders and a slide plate between said second shoulders. |
056429550 | claims | 1. A strongback comprising: a first end member having a mating portion of a coupling, said mating portion of said first end member being generally located at a first axial position along a strongback axis; a second end member having a mating portion of a coupling, said mating portion of said second end member being generally located at a second axial position along said strongback axis, said first and second axial positions being mutually separated by at least a first predetermined distance; a first rigid linear member having a longitudinal axis generally parallel to said strongback axis and having first and second ends; a second rigid linear member having a longitudinal axis generally parallel to said strongback axis and having first and second ends, said second rigid linear member being offset from said first rigid linear member by a second predetermined distance; a third rigid linear member disposed at an oblique angle relative to said strongback axis and having a first end connected to said second end of said first rigid linear member and a second end connected to said first end of said second rigid linear member; a fourth rigid linear member disposed at an oblique angle relative to said strongback axis and having a first end connected to said first end member and a second end connected to said first end of said first rigid linear member; a fifth rigid linear member disposed at an oblique angle relative to said strongback axis and having a first end connected to said second end of said second rigid linear member and a second end connected to said second end member, wherein said first, second and third rigid linear members lie in a plane which is offset from said strongback axis. a cable; a first adaptor connected to an end of said cable; a strongback comprising a first coupling at an upper end, a second coupling at a lower end, a midsection between said first and second couplings, first means for connecting a top portion of said midsection to said first coupling, and second means for connecting a bottom portion of said midsection to said second coupling, said first and second couplings lying along a strongback axis at first and second elevations respectively, said midsection extending between a third elevation below said first elevation and a fourth elevation above said second elevation and below said third elevation, said midsection being offset from said strongback axis along its entire extent, said first coupling being connected to said first adaptor such that said cable supports said strongback; and a second adaptor connected to said second coupling of said strongback, said second adaptor having means for connecting to and supporting said component of said nuclear reactor. a first rigid linear member having a longitudinal axis generally parallel to said strongback axis and having first and second ends; a second rigid linear member having a longitudinal axis generally parallel to said strongback axis and having first and second ends, said second rigid linear member being offset from said first rigid linear member; and a third rigid linear member disposed at an oblique angle relative to said strongback axis and having a first end connected to said second end of said first rigid linear member and a second end connected to said first end of said second rigid linear member, wherein said first, second and third rigid linear members lie in a plane which is offset from said strongback axis. a first adaptor for connecting to an end of the cable; a strongback comprising a first coupling at an upper end, a second coupling at a lower end, a midsection between said first and second couplings, first means for connecting a top portion of said midsection to said first coupling, and second means for connecting a bottom portion of said midsection to said second coupling, said first and second couplings lying along a strongback axis at first and second elevations respectively, said midsection extending between a third elevation below said first elevation and a fourth elevation above said second elevation and below said third elevation, said midsection being offset from said strongback axis along its entire extent, said first coupling being connected to said first adaptor such that said first adaptor supports said strongback; and a second adaptor connected to said second coupling of said strongback such that said strongback supports said second adaptor, said second adaptor having means for connecting and supporting said component of said nuclear reactor, wherein said strongback midsection comprises: 2. The strongback as defined in claim 1, wherein said mating portion of at least one of said first and second end members is in the form of a clevis. 3. The strongback as defined in claim 1, wherein each connection between said members is a welded joint. 4. The strongback as defined in claim 3, wherein each said welded joint is spanned by a reinforcement rib which is welded to both of said members connected by said respective welded joint. 5. The strongback as defined in claim 1, wherein each of said rigid linear members is a metal tube having a square cross section. 6. An apparatus for suspending a component of a nuclear reactor inside a reactor pressure vessel, comprising: 7. The apparatus as defined in claim 6, wherein said strongback midsection comprises: 8. An apparatus for suspending a component of a nuclear reactor from a cable, comprising: 9. The apparatus as defined in claim 8, wherein said second coupling comprises a clevis. 10. The apparatus as defined in claim 8, wherein said first coupling comprises an apertured plate. 11. The apparatus as defined in claim 8, wherein said first and third rigid linear members are connected by a first welded joint, and said second and third rigid linear members are connected by a second welded joint. 12. The apparatus as defined in claim 11, wherein said first welded joint is spanned by a first reinforcement rib which is welded to said first and third rigid linear members, and said second welded joint is spanned by a second reinforcement rib which is welded to said second and third rigid linear members. 13. The apparatus as defined in claim 8, wherein each of said first, second and third rigid linear members is a metal tube having a square cross section. |
description | The following disclosure is based on and claims the benefit, under 35 U.S.C. §119(a), of German Patent Application No. 10 2013 222 140.1, filed on Oct. 30, 2013, which German Patent Application is incorporated in its entirety into the present application by reference. The invention relates to a reflective optical element, in particular for a microlithographic projection exposure apparatus or for a mask inspection apparatus. Microlithography is used for producing microstructured components, such as, for example, integrated circuits or LCDs. The microlithography process is carried out in a so-called projection exposure apparatus having an illumination device and a projection lens. The image of a mask (=reticle) illuminated with the illumination device is in this case projected by the projection lens onto a substrate (e.g. a silicon wafer) coated with a light-sensitive layer (photoresist) and arranged in the image plane of the projection lens, in order to transfer the mask structure to the light-sensitive coating of the substrate. Mask inspection apparatuses are used for inspecting reticles for microlithographic projection exposure apparatuses. In projection lenses or inspection lenses designed for the EUV range, i.e. at wavelengths of e.g. approximately 13 nm or approximately 7 nm, owing to the lack of availability of suitable light-transmissive refractive materials, reflective optical elements are used as optical components for the imaging process. One problem that occurs in practice is that reflective optical elements designed for operation in the EUV, in particular owing to the absorption of the radiation emitted by the EUV light source, experience heating and an accompanying thermal expansion or deformation, which can in turn result in an impairment of the imaging properties of the optical system. This is the case particularly if use is made of illumination settings having comparatively small illumination poles (e.g. in dipole or quadrupole illumination settings), in which the element heating or deformation varies greatly across the optically effective surface of the reflective optical element. Transferring solution approaches known for VUV lithography systems (having an operating wavelength e.g. of approximately 200 nm or approximately 160 nm) for overcoming the above-described problem of element heating to EUV systems has proven difficult. This is so in part because the number of optically effective surfaces available for active deformation compensation is, relative to VUV systems, greatly limited owing to the comparatively smaller number of optical elements or mirrors that are used in EUV lithography. (The number of elements or mirrors is kept small in order to avoid excessively high light losses on account of the necessary reflections). In order to overcome the above-described problem of element heating in EUV systems it is known, in particular, to use additional devices for realizing rigid-body movements and/or temperature changes in the region of the optically effective surface of the reflective optical elements designed for operation in the EUV. Such solutions, however, increase the complexity of the systems. It is an object of the present invention to provide a reflective optical element, in particular for a microlithographic projection exposure apparatus or for a mask inspection apparatus, which, in conjunction with comparatively little structural complexity, enables an effective avoidance or at least a reduction of thermal deformations or accompanying impairments of the imaging behaviour. A reflective optical element, in particular for a microlithographic projection exposure apparatus or for a mask inspection apparatus and addressing the above object, comprises, according to one formulation: an optically effective surface, an element substrate, a reflection layer system, and at least one deformation reduction layer which, upon the optically effective surface being irradiated with electromagnetic radiation, reduces a maximum deformation level of the reflection layer system in comparison with an analogous construction without the deformation reduction layer. The inventors have recognized that the maximum deformation level can be reduced by virtue of the fact that, in the case of the reflective optical element according to the invention, a deformation reduction layer is taken into account or incorporated from the outset and has the effect that an undesired thermally governed deformation of the reflective optical element toward the vacuum as far as possible does not actually occur in the first place. This deformation level can be defined here in each case in the direction of the normal vector of the element surface. If even pronounced aspheres are present, the deformation level can preferably be defined by the element surface being approximated by a spherical radius (as “basic radius”), onto which the respective asphere is modulated, the deformation level then being defined as deformation in the direction of the normal vector to the spherical surface described by said radius. In other words, the present invention includes, in particular, the concept of suitably incorporating a deformation reduction layer so as already to prevent thermally governed deformations from the outset and thus to render unnecessary or at least considerably simplify an active deformation compensation through rigid-body movements and/or temperature changes in the region of the optically effective surface. In accordance with one embodiment, the reflection layer system has at least one layer composed of a first material having a first coefficient of thermal expansion, and the at least one deformation reduction layer has a second material having a second coefficient of thermal expansion, wherein the first and second coefficient of thermal expansions have opposite signs. In accordance with this approach, as will be explained in even greater detail below, a thermally governed volume expansion of the reflection layer system during operation of the reflective optical element or of the optical system having the latter can be compensated for by an opposite volume contraction of the deformation reduction layer, said volume contraction ideally being of equal size in terms of absolute value, with the consequence that the effective volume change of the arrangement comprising reflection layer system, on the one hand, and deformation reduction layer, on the other hand, that results from the heating of the optically effective surface is reduced to almost zero. As a result, with optimum coordination of reflection layer system, on the one hand, and deformation reduction layer, on the other hand, in particular with regard to the respective layer thicknesses and materials, it is thus possible to achieve to a certain extent an intrinsic and self-consistent deformation compensation in the reflective optical element itself. In accordance with one embodiment, the first material comprises zirconium (Zr), yttrium (Y), molybdenum (Mo), niobium (Nb), silicon (Si), germanium (Ge), rhodium (Rh), ruthenium (Ru), ruthenium dioxide (RuO2) or ruthenium-silicon (RuSi). In accordance with one embodiment, the second material is selected from the group containing ZrMo2O8, ZrW2O8, HfMo2O8, HfW2O8, Zr2 (MoO4)3, Zr2 (WO4)3, Hf2 (MoO4)3, Hf2 (WO4)3, SCF3r ZnC2N2, ZnF2, Y2W3O12 and BiNiO3. Preferably, the second material used is an isotropic material such as e.g. a cubically crystalline or amorphous material (for avoiding mechanical stresses and resultant microcracks) having a comparatively high negative coefficient of thermal expansion and good thermodynamic stability. In accordance with one embodiment, the effective volume change ΔVeff of the arrangement comprising reflection layer system and deformation reduction layer that results from a heating of the optically effective surface by a predetermined temperature difference is a maximum of 90%, in particular a maximum of 50%, more particularly a maximum of 10%, of the volume change V1 of the reflection layer system that results from said heating. In this case, this condition can be met in particular for a heating of the optically effective surface by a temperature difference of at least 1K, in particular at least 5K, more particularly at least 10K. In accordance with one embodiment, the at least one deformation reduction layer has a heat distribution layer having an increased thermal conductivity in comparison with the element substrate. This heat distribution layer can have in particular a thermal conductivity of at least 100 W/mK. The heat distribution layer can furthermore comprise in particular at least one material selected from the group containing graphite, aluminium (Al), silver (Ag), gold (Au), copper (Cu) and ZrW2O8. What can be achieved in accordance with this approach, as will be explained in even greater detail below, is that the heat propagating within the reflective optical element upon electromagnetic radiation impinging on the optically effective surface is distributed better in a lateral direction (i.e. direction perpendicular to the light propagation direction or optical system axis), that is to say that a corresponding temperature distribution within the reflective optical element is “spread” with the consequence that undesired pronounced local deformations in particular in the region of the element substrate—which generally has only poor thermal conductivity—are avoided or at least greatly reduced. Particularly in the case of illumination settings having relatively small illumination poles (e.g. in dipole or quadrupole illumination settings), the heat distribution or deformation can thus be distributed uniformly over the optically effective surface of the reflective optical element before entry into the element substrate. In accordance with one embodiment, a heat insulation layer for delaying the entry of heat into the element substrate is arranged between the reflection layer system and the deformation reduction layer. This heat insulation layer can comprise quartz, in particular. In accordance with one embodiment, a further intermediate layer for avoiding the transfer of surface roughnesses to the reflection layer system is furthermore arranged between the reflection layer system and the deformation reduction layer. This intermediate layer can comprise quartz, in particular. In accordance with one embodiment, the reflective optical element is designed for an operating wavelength of less than 30 nm, in particular less than 15 nm. However, the invention is not restricted thereto, such that in further applications the invention can also be realized advantageously in an optical system having an operating wavelength in the VUV range (e.g. of less than 200 nm or less than 160 nm). The reflective optical element according to the invention can be a mirror, in particular a mirror for a microlithographic projection exposure apparatus or a mirror for a mask inspection apparatus. Furthermore, the reflective optical element according to the invention can also be a reticle for a microlithographic projection exposure apparatus. The invention furthermore relates to an optical system of a microlithographic projection exposure apparatus, in particular an illumination device or a projection lens, an optical system of a mask inspection apparatus, and also a microlithographic projection exposure apparatus and a mask inspection apparatus comprising at least one reflective optical element having the features described above. Further configurations of the invention can be gathered from the description and the dependent claims. FIG. 1 shows a schematic illustration for explaining the construction of a reflective optical element according to the invention in a first embodiment of the invention. The reflective optical element 10 comprises in particular an element substrate 12 produced from an arbitrary suitable (mirror) substrate material. Suitable element substrate materials are e.g. titanium dioxide (TiO2)-doped quartz glass, wherein merely by way of an exemplary embodiment and without the invention being restricted thereto, the materials sold under the trademarks ULE® (from Corning Inc.) or Zerodur® (from Schott AG) can be used. Furthermore, the reflective optical element 10 has a reflection layer system 14 in a manner known per se in principle, said reflective optical element, in the embodiment illustrated, comprising merely by way of example a molybdenum-silicon (Mo—Si) layer stack. Without the invention being restricted to specific configurations of said layer stack, a merely exemplary suitable construction can comprise 50 plies or layer packets of a layer system comprising molybdenum (Mo) layers having a layer thickness of 2.4 nm in each case and silicon (Si) layers having a layer thickness of 3.3 nm in each case. Optionally, provision can also be made of further functional layers such as, for example, a capping layer (“cap layer”), which can consist e.g. of Ru, Rh, SiC, C, Ir, Mo2C, Y2O3 or Si3N4, a substrate protection layer (SPL), which can consist e.g. of compounds comprising at least one of the elements Pt, Cu, Co, Sn, Ni and Ag, and/or diffusion barriers, which can consist for example of C, B4C, SixNy, SiC, Mo2C, MoSi2, Y5Si3 or Nb4Si. The reflective optical element 10 can be, in particular, a reflective optical element designed for operation in the EUV or a mirror of an optical system, in particular of the projection lens or of the illumination device of a microlithographic projection exposure apparatus or of the inspection lens of a mask inspection apparatus. The impingement of electromagnetic EUV radiation (indicated by an arrow in FIG. 1) on the optically effective surface 11 of the reflective optical element 10 during operation of the optical system has the consequence of a volume expansion of the reflection layer system 14 on account of the positive coefficient of thermal expansion of the material of the reflection layer system 14 (an average coefficient of thermal expansion of approximately 3.61.10−6 K−1 results in the case of an MoSi layer stack). Depending of the intensity distribution of the incident electromagnetic radiation (that is to say, in particular depending on the set illumination setting in the case of a near-pupil reflective optical element), said volume expansion proceeds inhomogeneously over the optically effective surface 11, which is illustrated in a greatly simplified and schematic manner in FIG. 2A. In order, then, at least partly to reduce the deformation of the reflective optical element 10 overall and in particular of the optically effective surface 11 thereof, this deformation being caused by the irradiation of the optically effective surface 11 with electromagnetic radiation, the reflective optical element 10 has a deformation reduction layer 15, which, in accordance with FIGS. 1-2 is situated on that side of the reflection layer system 14 which faces away from the optically effective surface 11. In the exemplary embodiment in FIG. 1, said deformation reduction layer 15 is constituted such that the deformation of the deformation reduction layer 15 caused by the impingement of electromagnetic radiation or the accompanying temperature increase proceeds precisely oppositely to the deformation of the reflection layer system 14. For this purpose, the deformation reduction layer 15 has a coefficient of thermal expansion having an opposite sign in comparison with the material of the reflection layer system 14, that is to say a negative coefficient of thermal expansion in the exemplary embodiment. Merely by way of example (and without the invention being restricted thereto), the deformation reduction layer can be produced e.g. from ZrW2O8 having a thickness of 118 nm (wherein ZrW2O8 is a crystal material having a cubic crystal structure and a coefficient of thermal expansion of −8.7.10−6 K−1 in the temperature range of 20K to 430K). In the abovementioned example (50 plies or layer packets of a layer system comprising Mo layers having a layer thickness of 2.4 nm and Si layers having a layer thickness of 3.3 nm), the resulting total thicknesses of 120 nm molybdenum (Mo) and 165 nm silicon (Si) can be compensated for by a 118 nm thick layer of ZrW2O8 with regard to the thermally governed deformation. In further embodiments, a layer of ZrW2O8 having a larger thickness can also be used in order additionally to compensate for the thermal expansion in the element substrate material (e.g. TiO2-doped quartz glass). Further suitable materials having a negative coefficient of thermal expansion are listed in Table 1: TABLE 1Coefficient of thermalMaterialexpansion [10−6 K−1]PbTiO3−3.3Sc2W3O12−2.2Y2W3O12−4.2Lu2W3O12−6.8NbOPO4−3.7ZrV2O7−7.1ZrW2O8−8.7ZrMo2O8−5.0K5Zr(PO4)3−0.5KZr2(PO4)3−1.7Zn(CN)2−18.1 Since the coefficient of thermal expansion itself is a function of temperature, in the case of an isotropic material (in particular having a negative coefficient of thermal expansion), the expansion or contraction can be assumed to be linear with temperature only for a limited temperature range. In order likewise to correct quadratic portions and higher-order portions, a plurality of suitable materials can also be mixed or stacked one above another in the layer construction. Furthermore, a deformation reduction layer 15 e.g. composed of ZrW2O8 (which e.g. with a thickness of 150 nm absorbs approximately 94% of EUV light having a wavelength of 13.5 nm) is preferably arranged as the bottommost layer (apart from a heat insulation layer and/or a substrate protection layer possibly present, as described further below) in the layer stack relative to the optically effective surface. In practice, materials, layer thicknesses and layer sequences of the reflection layer system 14, on the one hand, and of the deformation reduction layer 15, on the other hand, can in each case be coordinated with one another or optimized such that an at least almost complete mutual compensation of the temperature-governed deformation is obtained. This effect is illustrated only schematically and in a greatly simplified manner in FIG. 2B. The volume expansion of the reflection layer system 14, which takes place primarily in the region of high radiation intensities or illumination poles, is compensated for in accordance with FIG. 2B by an opposite volume contraction of the deformation reduction layer 15, said volume contraction ideally being of the same size in terms of absolute value, with the consequence that the effective volume change of the arrangement 13 comprising reflection layer system 14 and deformation reduction layer 15 that results from the heating of the optically effective surface 11 is reduced to almost zero. Furthermore, as already mentioned, it is also possible at least partly to compensate for the thermal expansion in the element substrate material (e.g. in the TiO2-doped quartz glass). As a suitable criterion for quantitatively describing the reduction according to the invention of the temperature-governed deformation, it is possible to specify in particular an upper limit for the effective volume change ΔVeff of the arrangement 13 comprising reflection layer system 14 and deformation reduction layer 15 that results from the heating by a predefined temperature difference. In this case, what can be achieved in particular by suitable coordination of materials and layer thicknesses of the reflection layer system 14, on the one hand, and of the deformation reduction layer 15, on the other hand, is that said resulting effective volume change ΔVeff is a maximum of 90%, in particular a maximum of 50%, more particularly a maximum of 10%, of the volume change of the reflection layer system 14 itself that results from said heating. In accordance with FIG. 2C, in the case of a reflective optical element without a deformation reduction layer, a layer thickness variation of, for example, 2 nm or of 4 nm or of 5 nm (x-axis) is brought about owing to element heating. An unwanted local phase change of the wavefront of approximately 4 nm or of approximately 8 nm or of approximately 10 nm (y-axis) is brought about as a result. In the case of a reflective optical element having a deformation reduction layer designed according to the invention, no or only a negligible wavefront change takes place. A variation of the reflectivity takes place both in the case of a reflective optical element without a deformation reduction layer and in the case of a reflective optical element with a deformation reduction layer. FIG. 2D shows by way of example for an EUV narrowband layer stack in the case of a layer thickness variation of 2 nm or of 4 nm or of 5 nm an unwanted local reflection reduction by approximately 0.1% or 1.6% or 2.5%, respectively. In the case of an EUV broadband layer stack, the contributions to an unwanted local layer thickness variation may be smaller. Currently expected local layer thickness variations caused by element heating are less than 5 nm, in particular less than 2 nm. In accordance with FIG. 1, a heat insulation layer 16 for delaying the entry of heat into the element substrate 12 is furthermore also arranged between the arrangement 13 comprising reflection layer system 14 and deformation reduction layer 15 and the element substrate 12, wherein said heat insulation layer 16 e.g. composed of quartz having a coefficient of thermal expansion of approximately 0.5·10−6 K−1 and a thermal conductivity of approximately 1.38 W/mK (which is comparatively low for instance relative to copper with a value of 400 W/mK, or silicon with a value of 150 W/mK) can be vapour-deposited. A further possible configuration of the deformation reduction layer according to the invention is explained below with reference to FIGS. 3A-B, wherein in comparison with FIG. 1 analogous or substantially functionally identical components are designated by reference numerals increased by 20. The reflective optical element 30 illustrated schematically in FIGS. 3A-B has, analogously to the reflective optical element 10 from FIG. 1, a deformation reduction layer 35 on that side of the reflection layer system 34 which faces away from the optically effective surface 31. In contrast to the embodiment from FIG. 1, however, the mode of action of the deformation reduction layer 35 in accordance with FIGS. 3A-B is based on the fact that the heat propagating within the reflective optical element 30 upon electromagnetic radiation impinging on the optically effective surface 31 is distributed better in particular in a lateral direction (i.e. direction perpendicular to the light propagation direction or optical system axis) (that is to say that a corresponding temperature distribution within the reflective optical element 30 is “spread”). This effect of a deformation reduction layer 35 configured as a “heat distribution layer” in the above sense is indicated once again merely in a greatly simplified manner in FIG. 3B and has the consequence that undesired pronounced local deformations in particular in the region of the element substrate 32 (which otherwise are caused by the comparatively poor thermal conductivity and a comparatively great temperature dependence of the coefficient of thermal expansion of the element substrate material such as e.g. ULE®) are avoided or at least greatly reduced. Exemplary suitable materials for the heat distribution layer or deformation reduction layer 35 are for example graphite (having a thermal conductivity of approximately 130 W/mK, and a coefficient of thermal expansion of approximately 6.5·10−6 K−1), or copper (having a thermal conductivity of approximately 390 W/mK and a coefficient of thermal expansion of approximately 16·10−6 K−1). The approach described above with reference to FIGS. 3A-B, thus takes account of the fact that particularly upon the setting of illumination settings having comparatively small illumination poles or greatly inhomogeneous intensity distributions on the optically effective surface of an e.g. near-pupil reflective optical element, comparatively high temperature gradients occur in the element substrate material since said element substrate material typically has only a low thermal conductivity and, moreover, a coefficient of thermal expansion greatly dependent on the respective temperature range. Resultant local deformations in the region of the element substrate in turn typically result in medium- to high-frequency optical aberrations in the lithographic process, which can be significantly reduced by the above-described approach of lateral heat distribution (or “spreading” of the location-dependent temperature distribution within the reflective optical element). In accordance with FIG. 4, a smoothing layer as a further intermediate layer 47 for avoiding the transfer of surface roughnesses to the reflection layer system 44 can furthermore be arranged between the reflection layer system 44 and the deformation reduction layer 45. This smoothing layer can comprise quartz and/or silicon, for example. It may be necessary that the smoothing layer must be smoothed in a separate work operation after application. This smoothing can be effected for example by mechanical processing and/or by ion beam figuring. In further embodiments of the invention, the above-described approach of using at least one heat distribution layer as deformation reduction layer 35 can also be combined with the approach described above with reference to FIG. 1, as is illustrated merely schematically in FIG. 5. In this regard, both at least one deformation reduction layer 58 acting as heat distribution layer and at least one deformation reduction layer 55 having an opposite coefficient of thermal expansion in comparison with the reflection layer system 54 can also be used for the deformation reduction according to the invention. In further embodiments, a layer stack having an alternating layer sequence composed of deformation reduction layers 55 and 58 can also be provided, wherein it is possible to produce the deformation reduction layer 55 as heat distribution layer in the above-described sense e.g. composed of graphite (e.g. having in each case a thickness of approximately 0.1338 μm) and the deformation reduction layer e.g. composed of ZrW2O8 (e.g. having a thickness of 0.1 μm) as layer material having opposite coefficient of thermal expansions in comparison with the material of the reflection layer system. Such a double-ply system can also be repeated multiply in a corresponding layer stack, wherein the abovementioned thicknesses can also be multiplied by an arbitrary factor (e.g. the factor 10 or 0.1). In this case, the thermal conductivity of the ZrW2O8, with a value of approximately 2 W/mK, is not significantly greater than that of ULE® (as exemplary element substrate material), but ZrW2O8, as described above, has a negative coefficient of thermal expansion of approximately −8.7·10−6 K−1. When designing the layer thicknesses of the compensation layers, it should be taken into consideration that a spreading of the local region to be compensated takes place via the heat distribution layers. In a further exemplary embodiment, it is also possible in each case for a ply ZrW2O8 having the thickness d1 to be supplemented by a ply composed of copper (Cu) having the thickness d2, wherein d2=0.5438·d1 holds true, wherein such a double-ply system can likewise be repeated multiply in the layer construction. FIG. 6 shows a schematic illustration of an exemplary projection exposure apparatus which is designed for operation in the EUV and in which the present invention can be realized. In accordance with FIG. 6, an illumination device in a projection exposure apparatus 600 designed for EUV has a field facet mirror 603 and a pupil facet mirror 604. The light from a light source unit comprising a plasma light source 601 and a collector mirror 602 is directed onto the field facet mirror 603. A first telescope mirror 605 and a second telescope mirror 606 are arranged in the light path downstream of the pupil facet mirror 604. A deflection mirror 607 is arranged downstream in the light path and directs the radiation incident on it onto an object field in the object plane of a projection lens comprising six mirrors 651-656. A reflective structure-bearing mask 621 on a mask stage 620 is arranged at the location of the object field and is imaged with the aid of the projection lens into an image plane in which a substrate 661 coated with a light-sensitive layer (photoresist) is situated on a wafer stage 660. Of the mirrors 651-656 of the projection lens, in particular the mirrors 651 and 652—arranged in the initial region of the projection lens relative to the optical beam path—can be configured in the manner according to the invention since the effect achieved according to the invention is then particularly pronounced owing to the summed reflection losses that are still comparatively low at these mirrors 651 and 652 and thus the relatively high light intensities. However, the invention is not restricted to application to the mirrors 651 and 652, such that in principle other mirrors can also be configured in the manner according to the invention. Even though the invention has been described on the basis of specific embodiments, numerous variations and alternative embodiments are apparent to the person skilled in the art, e.g. by combination and/or exchange of features of individual embodiments. Accordingly, for the person skilled in the art, such variations and alternative embodiments are concomitantly encompassed by the present invention, and the scope of the invention is restricted only within the meaning of the appended patent claims and the equivalents thereof. |
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062401556 | claims | 1. A method of performing maintenance on a structural member inside a reactor pressure vessel comprising the steps of: lowering an annular guide rail having a second lug until said guide rail rests on an upper flange of a core shroud provided inside a reactor pressure vessel; detachably mounting said second lug of said guide rail to a first lug, mounted on said core shroud; mounting a turntable on said guide rail so as to be movable on said guide rail; mounting a discharging nozzle moving apparatus on said turntable, said discharging nozzle moving apparatus having a multi-joint arm which can insert a discharging nozzle into a narrow space between said core shroud and an upper core grid plate; positioning said discharging nozzle at a position at which compressive remaining stress is added to a surface of said core shroud in a radial direction and an axial direction of said core shroud by said discharging nozzle moving apparatus; and discharging water to add compressive remaining stress to a surface of said core shroud from said discharging nozzle. lowering an annular guide rail having a second lug to an upper flange of a core shroud provided inside said reactor pressure vessel; detachably mounting said second lug of said guide rail to a first lug for mounting a shroud head, mounted on said core shroud; mounting a turntable on said guide rail so as to be movable on said guide rail; mounting a first discharging nozzle moving apparatus and a second discharging nozzle moving apparatus on said turntable, said first discharging nozzle moving apparatus having a multi-joint arm which can insert a first discharging nozzle into a narrow space between said core shroud and an upper core grid plate; positioning said first discharging nozzle at a position at which compressive remaining stress is added to an outer surface of said core shroud in a radial direction and an axial direction of said core shroud in a region between said reactor pressure vessel and said core shroud by said first discharging nozzle moving apparatus; positioning a second discharging nozzle at a position at which compressive remaining stress is added to an inner surface of said core shroud in a radial direction and an axial direction of said core shroud by said second discharging nozzle moving apparatus; discharging water for adding compressive remaining stress to the outer surface of said core shroud from said first discharging nozzle; and discharging water for adding compressive remaining stress to the inner surface of said core shroud from said second discharging nozzle. lowering an annular guide rail onto an upper flange of a core shroud provided inside said reactor pressure vessel; detachably mounting said guide rail to said core shroud; mounting a turntable on said guide rail so as to be movable on said guide rail; mounting a discharging nozzle moving apparatus on said turntable, said discharging nozzle moving apparatus having an arm which can insert a discharging nozzle into a narrow space between said core shroud and an upper core grid plate; positioning said discharging nozzle at a position at which compressive remaining stress is added to a surface of said core shroud in a radial direction and an axial direction of said core shroud by said discharging nozzle moving apparatus; and discharging water to add compressive remaining stress to a surface of said core shroud from said discharging nozzle. lowering an annular guide rail onto an upper flange of a core shroud provided inside said reactor pressure vessel; detachably mounting said guide rail to said core shroud; mounting a turntable on said guide rail so as to be movable on said guide rail; mounting a discharging nozzle moving apparatus on said turntable, said discharging nozzle moving apparatus having a rotating cover arm which can insert a discharging nozzle into a narrow space between said core shroud and an upper core grid plate; positioning said discharging nozzle at a position at which compressive remaining stress is added to a surface of said core shroud in a radial direction and an axial direction of said core shroud by said discharging nozzle moving apparatus; and discharging water to add compressive remaining stress to a surface of said core shroud from said discharging nozzle. 2. A method of performing maintenance on a structural member inside a reactor pressure vessel according to claim 1, wherein said step of lowering the guide rail is done after a plurality of third lugs on said guide rail are engaged with a plurality of guides provided inside said reactor pressure vessel. 3. A method of performing maintenance on a structural member inside a reactor pressure vessel according to claim 1, wherein reactor water inside said reactor pressure vessel is cleaned and utilized as supply water for said discharging nozzle. 4. A method of performing maintenance on a structural member inside a reactor pressure vessel according to claim 1, wherein crud suspended in the reactor water inside said reactor pressure vessel is sucked up and removed from said reactor pressure vessel. 5. A method of performing preventive maintenance on a structural member inside a reactor pressure vessel comprising the steps of: 6. A method of performing preventive maintenance on a structural member inside a reactor pressure vessel according to claim 5, wherein said step of lowering the guide rail is done after a plurality of third lugs on said guide rail are engaged with a plurality of guides provided inside said reactor pressure vessel. 7. A method of performing preventive maintenance on a structural member inside a reactor pressure vessel according to claim 5, wherein reactor water inside said pressure vessel is cleaned and utilized as supply water for said discharging nozzle. 8. A method of performing preventive maintenance on a structural member inside a reactor pressure vessel according to claim 5, wherein crud suspended in the reactor water inside said reactor pressure vessel is sucked up and removed from said reactor pressure vessel. 9. A method of performing maintenance on a structural member inside a reactor pressure vessel comprising the steps of: 10. A method of performing maintenance according to claim 9, wherein said step of lowering said guide rail is effected after performing said step of mounting said turntable thereon and said step of mounting said discharging nozzle moving apparatus, and said guide rail is lowered while being guided by guide rods separated from each other and each fixed to the reactor pressure vessel. 11. A method of performing maintenance on a structural member inside a reactor pressure vessel comprising the steps of: |
052672875 | summary | BACKGROUND OF THE INVENTION This invention relates to nuclear reactor fuel assemblies and in particular to assemblies which are mounted side by side in a reactor core. The fuel rods of each assembly are held by spacer grids between an upper end fitting or top nozzle and a lower end fitting or bottom nozzle. The reactor coolant flows upwardly from holes in the lower end fitting along the fuel rods, and upwardly through holes in the upper end fitting. When the fuel assembly is loaded in a reactor core, an upper core plate over the fuel assembly reacts against fuel assembly holddown spring members on the upper end fitting, to provide a downward force. This force combines with the fuel assembly weight to prevent fuel assembly liftoff from hydraulic forces during operation of the reactor pumps. The holddown spring members are exposed to the high pressure and temperature of the circulating coolant. They experience flexure while accommodating relative movement between the fuel assemblies and the support plate, and they experience some friction at the active surface which is in contact with the core support plate or when stacked with each other. This friction can have two adverse consequences. First, the spring coefficient can be affected if the point of contact of the spring active surface against the support plate or each other, cannot adjust as the spring flexes. Secondly, and perhaps more importantly, friction can produce wear of the upper core plate and spring which can affect the available spring force and the integrity of the spring member itself. SUMMARY OF THE INVENTION It is, accordingly, an object of the present invention to increase the lubricity of the contact surfaces between the spring members on the end fittings of nuclear fuel assemblies, and the core support plates in a nuclear reactor. According to one embodiment, the invention is directed to a method for fabricating an end fitting for a nuclear fuel assembly, wherein the spring members of the end fitting are coated, at least on the active external surface which contact the core support plate, with a lubricity-enhancing material. Suitable materials include a variety of nitrides, and Cr, TiC, CrC, ZrC, and NiTaB. In the apparatus embodiment, the invention is directed to a nuclear reactor having a substantially horizontally oriented core support plate and a plurality of nuclear fuel assemblies each having at least one spring member bearing against the support plate, wherein the spring members include a metallic coating to reduce friction at the bearing surface of the spring member against the support plate. Preferably, the support plate is stainless steel, the spring member is Inconel, and the coating on the spring member, is one of ZrN or TiN. The spring members in some types of conventional upper end fittings, are composed of a plurality of nested, cantilevered spring elements. Preferably, the coating is applied not only to the upper spring element, which contacts the support plate, but also to the other spring elements which are in contact with each other and the upper spring element. This reduces friction and wear within the spring member, thereby also maintaining a predictable spring rate and avoiding excessive wear and corrosion. |
claims | 1. An X-ray beam emitter comprising: a vacuum chamber having a target window; and an electron generator positioned within the vacuum chamber for generating electrons that are directed at the target window for forming X-rays which pass through the target window in an X-ray beam, the target window being supported by a support plate having a series of holes therethrough which allow passage of the electrons therethrough to reach the target window. 2. The emitter of claim 1 in which the target window has a thickness which substantially prevents the passage of electrons therethrough. claim 1 3. The emitter of claim 2 in which the electrons and X-ray beam travel in substantially the same direction. claim 2 4. The emitter of claim 3 further comprising an irradiation region into which the X-ray beam is directed for irradiating articles. claim 3 5. The emitter of claim 4 in which the emitter is a sterilization device where articles irradiated by the X-ray beam are sterilized. claim 4 6. An X-ray irradiation apparatus comprising: an X-ray beam system for directing at least one X-ray beam into an irradiation region, the X-ray beam system comprising at least one X-ray beam emitter, said X-ray beam emitter comprising: a vacuum chamber having a target window; and an electron generator positioned within the vacuum chamber for generating electrons that are directed at the target window for forming X-rays which pass through the target window as said X-ray beam, the target window being supported by a support plate having a series of holes therethrough which allow passage of the electrons therethrough to reach the target window. 7. The apparatus of claim 6 in which the target window of the X-ray beam emitter has a thickness which substantially prevents the passage of electrons therethrough. claim 6 8. The apparatus of claim 7 in which the electrons and X-ray beam travel in substantially the same direction. claim 7 9. The apparatus of claim 8 in which the X-ray beam system comprises more than one X-ray beam emitter for directing X-ray beams into the irradiation region from different directions. claim 8 10. The apparatus of claim 8 in which the X-ray beam system comprises at least three X-ray beam emitters positioned in a ring around the irradiation region, thereby forming a central irradiation chamber. claim 8 11. The apparatus of claim 9 in which the X-ray beam system comprises six X-ray beam emitters positioned in a ring around the irradiation region and abutting each other. claim 9 12. The apparatus of claim 10 in which the X-ray beam system comprises more than one ring of X-ray beam emitters joined together. claim 10 13. The apparatus of claim 8 in which the apparatus is a sterilization apparatus where articles are positioned within the irradiation chamber for sterilization. claim 8 14. The apparatus of claim 8 in which the X-ray beam system comprises at least one irradiation unit having at least one X-ray beam emitter. claim 8 15. The apparatus of claim 14 in which the X-ray beam system comprises more than one irradiation unit joined together. claim 14 16. An X-ray sterilization apparatus comprising: an X-ray beam system for directing at least one X-ray beam into an irradiation region, the X-ray beam system comprising at least one X-ray beam emitter, said X-ray beam emitter comprising: a vacuum chamber having a target window; and an electron generator positioned within the vacuum chamber for generating electrons that are directed at the target window for forming X-rays which pass through the target window as said X-ray beam, said X-ray beam for sterilizing articles positioned within the irradiation zone, the target window being supported by a support plate having a series of holes therethrough which allow passage of the electrons therethrough to reach the target window. 17. A method of forming an X-ray beam emitter comprising: providing a vacuum chamber having a target window; and positioning an electron generator within the vacuum chamber for generating electrons that are directed at the target window for forming X-rays which pass through the target window in an X-ray beam, the target window being supported by a support plate having a series of holes therethrough which allow passage of the electrons therethrough to reach the target window. 18. The method of claim 17 further comprising providing the target window with a thickness which substantially prevents the passage of electrons therethrough. claim 17 19. The method of claim 18 further comprising configuring the X-ray beam emitter so that the electrons and X-ray beam travel in substantially the same direction. claim 18 20. The method of claim 19 further comprising forming an irradiation region into which the X-ray beam is directed for irradiating articles. claim 19 21. A method of forming an X-ray irradiation apparatus comprising: forming an X-ray beam system for directing at least one X-ray beam into an irradiation region, the X-ray beam system comprising at least one X-ray beam emitter; and providing the X-ray beam emitter with a vacuum chamber having a target window, and an electron generator positioned within the vacuum chamber for generating electrons that are directed at the target window for forming X-rays which pass through the target window as said X-ray beam, the target window being supported by a support plate having a series of holes therethrough which allow passage of the electrons therethrough to reach the target window. 22. The method of claim 21 further comprising providing the target window with a thickness which substantially prevents the passage of electrons therethrough. claim 21 23. The method of claim 22 further comprising configuring the X-ray beam emitter so that the electrons and X-ray beam travel in substantially the same direction. claim 22 24. The method of claim 23 further comprising providing the X-ray beam system with more than one X-ray beam emitter for directing X-ray beams into the irradiation region from different directions. claim 23 25. The method of claim 23 further comprising providing the X-ray beam system with at least three X-ray beam emitters positioned in a ring around the irradiation region, thereby forming a central irradiation chamber. claim 23 26. The method of claim 25 further comprising positioning six X-ray beam emitters in a ring around the irradiation region and abutting each other. claim 25 27. The method of claim 25 further comprising forming the X-ray beam system from more than one ring of X-ray beam emitters joined together. claim 25 28. The method of claim 23 further comprising providing the X-ray beam system with at least one irradiation unit having at least one X-ray beam emitter. claim 23 29. The method of claim 28 further comprising joining more than one irradiation unit together to form the X-ray beam system. claim 28 30. The method of claim 22 further comprising forming the apparatus into a sterilization apparatus for sterilizing articles positioned within the irradiation region. claim 22 31. A method of forming an X-ray sterilization apparatus comprising: forming an X-ray beam system for directing at least one X-ray beam into an irradiation region, the X-ray beam system comprising at least one X-ray beam emitter; and providing the X-ray beam emitter with a vacuum chamber having a target window, and an electron generator positioned within the vacuum chamber for generating electrons that are directed at the target window for forming X-rays which pass through the target window as said X-ray beam, said X-ray beam for sterilizing articles positioned within the irradiation zone, the target window being supported by a support plate having a series of holes therethrough which allow passage of the electrons therethrough to reach the target window. 32. A method of forming X-rays comprising: providing a vacuum chamber having a target window; positioning an electron generator within the vacuum chamber for generating electrons; and directing the electrons at the target window to form X-rays which pass through the target window in an X-ray beam, the target window being supported by a support plate having a series of holes therethrough which allow passage of the electrons therethrough to reach the target window. 33. The method of claim 32 further comprising providing the target window with a thickness which substantially prevents the passage of electrons therethrough. claim 32 34. The method of claim 33 further comprising configuring the X-ray beam emitter so that the electrons and X-ray beam travel in substantially the same direction. claim 33 |
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claims | 1. An electron emission lithography apparatus comprising: an electron emission source, which is a carbon nanotube formed in a porous substrate, installed within a chamber; a stage, which is separated from the electron emission source by a predetermined distance and on which a sample is mounted, wherein the electron emission source has electron emission power; a magnetic field generator, which can apply a magnetic field to electrons emitted from the carbon nanotube and control paths of electrons emitted from the carbon nanotube so that the electrons reach predetermined positions of an electron beam resist on the sample; and an insulator thin film patterned on the carbon nanotube. 2. The electron emission lithography apparatus as claimed in claim 1 , wherein the porous substrate comprises Si or Al 2 O 3 . claim 1 3. An electron emission lithography method using a carbon nanotube, comprising: (a) applying voltage to a substrate having a carbon nanotube to emit electrons from the carbon nanotube; (b) controlling the emitted electrons to reach a position corresponding to the carbon nanotube on a sample by applying a magnetic field to the emitted electrons using a magnetic field generator and by controlling the intensity of the magnetic field applied to the electrons according to the distance between the carbon nanotube and an electron beam resist on the sample; and (c) performing lithography on the electron beam resist formed on the sample by the electrons. |
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048511552 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 schematically illustrates a solidification processing apparatus for radioactive waste materials according to the invention. The radioactive waste materials treated herein are ashes of burned waste materials, dried powders of concentrated waste liquids, powdery waste materials of used ion-exchange resins, miscellaneous incombustible solid waste materials such as concrete and heat insulators and solid waste material such as metals. As a solidifying agent, for example, a vinyl type monomer may be used which is able to polymerize at low temperatures with ease. A tank 2 for the solidifying agent comprises a catalyst hopper 1 through which a catalyst is poured into the tank 2 through a catalyst inlet provided in the tank 2. The tank 2 is preferably provided with mixing blades rotatively driven by a motor for mixing the solidifying agent with the catalyst. A solidifying agent supply pipe 3 extends from a bottom of the tank 2 to a waste material vessel 4 and communicates with a cover 5 of the vessel 4 through a flange 25 at a lower end of the pipe 3. Therefore, the solidifying agent in the tank 2 is able to be poured into the waste material vessel 4 through the solidifying agent supply pipe 3. Valve means 19 is provided in the solidifying agent supply pipe 3 to control the amount of the solidifying agent to be poured into the vessel 4 to a predetermined value. The cover 5 seals the waste material vessel 4 in an air-tight manner. The cover 5 is provided with a vacuum valve connected to a vacuum deaerating unit 8. The vacuum deaerating unit 8 serves to remove gases in the waste material vessel 4 to bring the vessel into a negative pressure condition, thereby promoting the pouring of the agent into the vessel 4. The waste material vessel 4 is arranged on vessel transferring means 15 so as to be transferred to a heating and curing chamber 12 for polymerizing and setting the solidifying agent after the solidifying agent poured into the vessel has impregnated the waste materials in the vessel 4. Particular vessels for treating radioactive waste materials as the vessel 4 are not needed. For example, a drum can provided with concrete layers on its inside may be used. The waste material vessel 4 may be provided with impregnation detecting means 7 and with pouring control means 6 connected to the impregnation detecting means 7 and the valve means 19. These means control the amount of the solidifying agent to an appropriate value. A communication tube 26 is provided in the waste material vessel 4 vertically extending to a bottom of the vessel 4 as a sensor included in a practical example of the impregnation detecting means 7 (FIG. 2a). The waste material is filled in the vessel 4 to a level lower than an upper end of the communication tube and the solidifying agent is poured onto an upper end of the waste materials in the vessel 4. The solidifying agent impregnates the waste materials and arrives at the bottom of the vessel 4. Then the solidifying agent enters the communication tube. Therefore, a level of the solidifying agent in the communication tube is detected by a liquid level indicator, for example, provided in the impregnation detecting means 7. As an alternative, the solidifying agent is poured into the communication tube 26 (FIG. 2b). The solidifying agent which has arrived at the bottom of the vessel impregnates from the bottom to the top of the waste materials. A level of the agent on the waste materials is detected by a liquid level indicator provided in the impregnation detecting means 7. As another alternative embodiment of the impregnation detecting means 7, electrostatic capacity measuring terminals 27 are provided in the vessel 4 so as to be in contact with or in the proximity of the bottom of the vessel. Change in electrostatic capacity of the terminals is detected when the solidifying agent poured from the upper end of the waste material has fully impregnated the materials to their bottom. In the solidification processing apparatus according to the invention, sufficient impregnation of the solidification agent in the waste materials is required in order to obtain an appropriate solidified body. The impregnation detecting means 7 ensures complete pouring and impregnation of the agent with great certainty. In order to restrain vaporization of the solidifying agent from the waste material vessel 4 after pouring the agent into the vessel 4, an inner lid 9 may be fitted in the waste material vessel 4, and an inner lid capping unit 10 is mounted thereon. The heating and curing chamber 12 is provided with indirect heating means 11 for heating the vessel 4 together with the waste materials transferred in the chamber 12, thereby setting the solidifying agent and solidifying the waste materials by heating. The indirect heating means may be a heater arranged around the waste material vessel 4. However, a steam heater is preferable in the case using vinyl type monomer (styrene, methyl methacrylate or the like) as the solidifying agent because the steam heater is preferable to prevent deflagration of vaporized combustible monomer and to keep heating temperatures of 50.degree.-70.degree. C., at which the polymerization of the agent is promoted. The heating and curing chamber 12 is further provided with temperature detecting means 21 connected to the indirect heating means 11 for measuring temperatures on outer surfaces of the waste material vessel 4 and in the heating and curing chamber 12, and with valve means 22 connected to the temperature detecting means 21 for controlling the flow of the steam according to outputs of the temperature detecting means 21. In order to collect vaporized gases from the solidifying agent such as vinyl type monomer, a monomer recovery unit 16 may be provided, which is adapted to be connected to the solidifying agent tank 2, the vacuum deaerating unit 8 and the heating and curing chamber 12, respectively. The vaporized gases of the vinyl type monomer are adsorbed or condensed by known adsorbing or condensing means such as active carbon in the monomer recovery unit 16, thereby enabling the recovered monomer to be used again. Moreover, there are provided a filter 17 and an exhausting blower 18 adapted to be connected to the monomer recovery unit 16 for filtering exhaust gases after the recovery of the vinyl monomer and exhausting the filtered gases through the blower 18. The vinyl type monomer allows the heating and curing to be effected at low temperature, and is inexpensive in itself and able to be recovered as above described to reduce the operating cost. Therefore, the vinyl type monomer is advantageous as a solidifying agent for this purpose. The operation of the solidification processing apparatus for radioactive waste materials constructed as above described will be explained hereinafter. A solidifying agent (vinyl type monomer) and a catalyst are poured into the solidifying agent tank 2 and mixed with each other. Thereafter the mixed agent and catalyst of a suitable amount and a suitable viscosity are introduced by dropping onto the powdery or granular radioactive waste materials in the vessel 4. The dropped agent and catalyst progressively impregnate the powdery or granular radioactive waste materials. The impregnated amount is always monitored by the impregnation detecting means 7 and when the solidifying agent becomes a suitable amount, the valve means 19 is closed by the pouring control means 6 to stop the solidifying agent feeding to the waste materials. In this manner excessive pouring of the solidifying agent is prevented. The waste material vessel 4 enclosing the waste materials impregnated with the vinyl type monomer is covered by the inner lid 9 and the inner lid capping unit 10 and transferred into the heating and curing chamber 12 by means of the vessel transferring means 15. In the chamber 12 the vessel 4 covered by the inner lid 9 is arranged in an atmosphere of 50.degree.-70.degree. C. heated by the steam heating which promotes the polymerization reaction of the vinyl type monomer. The polymerization reaction temperature of the radioactive waste materials in the vessel 4 is measured by the temperature sensor 21 secured to an outer surface of the vessel 4. A substantial completion of the polymerization of the solidifying agent is detected by a peak of the polymerization reaction temperature. The temperature of the atmosphere in the heating and curing chamber 12 is also measured by the temperature sensor 21 and is kept substantially at constant by controlling the valve means 22 by referring to the detected temperature. Therefore, the time required for setting the solidifying agent is shortened, and any excessive heating is avoided. In heating, safety is ensured because of the indirect heating. In order to prevent the pressure in the vessel 4 from rising due to the vaporized gases of the solidifying agent and in order to absorb or condense the vaporized gases of the solidifying agent for reuse, the vaporized gases are extracted from the solidifying agent tank 2, the waste material vessel 4, the heating and curing chamber 12 and the like and fed into the monomer recovery unit 16 in which the monomer is recovered. The remaining gases from which the monomer has been recovered is filtered by the filter 17 to remove noxious gases and then exhausted through the exhaust blower 18. The waste material vessel 4 enclosing the waste materials which have been solidified by heating and setting the solidifying agent in this manner is equipped with a lid 13 and a lid capping unit 14 and stored in a particular location. An actual solidification process for waste materials with the apparatus will be explained. Using mimic ashes produced by various solid materials instead of radioactive waste materials, the solidification process was carried out by the use of the apparatus according to the invention. The mimic incineration ash (true specific gravity: 3.0) of 225 Kg was filled to a volume 150 l (bulk specific gravity: 1.5) in a drum can of 200 l (inner capacity: 170 l) having a concrete layer on an inside with the aid of vibration. Styrene monomer added with azobisisobutyronitrile of 2% as a polymerization initiator was used as a solidifying agent. After the solidifying agent had been vacuum deaerated at room temperatures, 80 l of the agent which was more than 75 l of volume of voids in the ash filled in the drum was poured to impregnate the ash. Then the drum was covered by an inner cover made of concrete and heated and cured at 60.degree. C. in a heating and curing chamber to polymerize and setting the agent, thereby obtaining a good solidified body having a specific gravity of 1.9 and a uniaxial compressive strength of 150-200 Kg/cm.sup.2. The solidification processing apparatus for radioactive waste materials according to the invention comprises series of means for pouring, for example, a vinyl type monomer as a solidifying agent superior in impregnation into a vessel filled with the waste materials and thereafter heating and curing the solidifying agent at relatively low temperatures such as 50.degree.-70.degree. C. to polymerize and set the agent and has the following advantages. (1) Any pretreatments such as classifications and crushing or pulverization of radioactive waste materials are not needed. As a result, the apparatus is simple in construction. (2) Mixing and kneading operation of radioactive waste materials with the solidifying agent are not needed. Therefore, any kneader and extruder are dispensed with. (3) As radioactive waste materials are processed by the series of operations of the means of the apparatus, operators are not exposed to radioactive materials. (4) The apparatus according to the invention is easy in maintenance because of its simplicity in construction. (5) As the setting of the solidifying agent is promoted by indirect heating, the setting of the solidifying agent or the solidification of the radioactive waste materials is carried out in comparatively short time. Moreover, as the heating is indirect, there is no risk of deflagration even for a combustible solidifying agent, so that the apparatus is superior in safety. While the invention has been particularly shown and described with reference to preferred embodiments thereof, it will be understood by those skilled in the art that the foregoing and other changes in form and details can be made therein without departing from the spirit and scope of the invention. |
054229227 | summary | BACKGROUND OF THE INVENTION (1) Field of the Invention The present invention relates to boiling water reactors (BWR), particularly, to preferable fuel assemblies and reactor cores for labor-saving fuel shuffling operation by increasing size of the fuel assembly and reducing number of the fuel assemblies with ensuring thermal margin and reactor shut down margin. (2) Description of the Prior Art A fuel assembly for BWR is, in general, composed of a bundle of fuel rods forming a square lattice, each of the fuel rods is manufactured by inserting a plurality of fuel pellets containing fissile material into a cladding tube and sealing, and a channel box having a hollow square cross section, an outer side of which is about 14 cm, which covers the above fuel bundle. A reactor core is formed in a cylindrical shape by further bundling of the above fuel assemblies. As for fuel, enriched uranium or/and plutonium-enriched uranium is used in a chemical form of oxide. As reactivity of reactor core decreases with burning of fuel, the fuel is loaded into the reactor core more than the critical amount at beginning of the reactor operation cycle so that the reactor maintains criticality. Excess reactivity yielded by loading of the excess fuel is controlled by adjusting neutron absorption in the reactor core with mixing burnable poison such as gadolinia etc. into the fuel, and inserting control rods having cruciform cross section, which comprise boron carbide or hafnium, among a plurality of adjacent fuel assemblies. In order to allow inserting the cruciform control rod into the reactor core, water gap regions being filled with non-boiling water, of which gap size is almost twice of a blade thickness of the control rod, are provided around the channel box of the fuel assemblies. Moreover, water rods filled with non-boiling water are provided at center of the fuel assembly in view of neutron flux flattening. Atomic numbers ratio of hydrogen to uranium in the reactor core average (optionally it is called H/U ratio hereinafter) which depends on a size of the above non-boiling water region and the amount of fissile material is adjusted in a range of 4-5 in order to make necessary enrichment of the fuel lowest mainly in view of uranium resource saving. On the other hand, the excess reactivity increases at shut down of the reactor because of increase in water by phase change of steam to water. Accordingly, it becomes important to ensure reactor shut down margin. Regarding to methods for increasing reactor shut down margin of the fuel assembly and the reactor core, the following two methods are well known as prior art; (1) JP-A-63-231293 (1988) In accordance with this prior art, neutron average energy in the reactor core is reduced, a difference in neutron moderating effect between upper portion and lower portion of the reactor core is reduced, and consequently, the reactor shut down margin is increased, by making a ratio of transverse cross section area of the water gap region which is a saturated water region outside the channel box to transverse cross section area of pellets in all fuel rods in the channel box at least one. (2) JP-A-2-12088 (1990) In accordance with this prior art, an excess reactivity of the reactor core is reduced and, consequently, the reactor shut down margin is increased, by composing the fuel assembly so as to have a non-boiling water region of which area is at least 9.1% of the transverse cross section area of the channel box. Hitherto, increase of output power has been achieved in general by increasing in the number of fuel assemblies. However, the increase in the number of fuel assemblies in the reactor core causes increase in the numbers of fuel assemblies to be shuffled and to be transferred in periodical inspection of the reactor core, and consequently, necessary period and man-hour for fuel exchange operations increase and an utilization factor for the plant can be lowered. Therefore, there is a problem that a scale merit which is expected by the increase of the output power is not necessarily obtained. Accordingly, in view of labor saving for fuel exchange operation, it is effective to increase size of a fuel assembly for reducing total number of fuel assemblies in the reactor core. On the other hand, the increase in size of a fuel assembly causes increment of local power peaking factor in a diametral direction because of increase in heterogeneity of the reactor core. Further, the number of the fuel assemblies in the reactor core decreases by increasing size of the fuel assembly under a condition that the reactor core has a constant size. Accordingly, the number of control rods being inserted among the fuel assemblies also decreases. It means relative decrease in total length of the control rod blade, reducing in control rods worth, and decrease in the reactor shut down margin. Therefore, it is necessary to have means for preventing above described problems when increasing size of the fuel assembly. When the above described prior art are applied for increasing size of the fuel assembly, the following defects exist; In accordance with the prior art, JP-A-63-231293 (1988), the reactor shut down margin can be increased by reducing the neutron moderating effect, but thermal margin is decreased by increase in the ratio of the transverse cross section area of the water gap region to the transverse cross section area of the total fuel pellets. In accordance with the prior art, JP-A-2-12088 (1990), the transverse cross section of the non-boiling water region is defined by taking the internal transverse cross section of the channel box as a base. Therefore, there are some cases in which effective increment of the reactor shut down margin can not be achieved depending on a loading condition of the fuel rods in the fuel assembly. As for the thermal margin, the situation is the same. Further, the above defined value for the transverse cross section of the non-boiling water region varies depending on the kind of the fuel material such as uranium-plutonium mixed oxides, or enriched uranium. SUMMARY OF THE INVENTION (1) Objects of the Invention The first object of the present invention is to provide a fuel assembly and a reactor core therefor which are capable of increasing size of the fuel assembly. The second object of the present invention is to provide a fuel assembly and a reactor core therefor which are capable of increasing size of the fuel assembly with ensuring thermal margin. The third object of the present invention is to provide a fuel assembly and a reactor core therefor which are capable of increasing size of the fuel assembly with ensuring reactor shut down margin. In the present invention, increasing size of the fuel assembly is aimed at about 1.5 times of the conventional fuel assembly in consideration of reducing the number of the fuel assemblies about a half of the conventional one. (2) Methods for solving the Problems In order to achieve the above first and the second objectives of the present invention, a fuel assembly having a plurality of fuel rods which are composed by inserting a plurality of fuel pellets containing fissile material into cladding tubes and sealing the cladding tubes, and at least a moderating rod filled with a moderator for moderating neutrons which are generated by nuclear fissions of the fissile material, characterized in having an average ratio at least 0.4 in the axial direction of a sum of transverse cross section area of the portion filled with the moderator of the moderating rods to a sum of transverse cross section area of the fuel pellets is provided. In order to achieve the above first and the third objectives of the present invention, the above fuel assembly preferably having the transverse cross section area for the portion filled with the moderator of 14-50 cm.sup.2 per moderating rod is provided. Further, preferably, the above fuel assembly characterized in having a value in a range of 2.7-3.4 for a ratio of a sum of transverse cross section area of the moderator at a horizontal cross section surrounded by hypothetical planes which are imaginarily formed by extending outer hem of upper tie plate, which bundles upper ends of a plurality of the above fuel rods, downward vertically to the horizontal cross section to a sum of transverse cross section area of the fuel pellets is provided. Further preferably, in order to achieve the first objective of the present invention, the above fuel assembly characterized in having at least a double wall tube in which water level goes up and down depending on flow rate of the moderator as for one of themoderating rods is provided. Further, in order to achieve the above first to third objectives of the present invention, a reactor core having the above fuel assembly according to the present invention is provided. Further preferably, in order to achieve the first and the second objectives of the present invention, the above reactor core characterized in having a ratio utmost 0.7 of a sum of transverse cross section area for the moderator being filled in the water gap region around the fuel assembly to a sum of transverse cross section area for the above fuel pellets is provided. In order to achieve the first and the third objectives of the present invention, the above reactor core preferably having a control rod which is composed of a plurality of absorbing rods containing neutron absorber bundled in a shape having a cruciform cross section, and being inserted into the water gap region around the fuel assemblies, and a ratio at least 0.20 for a sum of surface area of the above absorbing rods to a sum of surface area of the above fuel rods is provided. Further preferably, the above reactor core having a control rod which is composed of a plurality of absorbing rods containing neutron absorber bundled in a shape having a cruciform cross section, and being inserted into the water gap region around the fuel assemblies, and a ratio at least 0.4 for a sum of transverse cross section area of the above absorbing rods to a sum of transverse cross section area of the above water gap region is provided. Further preferably, the above reactor core having a value between 3.0-3.5 for a ratio of a sum of transverse cross section area of the above moderator to a sum of transverse cross section area of the above fuel pellets is provided. In accordance with the present invention which is composed as above described, a transverse cross section area of the moderator in the moderating rod is increased, a transverse cross section area of the water gap region is decreased, values of local power peaking factor is decreased, and thermal margin is ensured, by making a ratio of transverse cross section area of the portion filled with moderator in the moderating rod to transverse cross section area of the fuel pellets averaged in the axial direction at least 0.4. Moreover, the excess reactivity is reduced and the reactor shut down margin is ensured, by making the transverse cross section area of the moderator per moderating rod 14-50 cm.sup.2. Further, the excess reactivity is reduced and the reactor shut down margin is ensured, by making a ratio of a sum of transverse cross section area of the moderator at a horizontal cross section surrounded by hypothetical planes which are imaginarily formed by extending outer hem of upper tie plate downward vertically to the horizontal cross section to a sum of transverse cross section area of the fuel pellets a value in a range of 2.7-3.4. By using a double wall tube water rod as for at least one of the moderating rods, effects of spectrum shift are multiplied to reduce necessary uranium enrichment, and an operation without inserting control rods can be performed. The local power peaking factor is reduced and the thermal margin can be ensured by making a ratio of a sum of transverse cross section area of the moderator being filled in the water gap region around the fuel assemblies to a sum of transverse cross section area of the fuel pellets utmost 0.7. The control rod worth is increased and the reactor shut down margin is ensured by making a ratio of a sum of surface area of the absorbing rods to a sum of surface area of the fuel rods at least 0.20. Further, the control rod worth is increased and the reactor shut down margin is ensured by making a ratio of a sum of transverse cross section area of the absorbing rods to a sum of transverse cross section area of the water gap region at least 0.4. The excess reactivity is decreased and the reactor shut down margin is ensured by making a ratio of a sum of transverse cross section area of the moderator to a sum of transverse cross section area of the fuel pellets a value in a range 3.0-3.5. |
summary | ||
047479930 | description | DETAILED DESCRIPTION OF THE INVENTION This invention provides new annular ring seals for establishing a permanent seal across the annular thermal expansion gap between the peripheral wall of a nuclear reactor vessel and a containment wall. The new seals may be installed in place of existing prior art seals of either the removable or permanent types. Although their installation may be considered permanent, they can be removed and repaired or replaced if the need arises. The seals according to the present invention are constructed to cyclicly contract and expand, respectively, with cyclic expansion and contraction of the reactor vessel relative to the containment wall during reactor operation, while maintaining the integrity of the seal, most preferably while maintaining the water-tight integrity of the seal. The seals include a flexure member which provides the required flexibility. The seals are structured so that the flexure member is protected from falling objects by virtue of its positioning. Further, the seals include an annular ring plate member which is strong enough to accommodate the force of even the heaviest object which might be accidentally dropped on the plate member during the reactor refueling operation, without complete or sudden loss of shielding water. Thus, the flexure member of each annular ring seal has an annular base plate which sealingly engages and is affixed to the annular ring plate of the seal near an inner surface thereof proximate the reactor vessel wall, and a leg which extends from the annular base plate, is affixed to the reactor vessel, and sealingly engages the reactor vessel. The flexure member most preferably has an L-shaped cross-section in which case the leg is a cylindrical leg which extends perpendicularly from the annular base plate. The perpendicular, i.e., vertical, extension of such a cylindrical leg from the annular base plate is a structural feature which minimizes the possibility of damage thereto from falling objects by minimizing the exposed horizontal cross-sectional area of the flexure member. However, protection of the flexure member may be accomplished by providing a leg which makes an acute angle with the annular base plate, or, alternately, by a leg which makes an obtuse angle with the annular base plate, so long as the leg is protected from vertically dropped objects by being nested under the reactor vessel flange or an annular ledge. Thus, the flexure member may have a conical cross-section or a cross-section of any other configuration. This protected flexure member feature significantly contributes to satisfying the need for assurance that a sudden or complete loss of shielding water from the refueling canal will not occur by virtue of a structural failure of the ring seal. The protected flexure member need not be strong enough to directly withstand the impact of a dropped object, but may be manufactured from sheet metal, such as stainless or carbon steel, and must be thin enough to provide the flexibility required to accommodate the cyclic thermal expansion and contraction required of the seal during reactor operations, while having at least the strength to withstand the weight of the shielding water introduced during the reactor refueling operation. These new ring seals may advantageously be provided with a backup member to provide a back-up structure for the flexure member to preclude the possibility of a major leakage in the event of a structural failure of the flexure member, however improbable. The backup member serves to restrict the flow of water in the event of a structural failure of the flexure member, thereby providing the refueling crew with time to make safe the exposed reactor core or any exposed fuel assemblies, by redundantly assuring the prevention of the catastrophically sudden and/or complete loss of shielding water from the refueling canal. The backup member includes a backup plate which is positioned inboard of the flexure member, i.e., in closer proximity to the reactor vessel, and which extends from the annular ring plate toward the reactor vessel. The backup plate is most preferably a cylindrical plate extending vertically in cross-section, however, the backup plate may have a conical cross-section of virtually any slope, or a cross-section of any other configuration. The backup plate preferably does not contact the reactor vessel, since the backup member is not intended to sealingly engage the reactor vessel. Sealing engagement of the backup member to the vessel would reduce the flexibility of the structural arrangement needed for thermal expansion/contraction and, in any event, would be difficult to achieve in a permanent ring seal because of the non-accessibility of the juncture of the backup member with the vessel during installation. The backup member may further include a flashing which biasingly engages the reactor vessel, but which does not sealingly engage the vessel. The term "flashing" as used herein is meant to include a member having any cross-section, such as a conical, hemispherical or spherical cross-section as provided by, for example, a conical ring, a half-moon ring, or an O-ring, respectively. The invention can be better understood by reference to FIG. 1, an elevational view partly in cross-section of a nuclear reactor containment arrangement showing the pertinent part of a nuclear steam generating system of the pressurized water type incorporating a permanent ring seal according to this invention. A nuclear reactor vessel 1 is shown which forms a pressurized container when sealed by its head assembly 3. The reactor vessel 1 has coolant flow inlet means 5 and coolant flow outlet means 7 formed integral with and through its cylindrical wall. The reactor vessel 1 contains a nuclear core (not shown) consisting mainly of a plurality of clad nuclear fuel elements which generate substantial amounts of heat depending primarily upon the position of control means, pressure vessel housing 9 of which is shown. The heat generated by the reactor core is conveyed from the core by coolant flow entering through inlet means 5 and exiting through outlet means 7, as is well established in the art. The reactor vessel 1 and head assembly 3 are maintained within a reactor cavity defined by a concrete containment wall 11. The reactor cavity is divided into an upper portion which defines and is commonly utilized as a refueling canal 13 and a lower portion which is a well 15 completely surrounding the vessel structure itself. A shelf 17 is provided in the containment wall 11 and divides the upper portion of the reactor cavity from the lower portion. An open annulus defined between the containment wall 11 and the reactor vessel 1 is an annular thermal expansion gap 19 and is provided with an annular ring seal 21 at least during refueling operations, but preferably is positioned permanently. The ring seal 21 accommodates the normal thermal expansions and contractions of material experienced during reactor operation, while maintaining the integrity of the seal essential during refueling operations in which the upper portion of the reactor cavity defining refueling canal 13 is filled with water in order to provide a radiation shield for the exposed core and refueling assemblies. Preferably, the annular ring seal 21 is a permanent seal which sealingly engages and is affixed to the containment wall 11 (or fixture thereof), preferably along the shelf 17, and to a peripheral wall 23 (or fixture thereof) of the reactor vessel 1, preferably to a lower flange 25 (or fixture thereof) positioned along the peripheral wall 23 near the reactor vessel opening (not shown), the opening being exposed when the head assembly 3, which includes an upper flange 27, is removed during the refueling operation. The preferred annular ring seal 21 of this invention can be better appreciated by reference to FIG. 2 which shows, in cross-section, a portion of a ring seal 21 and its interface with the lower flange 25 and the shelf 17 of the containment wall 11. In this preferred embodiment, the annular ring seal 21 comprises an annular ring plate (shown generally at 29). The annular ring plate 29 has a step-shaped cross-section and has a first annular portion 31, a second annular portion 33, and a cylindrical portion 35 which interconnects the first and second annular portions 31, 33, respectively, to form the step-shaped cross-section. The cross-section may alternately have any configuration. The first annular portion 31 surrounds the second annular portion 33 and includes an outer peripheral surface (shown generally at 37) for the annular ring plate 29. The second annular portion 33 includes an inner peripheral surface (shown generally at 39) for the annular ring plate 29. With continuing reference to FIG. 2, the annular ring seal 21 further comprises a flexure member 41 shown as having an L-shaped cross-section. The flexure member 41 has an annular base plate 43, which sealingly engages and is affixed to the annular ring plate 29 near the inner peripheral surface 39 thereof, the inner peripheral surface 39 of this embodiment being provided in the second annular portion 33. The flexure member 41 also has a leg 45, which is joined to and extends upwardly from the annular base plate 43 and which sealingly engages and is affixed to an annular ledge 47 provided along the lower flange 25 for engaging the ring seal 21. Leg 45 is shown as a cylindrical leg extending perpendicularly upwardly from the annular base plate and forming an L-shaped cross-section therewith. As discussed previously, however, the leg may make an acute angle or an obtuse angle with the annular base plate and still provide a structure in which the flexure member is protected from falling objects by virtue of its positioning. Thus, for example, when the leg and annular bsse plate make an angle other than a right angle with one another, the leg should lie beneath and be protected by, for example, the annular ledge. Further, the extremity of the leg which engages the peripheral wall or a fixture thereof, such as the flange or annular ledge, should be adapted as necessary. The annular ledge 47 has a first surface 49 which overlappingly and sealingly engages the leg 45 of the flexure member 41. FIG. 2 shows a most preferred embodiment of the present invention in which the annular ring seal 21 also comprises a backup member 51 including backup plate 53 and a flashing 55. The backup plate 53 is shown as a cylindrical plate and is surrounded by the flexure member 41 and has a first perimeter 57 and a second perimeter 59. The first perimeter 57 is joined to the inner peripheral surface 39 from which the backup plate 53 extends upwardly. The flashing 55 extends from the second perimeter 59 and is biased to engage a second surface 61 of the annular ledge 47. In embodiments without a flashing 55, the second perimeter 59 of the backup plate 53 may extend toward the underside second surface 61 of the annular ledge 47, without directly engaging the same, but rather defining an gap therewith sufficient to allow for thermal expansion and contraction. The backup member 51 provides a backup structure for the flexure member 41 to preclude the possibility of major leakage of shielding water (not shown) in the event of a structural failure of the flexure member 41. The backup member 51 does not form a water-tight seal with the annular ledge 47, but rather functions as a flow restrictor to prevent a catastrophically sudden and/or complete loss of shielding water from the refueling canal 13 during the refueling operation, in the event of a structural failure from, for example, the accidental dropping of a fuel assembly onto the annular ring seal 21. By restricting the flow, the refueling crew is given the necessary time to make safe any exposed radiation sources, thereby further assuring the safety of the crew and preventing contamination of the environment. FIG. 2 also shows the shelf 17 of the containment wall 11 fitted with a mating plate 63 which has a rectangular cross-section. The mating plate 63 may be, for example, imbedded into or bolted onto the shelf 17 through apertures (not shown) provided in the respective parts and preferably sealingly engages the shelf 17. A lip or recess 65 is shown defined along the outer peripheral surface 37 of the first annular portion 31 of annular ring plate 29. The thickness of the first annular portion 31 in the area of the recess 65 is less than the thickness of the remainder thereof to permit plastic deformation of the first annular portion 31 and to facilitate overlapping and sealing engagement thereof with the mating plate 63. Recess 65 also facilitates initial positioning of the annular ring seal 21 during installation and prevents gross movement thereof in the event of a weld failure. Annular ring seal 21 may also comprise a plurality of support arms 67, each support arm 67 extending from the cylindrical portion 35 of the annular ring plate 29 toward the reactor vessel 1. Each arm is adapted to rest on the reactor vessel 1, such as on a third surface 69 of the annular ledge 47, thereby providing auxiliary support for the annular ring seal 21. Each support arm 67 is preferably provided with alignment means, shown in FIG. 2 as a threaded aperture 71 defined in each support arm 67 along the portion thereof which overlaps the annular ledge 47 and a leveling bolt 73 threadingly inserted through the threaded aperture 71. Each bolt 73 engages the annular ledge 47 for initially aligning the annular ring seal 21 during installation and for providing auxiliary support for the annular ring seal 21 during the refueling operation. Bolts 73 may optionally be removed during reactor operation. Support arms 67 may be welded onto the annular ring plate 29 and may be conveniently used to support open grating plates (not shown) for personnel to stand on during the refueling operation, which open grating plates rest upon and span the distance between support arms 7. Support arms 67 may be preferably placed at from 20 to 30 degree intervals along the annular ring seal 21. FIG. 3 is a cross-sectional plan view of the reactor containment incorporating the annular ring seal 21 of FIG. 2 through section line A--A. Reactor vessel 1 is shown schematically. The annular ring plate 29 of the annular ring seal 21 is more clearly shown as being comprised of a plurality of arc sections 29a, 29b, 29c, 29d. The four arc sections 29a, 29b, 29c, 29d shown have equal arc lengths. Other embodiments are contemplated which have any number of arc sections, such as three arc sections, and the arc lengths thereof need not be equal. Each arc section may have a pair of flanges 75, shown in FIG. 3 as flange pairs 75a, 75b, 75c, 75d, respectively, for interconnection thereof with adjacent arc sections using flange connection means. Alternate connection means are contemplated however and should be considered to be within the scope of the present invention. The flange connection means shown in FIG. 3 comprises a plurality of dowel pins 77 and a plurality of apertures 79 provided in each flange 75. The apertures 79 of each of a pair of flanges from adjacent arc sections, for example, flanges 75a and 75b from adjacent arc sections 29a and 29b, are aligned respectively and are adapted to receive one dowel pin 77. Adjacent arc sections may further be provided with a sealing weld (not shown). FIG. 4 is an enlarged cross-sectional view of a portion of another preferred annular ring seal according to the invention. Annular base plate 43' is shown offset and positioned above the plane occupied by annular ledge 47 and annular shelf 17. Leg 45' of flexure member 41' extends downwardly from the annular base plate 43'. By way of example, when the flexure member 41 is designed to withstand up to 12.1 psig of water pressure, and the annular ring plate 29 is designed to withstand the impact of a fuel assembly weighing on the order of 700 kg, the annular ring plate 29 is constructed of 2.54 centimeter thick stainless or carbon steel and the seams joining portions 31, 35 and 33 are welded together. The flexure member 41 is likewise constructed of stainless or carbon steel and has a thickness ranging from 1.6 millimeters to 3.2 millimeters. The backup plate 53 of the backup member 51 is likewise constructed of stainless or carbon steel and has a thickness on the order of 1.27 centimeters. The flashing 55 is biased against the annular ledge 47 and is constructed of stainless steel of approximately 0.3 to 0.4 millimeter. The flashing 55 may alternately be constructed of an elastomer such as rubber and have a thickness of from about 0.8 to 1.6 millimeters. It is preferable that the various metal portions be all welded together so that the ring seal 21 is gasket free. Further, surface non-uniformities in mating plate 63 are compensated for during installation by using a weld prep 81, shown in FIG. 2, so that the weld prep is compressed on installation by means of a roller or a weld plate. The annular base plate 43 of the flexure member 41 may be affixed to the inner peripheral surface 39 of the second annular portion 33 of the annular ring plate 28 by field welding and, optionally and/or additionally, plug welds 83 may be provided through intermittent apertures 85 provided in annular base plate 43 and spaced apart by from 15 to 30 centimeters. Likewise, the leg 45 of the flexure member 41 may be welded to the first surface 49 of the annular ledge 47 and plug welds 83 may be provided through intermittent apertures 87 provided in the leg 45 and spaced apart by from 15 to 30 centimeters. Fillet welds 89 may also be advantageously used, as shown in FIG. 2, at the ends of annular base plate 43 and leg 45, such weld types being well known in the art. It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptations, and the same are intended to be comprehended within the meaning and range of equivalents of the appended claims. |
043022880 | summary | BACKGROUND The invention relates to a system for controlling the fluid level in a vessel. Among numerous applications for such a system is the maintenance of the water level in a nuclear reactor. In well-known commercial boiling water nuclear power reactors, for example as used in the Dresden Nuclear Power Station near Chicago, Ill., a core of fuel material contained in a pressure vessel is submerged in a fluid, such as light water, which serves both as a working fluid and a neutron moderator. The water is circulated through the core whereby a portion thereof is converted to steam. The steam is taken from the pressure vessel and applied to a prime mover such as a turbine. The turbine exhaust steam is condensed and, along with any necessary make-up water, returned to the pressure vessel as feedwater. Reactor power level is controlled by a system of control rods, containing neutron absorber material, which are selectively insertable into the core. Further information on nuclear reactors may be found, for example, in "Nuclear Power Engineering", N. M. El-Wakil, McGraw-Hill Book Company, Inc., 1962. Nuclear reactors are provided with a protection system which monitors various aspects of reactor operation including water level. If a fault develops in the water level control system and it fails to maintain a predetermined water level in the vessel, the water level becomes "out of limits", either too high or too low, and the protection system "scrams" the reactor that is, it causes rapid insertion of the control rods whereby the reactor is shut down automatically. Such reactor shutdowns are undesirable for a variety of reasons. Even if the fault is corrected readily, restart of the reactor is a relatively lengthy process. Meanwhile, customers may suffer a loss of power or the power must be supplied from other, usually more costly, sources. Thus an object of the invention is an improved liquid level control system which is tolerant of a failure of components therein. Another object is a level control system having redundant control channels with means for switching control automatically to another channel upon an excursion of the liquid level beyond prescribed limits. SUMMARY These and other objects are achieved, according to the invention, by providing at least one rudundant channel for control of the flow of feedwater to the vessel. A plurality of water level sensors are positioned on the vessel at the positions of normal operating water level upper and lower limits. When a majority of the upper or lower limit sensors indicate an excursion of the water level beyond a limit, feedwater flow control is switched automatically from the initial control channel to a redundant control channel whereby a component failure in the initial channel does not result in a reactor scram. Means are also provided to switch control immediately to the redundant channel in response to a rapid change in the water level control signal of the initial channel. Another aspect of the feedwater flow control system is detection of the steam outflow and feedwater inflow flow rates. In normal operation feedwater flow is controlled in accordance with the vessel water level and the difference between the water equivalent of the steam flow and the feedwater flow. In such a system, component failure usually is manifested by a rapid change in the difference signal. Thus in accordance with the invention a rapid change in the difference signal causes the steam and feedwater flow aspect of water level control to be switched out of the circuit, control thereupon being assumed solely by the water level sensor arrangement without causing a reactor scram. |
052951677 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to FIGS. 1 and 2, the service pole caddy system 2 in accordance with the preferred embodiment of the invention comprises a rigid frame made up of welded tubes. The rigid frame comprises right and left frames, each comprising a vertical tube 4, an oblique tube 6, a vertical tube 8, an oblique tube 10, a horizontal tube 12 and a horizontal tube 14. The right and left frames are connected by four horizontal tubes 16. The rigid frame is mounted on a truck 18 via plate 19. Truck 18 rides on a rail 20 which is supported by an I-beam 22 of the refueling bridge, as shown in FIG. 2. A monorail hoist 24 has a motorized trolley which rides on an I-beam 26 supported by the rigid frame. The travel of hoist 24 is stopped at the ends of I-beam 26 by respective stop angles 28. The maximum travel of the hoist may be about 46 feet or any other length depending on the dimensions of the specific reactor with which the pole caddy system is used. The hoist 24 includes a control box 30 and a hook 32 which is coupled to a pole connector to be described in detail below. As best seen in FIGS. 3 through 5, the rigid frame further comprises a rigid structure welded to tubes 4 for supporting the pole service caddy. This structure includes two vertical tubes 34 separated by a distance equal to the distance between tubes 4; two horizontal tubes 36 each welded at respective ends to a tube 4 and a tube 34; a horizontal tube 38 welded at respective ends to tubes 4; and a horizontal tube 40 welded to one of the tubes 34 and parallel to horizontal tube 38. Horizontal tubes 38 and 40 are in turn joined by an angle 42. As shown in FIG. 4, a lower rack 44 is welded into the rectangular space defined by tubes 36, 38, 40 and angle 42. Lower rack 44 is a four-by-two array of tubular cells of square cross section, each tubular cell being designed to receive one end of a generally upright service pole. Lower rack 44 has a perforated plate 45 (see FIG. 3) at the bottom which supports the ends of the service poles while allowing water from the poles to drain through and into the pool below. As shown in FIG. 5, the rigid structure further includes two horizontal angles 46 each welded at respective ends to a tube 4 and a tube 34; a horizontal angle 48 welded at respective ends to tubes 4; and a horizontal angle 50 welded to one of the tubes 34 and disposed parallel to angle 48. Horizontal angles 48 and 50 are in turn joined by an angle 52. An upper rack 54 is welded into the rectangular space defined by angles 46, 48, 50 and 52. Upper rack 54, like lower rack 44, is a four-by-two array of tubular cells of square cross section, each tubular cell being designed to receive a mid-portion of a generally upright service pole. The upper rack is supported by a channel member 56, which is welded to tube 40 at the lower end and to angle 50 at the upper end. The lower and upper racks are positioned in alignment at different elevations and together maintain the service poles 58 in a generally vertical storage position, one pole to a cell, as shown in FIG. 6. In an exemplary embodiment, each cell has a width of 31/4 inches, whereas each pole has an outer diameter of 21/2 inches, allowing for easy storage and removal. The frame shown in FIG. 5 further comprises a keyway plate 60, which is swingably mounted on a hinge 61 welded to angles 48 and 52. In the down position, keyway plate 60 has a portion which rests atop angle 48 and is supported thereby. Keyway plate 60 has a two-position slot 62 with two throats 68, 70 of width less than the outer diameter of the service poles, but greater than the diameter of the neck of the pole end connector (to be described in detail below). At a first position, slot 62 has opposing recesses 64 which are arcs of a circle of radius 11/2 inches. The tubing of a service pole intersecting the slot 62 and positioned between recesses 64 cannot be displaced laterally except for a small amount of play. This play allows for guided vertical displacement of the pole when in position B shown in FIG. 6. The slot 62 ends at a second position in a semicircle 66 of diameter 21/8 inches, which is slightly greater than the pole neck diameter but less than the pole outer diameter. The upper edge of slot 62 at the second position is chamfered to form a seat 72 which supports a service pole in position C shown in FIG. 6. As best seen in FIG. 6, the service poles 58 are lifted, lowered and carried by means of hoist 24, which is an electric hoist with a motorized trolley that rides on monorail 26. The hoist has a coupling 74 adapted to couple with the end connector of a service pole. A service pole 58 can be lifted out of the storage caddy, carried laterally and then lowered into guided pole handling position B. A pole in position B can be moved into pole assembly position C by first displacing the pole vertically while in position B until its neck is lined up with slot 62 and then pushing the pole so that its neck passes through throat 68. When the pole overlies position C (see FIG. 6), the pole is lowered until its seating portion abuts and is supported by support seat 72. Then another pole can be retrieved from the caddy and placed directly over the pole being supported in position C. Thereafter the bottom end connector of the second pole is coupled to and locked on the top end connector of the pole being supported in position C. This operation is repeated until the multi-pole assembly is completed. The multi-pole assembly can then be carried by the hoist to open pole handling position A and lowered as necessary to line up the tool, mounted on the lower end of the assembly, with the component to be manipulated by the tool. The service pole position is also determined by the position of the movable rigid frame. As previously described, the rigid frame is mounted on truck 18, which rides on rail 20 supported by I-beam 22 of the refueling bridge. Some of the components of truck 18 are shown in FIGS. 9A and 9B. The truck has two wheels 118 mounted on respective shafts 120. Other truck components include four bearings 122, four spacers 124, four hold-down plates 126, two rubber wiper rails 128 and four cam followers 130. The rigid frame, in addition to riding on a rail supported by the refueling bridge I-beam, is also provided with a pair of wheels 76, 76' and a pair of casters 78, 78' (see FIG. 1). As best seen in FIG. 2, wheels 76 and 76' ride on a lower leg of a refueling bridge channel 80, while casters 78, 78' bear against the vertical central member of channel 80. The casters 78, 78' are supported by wheel support 84, which is an angle welded to vertical tubes 4 (see FIGS. 1 and 7). The wheels 76, 76' are supported by brackets 82 via hubs 86 (see FIG. 7). Brackets 82 are welded to horizontal tube 38. Referring to FIG. 3, the frame further comprises two supporting angles 88, each supporting angle being welded to a vertical tube 4 and a vertical tube 34. Angles 88 support the auxiliary personnel work platform, which is generally designated by numeral 90 in FIG. 8. The work platform 90 is generally L-shaped, the two legs of the L being situated adjacent two sides of the pole storage station. The work platform has a welded base frame comprising angles 92, 94, 96, 98, 100, 102 and channels 104, 106. This base frame has a floor plate 108 welded thereon, on which the personnel stand. The perimeter of the work platform has a front railing 110 and two side railings 112, as best seen in FIG. 1. The perimeter of the base frame has a front kick plate 114 and two side kick plates 116 (see FIG. 8). Tubes 4, 6, 8, 34 and 36 are preferably 4".times.4".times.1/8"; tubes 12 and 14 are preferably 4".times.4".times.1/4"; and tubes 10, 16, 38 and 40 are preferably 3".times.3".times.1/8". All tubes can be made of ASTM A500 steel or any other functionally equivalent material. All angles referred to hereinabove are preferably 3".times.3".times.1/4" and can be made of ASTM A36 steel. The service poles used with the service pole caddy system of the invention have a diameter which is larger than that of conventional service poles. These large-diameter poles are designed to transmit torques adequate for all expected operations. The end connectors of these poles have a plug and socket design with a twist pin and a locking collar. This design maximizes the rigidity of the multi-pole assembly. Solid body end connectors are welded to each end of the poles to prevent contaminated water from entering the inside of the poles and to achieve pole buoyancy. Internal pole contamination is prevented and exterior decontamination can be easily accomplished. The solid body of each pole connector provides the equivalent shielding of eight feet of water in a four-pole assembly, thus preventing radiation streaming. Referring to FIG. 10, each service pole 58 comprises stainless steel tubing 134 with solid body end connectors 132 and 136 welded to respective ends thereof. End connector 132 has a pair of diametrally opposed, radially outwardly extending circular cylindrical twist pins 138 which cooperate with corresponding slots in locking collar 140 formed in end connector 136. The hoist 24 is provided with a coupling having twist pins identical to those on end connector 132 to enable the hoist to couple with the end connector 136 of any service pole which needs to be hoisted. Each end connector 132 is provided with threads 142. When threaded coupling 144 is screwed onto threads 142 of an end connector 132 coupled to an end connector 136, the connectors are effectively locked together. The pieces of the split ring are then attached to coupling 144, thereby preventing the coupling from being accidently screwed off and dropped into the reactor. Each end connector 136 has a neck 148 and a chamfered portion 150. When neck 136 passes through throat 68 of keyway plate 60 (see FIG. 5) to the second position, the pole is lowered until chamfered portion 150 rests on support seat 72, whereby the keyway plate supports the service pole, as indicated at position C in FIG. 6. The chamfered portion 150 form-fits with the chamfered support seat 72 such that lateral displacement of the chamfered portion 150 of the service pole seated thereon is resisted. The service poles can be of two different sizes. For example, one set of poles could have dimensions A and B (see FIG. 10) of 48 and 385/8 inches respectively, while another set of poles could have dimensions A and B of 96 and 865/8 inches respectively. The stainless steel tubing 134 has an outer diameter of 21/2 inches and a wall thickness of 0.083 inch. The preferred embodiment has been described for the purpose of illustration only. Various modifications of the service pole caddy system in accordance with the invention will be apparent to a skilled engineer. For example, the service pole caddy system disclosed herein can be readily adapted to cooperate with a refueling bridge having supporting structure different than that disclosed herein. |
summary | ||
abstract | A method of sample extraction entails making multiple, overlapping cuts using a beam, such as a focused ion beam, to create a trench around a sample, and then undercutting the sample to free it. Because the sidewalls of the cut are not vertical, the overlapping cuts impinge on the sloping sidewalls formed by previous cuts. The high angle of incidence provides a greatly enhanced mill rate, so that making multiple overlapping cuts to produce a wide trench can requires less time than making a single, deep cut around the perimeter of a sample. |
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055966136 | abstract | A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs. |
abstract | The boron concentration can be measured without intervention into the cooling circuit of a nuclear power station. A mobile emitter and a mobile receiver are provided, with the interposition of at least one coolable region, for placement on a coolant-carrying component of the cooling circuit. |
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summary | ||
description | This application is a divisional of, and claims priority to, U.S. patent application Ser. No. 14/819,271, filed Aug. 5, 2015, entitled “Production of Copper-67 from an Enriched Zinc-68 Target,” now issued as U.S. Pat. No. 10,006,101, which claims priority to U.S. Provisional Patent Application No. 62/035,113, filed Aug. 8, 2014, entitled “Production of Copper-67 from an Enriched Zinc-68 Target,” the entire contents of both of which are incorporated herein by reference. This application is related to a method and apparatus for the production of radiopharmaceutical copper-67. Further, the application describes a sublimation apparatus and target assembly of the sublimation apparatus used to improve the methods of producing copper-67. Nuclear medicine is a branch of medicine that relies on radiation to both diagnose and treat a variety of conditions, including many types of cancers, heart diseases, and other disorders. Within nuclear medicine, diagnostic or imaging techniques use radioisotopes that are either gamma or positron emitters. Typically, the majority of medical procedures involving radioisotopes are for diagnostic applications. A smaller percentage of the procedures are for therapeutic purposes. In either case, these radioisotopes are relatively short-lived (i.e., a short half-life) and are linked or conjugated to chemical compounds known as radiopharmaceuticals. A radiopharmaceutical preferably binds to one or more sites of a tissue or cancer cell. As many cancer cells have a limited number of available binding sites, the administration of a non-labeled bio-conjugate will often times block one or more of the cellular sites. Therefore, radioisotopes used in the labeling of a bio-conjugate preferably have high specific activity to minimize the attachment of non-labeled bio-conjugates that have little to no therapeutic or diagnostic effect. With the use of a gamma-detecting camera, radiopharmaceuticals are used to construct 3-D images of different organs and tissues, thereby providing information on organ function or disease. This data may in turn be used for reliable and accurate medical diagnosis. One such radiopharmaceutical, copper-67, has a half-life of about 62 hours, and has a gamma-ray energy suitable for imaging. Copper-67's beta particle is also of sufficient energy for therapy with a cell range of less than 2 mm and the gamma ray is detectable using a SPECT camera. In addition, the chelation chemistry of copper is well established and copper is well tolerated by the body, particularly at the trace levels administered to patients. Furthermore, a copper-67 radiopharmaceutical has sufficient range to target and irradiate small tumors without damaging surrounding healthy tissue. Copper-67 has been used in studies to treat non-Hodgkin's lymphoma and shows promise in treating many types of cancer. The half-life of copper-67 also delivers a low systematic radiation dose to the patient and allows for its transportation from a generation facility to a medical center or research laboratory. Methods for producing copper-67 have included nuclear reactors and bombarding zinc oxide with high energy protons. Like reactor generation, producing copper-67 using high energy proton accelerators has high inherent capital and operational costs, scheduling issues, and product contamination risks. The specific activity of copper-67 from proton production methods also exhibit wide variability. As nuclear medicine continues to be an important part of non-invasive disease diagnosis and treatment, there exists the need to produce copper-67 without the described drawbacks. Overview The current disclosure is directed towards a method of producing radioisotopes and a sublimation assembly, apparatus, and vessel used for producing and isolating radioisotopes, for example, copper-67. In one embodiment, the copper-67 may be produced with a purity, specific activity, and consistency suitable for diagnostic and therapeutic applications. The description itself is not intended to limit the scope of this patent. Rather, the inventors have contemplated that the claimed invention might also be embodied in other ways, to include different elements or combinations of elements similar to ones described in this document, in conjunction with other present or future technologies. As used herein, beam energy contains the units MeV, current is described in microAmps, and the power of the beam is expressed as kW. Sublimation is a separation and purification technique for inorganic solids. In general, the separation of an inorganic mixture using sublimation includes placing a solid mixture in a flask positioned within a heating element. Located above the flask is a condenser, which may include a continuous flow of water or other coolant fluid. At times, the condenser may comprise a hollowed and exposed inner volume in which a dry-ice slush bath may be used to cool the condenser. The flask containing the mixture may be heated to a select temperature, and the interior volume of the sublimation apparatus may be placed under a dynamic vacuum. The inorganic component with a greater vapor pressure at the given temperature of the solid mixture and the pressure of the sublimation apparatus may therein condense upon a solid surface of the condenser. Following a certain period of time, the heat may be removed from the flask, and thereafter the condenser removed. The separated, and typically the desired, inorganic component may then be scrapped from the surface of the condenser. In one embodiment, the sublimation apparatus may generally include a support to position a solid mixture within the sublimation vessel and a collection vessel. The sublimation vessel may generally have a heat volume portion configured to fit within a region of a heating element, and a warm volume portion that extends from the heat volume portion to a location outside of the region of the heating element. The collection vessel may include an upper end and an opposite, open end, with an internal sidewall extending from the upper end to the opposite open end, thus forming an internal volume to the collection vessel. Upon applying heat to the sublimation vessel via a heating element, for example, and consequently heating the solid mixture positioned within the heat volume portion of the sublimation vessel, one or more metal vapors may condense upon the internal sidewall, causing the metal vapors to collect within the internal volume of the collection vessel. Furthermore, described is a method for producing the radioisotope, copper-67, from an isotope-enriched metal target comprising zinc-68, hereafter referred to a “zinc-68 target,” a “metal target,” or a “target.” The method of separation may include positioning a solid mixture comprising copper-67 and zinc-68 in a sublimation apparatus. The sublimation apparatus may include, for example, the previously described structure. The method may incorporate heating the solid mixture to a temperature sufficient to form metal vapor comprising greater than 90% by weight zinc-68. The heated metal vapor thereafter may condense within the internal volume of the collection vessel and onto the internal sidewall. Condensation may occur as zinc-68 has an appreciable vapor pressure at a temperature from 300° C. to 700° C. and at a nominal pressure of 10−5 mbar within the sublimation vessel. The method of producing copper-67 may begin by positioning the zinc-68 target in a target assembly and directing an electron beam with an energy of at least 20 MeV and an average power of at least 1 kW, onto a first in a series of at least three, substantially parallel Bremsstrahlung converter plates. As discussed within the instant application, Bremsstrahlung refers to electromagnetic radiation produced by high energy electrons deflected (decelerated) in an electric field of another charged particle, such as an electron or atomic nucleus. A Bremsstrahlung converter is a material that produces Bremsstrahlung radiation when high energy electrons strike the converter, thereby converting electron energy into photon energy. Typically, converter plates are made from materials with high atomic numbers as the Bremsstrahlung radiation's efficiency increases with the energy of incident electron, the atomic number, and thickness of the target material. For example, tungsten and tantalum both have relatively high electron-photon conversion rates, high melting points, and may withstand high electron power densities. Therefore, converter materials include, but are not limited to, tungsten, tantalum, or heavier metals, such as osmium. Independently, each of the at least three converter plates may have a thickness between 0.75 mm to 3.0 mm and a minimum plate separation between 1 mm and 4 mm. The use of three converter plates may generate sufficient Bremsstrahlung radiation that impacts the zinc-68 target over a period of time. For example, and without limitation, the period of time may be at least 1 hour. As specific activity of a radioisotope is typically reported in units of activity per unit mass (Curies/gram), the above configuration and dimensions may produce a solid mixture with a measured activity of copper-67 of at least 1 μCi/g-target. Alternatively, if a single converter plate is used, the plates may have a thickness in the range of about 2 mm to about 8 mm about, 2 mm to about 5 mm, or about 2 mm to about 3 mm. Similarly, the converter plates may be tungsten, tantalum, tantalum-coated tungsten plates, or another heavy z-metal, such as osmium. The multi-converter plate may be designed to stop high energy incident electrons and prevent the photo-production target from excessive heating by absorbing low energy electrons that may otherwise deposit thermal energy into the metal target. To minimize the pass through of electrons and subsequent heating of the metal target, it is possible to increase the total thickness of the converter. However, an increase in converter thickness may result in a decrease in the average photon flux at the energy of maximum production within the zinc-68 target and may produce lower photo-yields of the desired radioisotope, for example, copper-67. Alternatively, a decrease in converter thickness may increase photon flux at the energy of maximum production in the metal target but may increase target heating by allowing a large portion of the electrons to penetrate through the converter plates and into the target. The spacing between the converter plates may be organized to keep the turbulence of the water flowing between the plates in order to maximize the heat removal. The thicker the plates or, alternatively, the thicker the water between the plates may potentially shield the gamma rays and decrease the overall yield of the metal target. For example, it may not be suitable to have different spacing between the plates, respective of one another, because the water may find the least resistance within the assembly and the narrowly spaced plates may have insufficient water flow and overheat. Furthermore, the farther away the metal target is from the converter, the less “concentrated” the gamma-ray beam may become. Consequently, the concentration of the gamma-ray energy may be represented by 1/R2, where R is the distance away. While multiple converter plates provide heat removal by water, or similar coolant fluids, further embodiments may rid such features as water in the converter may block the electrons and gamma rays targeted for the metal target. Accordingly, converter designs such as a liquid metal converter made from lead bismuth eutectic (LBE) may be incorporated into the target assembly. Therefore, such converter designs may run at high power and dissipate heat without water flow in the converter. An optimum Bremsstrahlung converter design, for example, is one that may produce maximum high energy Bremsstrahlung photons above 10 MeV for photo-nuclear reactions. A Bremsstrahlung converter may have multiple stacked converter plates in series to improve production yields of copper-67. If multiple stacked converter plates are used, the total thickness of all converter plates may have a thickness in the range of about 2 mm to about 8 mm, about 2 mm to about 6 mm, or about 2 mm to about 4 mm. The separation of electron linear accelerator (linac) generated copper-67 from a bulk zinc-68 target may require separating small, near trace amounts of copper-67 from the solid target mixture. For example, the mass of non-converted zinc-68 in the solid mixture may be seven to nine orders of magnitude greater in mass than the copper-67. Accordingly, the separation process may account for the dilution of copper-67 in the solid mixture, and therefore, minimize the loss of the very small amounts of copper-67 in the solid mixture. In other words, one of the technical problems that may be addressed is the select separation and isolation of the small amounts of copper-67 in the solid mixture. Photonuclear production of copper-67 using an electron linac represents an alternative to proton and neutron induced production methods. When high energy photons are absorbed by a nucleus of the target material, the nucleus becomes unstable. The unstable nucleus may then release excess energy in the form of one or more particles, e.g., proton, neutron, α, β, or γ, etc., and decays to a lower energy state. This process may be expressed as: T+γ→P+b, wherein, T represents the target nucleus, γ the incident radiation particle, e.g., a gamma ray, P the product nuclide, and b, the emitted particle. As stated, the use of electron linacs for copper-67 production is a convenient and relatively inexpensive alternative to nuclear reactor and proton accelerators. The described sublimation and purification methods may have copper-67 recovery yields of at least 80%, at least 90%, at least 95%, or at least 98%. The process of copper-67 production and improved efficiencies in the separation and isolation of copper-67 described may address and provide answers to many technical and commercial issues. To produce high concentrations of copper-67 in an irradiated target, the power of the electron beam or the irradiation time may be increased. Therefore, a relationship may exist between the average power of the electron beam used to irradiate a zinc-68 target and the time of irradiation in an irradiated target. Additionally, in a photon induced reaction, the yield of radioisotopes may be increased by increasing the production rate, which may, in part depend, upon the photon flux, the number of target atoms, and the cross section of the radiation induced photo-nuclear reaction. The number of Bremsstrahlung photons created using an electron linac may depend on electron beam parameters, such as electron energy, current, beam divergence, and beam size. Likewise, different converter materials and design may affect efficiencies of electron to photon conversions. Additionally, a metal target may not be only subjected to irradiation from gamma rays, but also irradiation from gamma and electrons of insufficient energy to cause photo-conversion. Moreover, a photo-production metal target may be also subjected to irradiation from primary and secondary electrons from the electron beam. The metal target may therefore absorb power delivered by both incident electrons and photons, resulting in additional heat generation. Accordingly, there exists an appropriate balance between electron beam power and the amount of heat dissipated. Also, since radioisotope yields are proportional to the incident beam power, the power density from gammas and secondary electrons in the target may be maximized to increase isotope production yield. Still, the target may be limited to maximize the photon flux within the target. In both instances, the maximization of electron beam power and photon flux may increase the thermal power density in the target. Accordingly, if the melting point of target materials is low, the target may melt, and in some instances boil, unless sufficient cooling exists. To optimize and balance irradiation parameters, Monte-Carlo simulations may be used to model both the optimum photon flux and power deposition into a zinc-68 target. The photon flux may be calculated by utilizing standard MCNPX volume-averaged flux for a 40 MeV electron energy beam. Also, the total energy deposited by electrons and photons on the metal target and converter plates may be simulated using energy deposition tally of MCNPX. In an embodiment, because of its high atomic number, high density (19.3 g/cm3), and very high melting point (3422° C.), tungsten may be selected for the converter material of converter plates. It may also be advantageous to coat converter plates with tantalum to impart additional chemical stability. While there may exist no limit to production yields regarding how high an average beam power may be used, practical limitations exist to prevent partial melting, or perhaps, partial vaporizing of the target. For a zinc-68 target of about 30 g to about 50 g, a suitable average beam power may be tens of kilowatts, e.g. from about 5 kW to about 40 kW. To optimize photon flux distribution through the zinc-68 target, the appropriate size and shape of the target may maximize the integral flux, and correspondingly, the overall photo-yield of copper-67. A study in system parameters such as the beam energy (MeV), current (microAmps), and hence the power of the beam (kW), may maximize the photo-yield. However, there exists an operational balance to control the electron beam because of heat generation in the zinc-68 target. Therefore, a target assembly equipped with a cooling design may moderate the anticipated increase in temperature of the target, converter, and/or assembly. A given target metal is commonly composed of many isotopic species. For example, the isotopic amount of zinc-68 in natural zinc may be about 19%. The desired radioisotope generated by the photo-production process is the result of a gamma photoreaction with a specific isotope of the target metal. As indicated, in the case of the photoreaction to produce copper-67, the target metal is zinc, and the isotope of interest is zinc-68. If natural zinc is irradiated in a photo-production process, the other isotopes of zinc may be converted into unwanted or contaminating species, some of which may be radioactive. Accordingly, the use of a target enriched in the isotope of interest, i.e., the isotope that is converted to the desired radioisotope, may result in an increase in photo-yield and a reduction in contaminating species. However, isotopic enriched targets may be expensive and since only a small portion of the target metal may be converted into the desired radioisotope, it may be necessary to develop a process to recover the unconverted enriched target metal. For example, photoreaction using Bremsstrahlung may often convert a small amount of a target isotope to the radioisotope of interest—as little as nanograms of radioisotope per gram of target. For these reasons, the zinc-68 target to be irradiated may be enriched in zinc-68 by at least about 90%, at least about 95%, and even at least about 99%. For example, the zinc-68 target obtained may comprise an enrichment of greater than 95% zinc-68, greater than 97% zinc-68, greater than 99% zinc-68, greater than 99.9% zinc-68 or even greater than 99.99% zinc-68. It may also be advantageous for the zinc-68 target to have trace copper impurities removed in order to minimize the amount of cold copper (non-copper-67) recovered in the separation process following irradiation (described in further detail herein). Highly enriched zinc-68 targets that contain low levels of cold copper may be obtained by repeated sublimation of the zinc-68 target. The recycling of the zinc-68 target may have an advantage in that the amounts of cold copper and other trace metal contaminants in the enriched zinc-68 target may reduce with each successive recycle. Accordingly, when a certain amount of Cu-67 is produced, the target may have few impurities of cold copper, accounting for a higher ratio of Cu-67 to cold copper or other impurities, as compared to a target that may not have been successfully sublimated or recycled. Therefore, it may be possible to obtain radioactive copper samples for medical applications with a higher ratio of copper-67 to non-radioactive (cold) copper after each zinc target recycle stage. From a theoretical perspective, the actual mass size of the zinc-68 target to be irradiated with Bremsstrahlung may not be limited, however, from a technical perspective, the zinc-68 target may have a mass size, for example, in the range of about 10 g to about 1000 g, 80 g to 300 g, or 10 g to about 60 g. Albeit, it is understood that smaller and larger sized targets may also be irradiated. To optimize operational system parameters in the production of a high specific activity product comprising copper-67, an investigation may be conducted to determine the following: the optimum electron beam energy for a given electron linac, keeping in mind that electron beam energy also has an effect on the maximum beam current; the design of the Bremsstrahlung converter in terms of material as well as geometry to maximize photon flux within the zinc-68 target; and the zinc-68 target geometry to maximize photon flux through the target. It is understood that a change or optimization of one operation parameter may, in turn, affect at least one of the other operational parameters. Therefore, an appropriate “tradeoff” when optimizing any one operational parameter may be assessed and analyzed. Referring to the figures, FIG. 1A illustrates a cross-sectional view of target assembly 10. Target assembly 10 may comprise a front housing 12 that includes a first section 12a, a second section 12b, and a third section 12c, the latter being joined to rear housing 14. Front housing sections 12a and 12b may be assembled and disassembled to allow access to plate cavity 15 and Bremsstrahlung converter plates 16. The first section 12a of the front housing 12 may include a front window fitting to seal front target window 18a. Likewise, the third section 12c of the front housing 12 may include a rear window fitting to seal rear target window 18b. Collectively, target windows 18a and 18b may allow access to plate cavity 15, which encloses converter plates 16. Front target window 18a may be made of any material that has little or no effect on electron beam 20, which passes through target front window 18a. Similarly, rear target window 18b may be made of any material that has little or no effect on the produced gamma photons, which also passes through rear window 18b. Converter plates 16 and metal target 27 may be configured in any suitable manner within electron beam 20. To remove heat generated in converter plates 16 by the impact of electron beam 20, the second section 12b of the front housing 12 may include coolant fluid input 22. The coolant fluid, for example water, may be added through coolant fluid input 22 at a select rate (volume/min) and enter plate cavity 15 to remove heat generated in converter plates 16. After passing around or between converter plates 16, the coolant fluid may be diverted to the third section 12c of the front housing 12 through conduit 24 (the coolant flow in FIG. 1A is represented by the depicted arrows). Thereafter, the coolant flow may be directed to target cavity 26, pass through target cavity 26, and then exit out coolant fluid output 28. In this configuration, the coolant fluid may remove heat generated in metal target 27 during irradiation. While FIG. 1A depicts one embodiment of a target assembly, alternative target assembly designs may provide similar irradiation conditions, for example, photon flux or coolant flow. Furthermore, it is conceivable, for example, that another coolant design may have two separate coolant fluid inputs and two corresponding outputs. Within target assembly 10, rear housing 14 is joined to back end 29 of the third section 12c of the front housing 12. For example, rear housing 14 may be joined via welding to back end 29 at joint 30. Rear housing 14 may be mechanically configured to be sealed by back-plate target plunger assembly 32 (including a back plate that opens and closes by the plunger), which in turn may be releasably attached to target crucible support 33. Target crucible support 33 may be mechanically configured to releasably attach to and from target crucible 34. Accordingly, one may mechanically manipulate back-plate target plunger assembly 32 to position target crucible 34, and hence, the metal target 27 in and out of target assembly 10. In one embodiment, rear housing 14 may be cylindrical. Metal target 27 may be configured in any geometric form for irradiation. For example, metal target 27 may be configured in the form of one or more plates or a solid cylinder. Metal target 27 may be positioned in target crucible 34 and then positioned within a target assembly 10, thereafter being irradiated with gamma rays produced by converter plates 16. The gamma rays may have an intensity of at least about 1.5 kW/cm2 to about 20 kW/cm2. For example, an arrangement of converter plates 16 may produce gamma rays with an intensity of from about 3 kW/cm2 to about 14 kW/cm2 or from about 3 kW/cm2 to about 8 kW/cm2. FIG. 1B illustrates a cross-sectional view of another embodiment of a target assembly 10B. Similar to target assembly 10 in FIG. 1A, target assembly 10B includes a front housing 12 that may include a first section 12a, a second section 12b, and a third section 12c, the latter being joined to rear housing 14B. The first and second sections, 12a and 12b of housing 12, may be assembled and disassembled to allow access to target cavity 26B and converter plates 16. First section 12a of housing 12 may include a front window fitting to seal a front target window 18a. Target assembly 10B may also include a coolant flow system (not shown, but may be similar to the system of coolant fluid input 22 in FIG. 1A) for cooling the converter plates 16 and the metal target (not shown). For example, water or another coolant fluid may be used to remove heat from the metal target when positioned in target cavity 26B. The coolant fluid may be contained within target housing 17. Target housing 17 may also include an input and output (not shown) so the coolant fluid may flow into housing 17, around a metal target positioned in target cavity 26B, and exit out of housing 17. An arrangement of three in series converter plates 16 shown in FIG. 1B may produce gamma rays with an intensity of from about 4 kW/cm2 to about 6 kW/cm2. In one example, converter plates 16 made of tungsten may be irradiated with an electron beam (such as electron beam 10 of FIG. 1A) having a beam energy in the range of about 25 MeV to about 100 MeV, e.g., 35 MeV to 55 meV, and a beam current in the range of about 30 microAmps to about 280 microAmps, e.g., 50 microAmps to 140 microAmps. The irradiation of converter plates 16 with the electron beam may result in the production of gamma rays with energies in the range of about 1 MeV to about 55 MeV, e.g., of about 1 MeV to about 40 MeV. For example, in some instances and for medical applications, the irradiation may be continued until the conversion to copper-67 yields a copper-67 total activity of at least about 2 μCi/g-target, at least about 5 μCi/g-target, at least about 10 μCi/g-target, or at least about 20 μCi/g-target. For example, when using a 40 g zinc-68 target (such as target 27 (FIG. 1A)), one may irradiate the metal target with Bremsstrahlung-produced gamma rays for a time until at least about 80 μCi of copper-67, at least about 400 μCi of copper-67, or at least until about 800 μCi of copper-67, is produced. In one instance, for example, one may irradiate a 40 g zinc-68 target with Bremsstrahlung-produced gamma rays for a time until from about 500 μCi to 500 mCi of copper is produced. Alternatively, a target assembly may provide a yield of copper-67 of at least about 5 μCi/g-target-kW-hr of beam energy, at least 20 μCi/g-target-kW-hr of beam energy, or at least about 50 μCi/g-target-kW-hr of beam energy. Irradiation times may be, for example, in the range of about 1 hour to 260 hours, 10 hours to 140 hours, or 40 hours to 96 hours. FIG. 1C illustrates a target holder 40 for insertion within target cavity 26 of FIG. 1A or target housing 17 of target assembly 10B, for example. Target holder 40 may hold a metal target, such as metal target 27 in FIG. 1A, contained within a target crucible, such as target crucible 34. Although depicted as cylindrical, the target holder 40 may take any shape. In use, a target crucible containing a metal target may be positioned within internal volume 41 of target holder 40 and may be held in place with threaded plug nut 42. Target holder 40 may also include any number of cooling fins 44 to facilitate the transfer of heat from the metal target to the coolant fluid that flows through target cavity 26 of FIG. 1A or target housing 17 of FIG. 1B. FIG. 2 is a schematic of an electron linear accelerator (linac) that may be used for producing photonuclear copper-67 from a zinc-68 target. Depicted within FIG. 2, quadrapole magnets (denoted 10 cm long QM (Quad2a)) may be used to help focus the electron beam down an axis of the accelerator. Dipole magnets may be included to allow the electron beam to be turned, and therefore to determine the energy of the electron beam. Additionally, moveable screens may be used to determine the size of the beam inside the accelerator tube. Corresponding to the schematic illustrated in FIG. 2, the total unloaded output energy may be about 50 MeV, with an energy reduction 0.118 MeV/microAmps of peak beam current after beam loading. Seen in FIG. 3, taking into account that at higher energies, the photo-produced copper-67 yield does not increase linearly with increase in electron energy, and considering the load characteristics of a pulsed electron linac, it may be possible to operate the electron linac in an optimal energy range for an optimal irradiation time. In one embodiment, the optimization may be done to produce a yield of copper-67 suitable for medical applications. For example, a cylindrical zinc-68 target that is about 2.5 cm in diameter, about 2.8 cm in length, and with one end of the cylindrical target facing the converter plates (converter plates 16 in FIG. 1A), a peak beam current may be calculated using a beam load function for different beam energies. Assuming an average duty factor of 0.1%, which is the fraction of time the beam is “on,” the average beam current and average beam power may be determined from the peak beam current for a given electron beam energy. For example, at 40 MeV of loaded beam energy, the peak beam current is 104 microAmps. Considering 0.1% duty factor, the average current is 104 microAmps. For these given electron beam parameters, the average power was found to be: Pavg=40 MeV×104 microAmps×0.1%=4.16 kW. Accordingly, the average power of an electron beam striking a converter for the photonuclear conversion of zinc-68 to copper-67 may likely be in the range of 3 kW to 8 kW. A MCNPX simulated photon flux through a 40 g cylindrical zinc-68 target may be used to calculate the average activity yield of copper-67 at various beam energies and corresponding beam currents. Photon activations on zinc targets may be performed at various beam energies followed by gamma spectroscopy. The optimal current and energy of electron beam may be determined based on the highest activity yield of copper-67. According to the measured activity values, optimal beam energy for the photo-production of copper-67 may be about 38 MeV. The values may be compared or measured against Monte Carlo simulation results to determine their agreement. Using Monte Carlo simulations, the optimum photon flux in a 40 g cylindrical zinc-68 target, the heat deposition in the target, as well as the converter with an electron beam of 40 MeV energy at 25 microAmps average beam current using several different converter designs and various thicknesses, may be investigated. For example, the optimum photon flux yield may peak using a 1.5 mm thick converter and gradually drop with increasing thickness of the converter. An increase in the converter thickness may also result in a corresponding decrease of the energy (heat) deposited into the target. However, there may exist a relationship between photon flux, converter design and thickness, and heat generation in the target. For example, an increase in the thickness of the converter from 1.5 mm to 4.5 mm may cause the optimum photon flux to decrease by about 18% with a corresponding 41% drop in the energy generation within the target. Considering the possible melting of a target posed by large amounts of heat generation some yield of copper-67 may be forgone in exchange for lower heat generation. As represented by FIG. 4, following the gamma irradiation of the target (metal target 27 in FIG. 1A) and allowing sufficient time for some of the relatively short-lived radioisotopes of copper to decay to near background levels, the target may be positioned in a sublimation apparatus 50. Sublimation apparatus 50 may include sublimation vessel 52, heating element 54, and translation stage 56 to vertically position heating element 54 and sublimation vessel 52 relative to the position of collection vessel 58 and crucible 60 containing solid mixture 62. Sublimation apparatus 50 may also include vacuum port valves 64, vacuum gauge 66, and inert gas port 68. After sublimation apparatus 50 is assembled, sublimation vessel 52 may be evacuated and back-filled with an inert gas, for example, argon, using inert gas port 68. Similarly, helium or nitrogen may be used. The purge/vacuum cycles, through vacuum port valves 64, may be used to remove trace levels of oxygen in sublimation vessel 52 prior to heating in order to minimize oxidation of zinc-68 to a zinc-68 oxide. Within the interior of sublimation vessel 52 is solid mixture 62, which as shown, is contained within crucible 60. Crucible 60 may be supported in sublimation vessel 52 with support 70. Positioned above crucible 60, and hence above solid mixture 62, is collection vessel 58. Sublimation vessel 52, as depicted, is represented as a cylindrical hollow tube, and may be made of quartz, though sublimation vessel 52 may similarly be made of a metal, e.g., titanium, or a ceramic oxide. In one embodiment, an advantage of making sublimation vessel 52 out of quartz is that an infra-red detector may be used to measure the temperature of collection vessel 58 during the sublimation heat cycle. The monitoring of and, if necessary, adjustment to, the temperature of the collection vessel 58 may optimize the fill efficiency of the target metal within the internal volume (discussed later) of collection vessel 58 during the sublimation heat cycle. In the sublimation of a solid mixture 62 containing zinc-68, or another target metal, copper-67 and other trace metals, for example, zinc-68 may have a greater vapor pressure than that of copper-67 at a given temperature and pressure. Accordingly, in the described sublimation process, the zinc-68 of solid mixture 62 may be selectively converted into the vapor phase upon heating by heating element 54. The zinc-68 may then condense in sublimation apparatus 50, and the copper-67, and optionally other trace metals, are retained in solid mixture 62. In one embodiment, an advantage to the separation process described herein is the manner in which zinc-68 may condense from solid mixture 62 within sublimation vessel 52. Under most sublimation conditions for a given temperature and pressure, at least about 95% or 98% or greater of the zinc-68 may be removed from the solid mixture by sublimation. For example, at least about 99.9%, even at least about 99.99%, on a weight basis of the zinc-68 in solid mixture 62 may be separated by sublimation. The copper-67 that remains in the solid mixture 62 may be further purified by chemical means, for example, by dissolving solid mixture 62 in an aqueous inorganic acid to form an acidic solution of metal ions. The copper-67 may then be separated from other trace metals by a metal-ion exchange. The zinc-68 sublimate may thereafter be recycled for use in another enriched target, and the process of producing copper-67 may therein be repeated, as discussed previously. Alternatively, copper-67 produced in the gamma irradiation of a zinc-68 target may be separated from solid mixture 62 at temperatures in the range of about 400° C. to about 700° C. in an environment of reduced pressure. The environment of reduced pressure in sublimation apparatus 50 may be created under a dynamic vacuum, using vacuum port valves 66, rather than static vacuum. However, it is understood that either type of vacuum may be used. Also, an exemplary range of pressures of the evacuated sublimation vessel 52 may be about 1 mbar or less (e.g., about 10−6 mbar). Using vacuum gauge 68, the pressure may be determined. Collection vessel 58, as depicted in FIG. 4, subpart A, and in FIG. 5, may be described as a one piece or multiple-piece unit vessel (each being described herein) that defines an internal volume of any shape, for example, a cylindrical or cone-shaped vessel that fits within sublimation vessel 52. Collection vessel 58 may be used as a receptacle in which sublimed metal vapors from a heated solid mixture may condense. As stated above, sublimation apparatus 50 may include translational stage 56 to position sublimation vessel 52, and optionally, heating element 54 to a location over crucible 60 and collection vessel 58. Once the components of sublimation apparatus 50 are in appropriate positions, sublimation vessel 52 may be secured and sealed. This may occur, for example, by using a high vacuum O-ring 72, located at the bottom of the sublimation vessel 52 and a vacuum source (Indicated in FIG. 4). Following the heating step of the process, translational stage 56 may also be used to move sublimation vessel 52, and optionally, heating element 54, away from collection vessel 58 and crucible 60 that contains the remaining solids of solid mixture 62. For example, translational stage 56 may be in a vertical relationship to the sublimation vessel 52, via a sublimation support assembly 74, such that the stage in connection with sublimation vessel 52 may both lower and raise sublimation vessel 52 over collection vessel 58 and crucible 60. Moreover, translational stage 56 may be in connection with heating element 54 to lower and raise heating element 54 over collection vessel 58 and crucible 60. Sublimation apparatus 50 may further include a control unit 76 that receives or monitors temperature data of solid mixture 62, crucible 60 that contains solid mixture 62, support 60, and pressure data, through vacuum gauge 68, within sublimation vessel 52. For clarity, the lines between control unit 76 and subpart A indicate that the control unit 76 is in data communication with the sublimation apparatus 50. Such communication, for example, may be through hardwire or wireless sensors (not shown) providing the data. Based on the given temperature data, the control unit 76 may also be used to automatically adjust the operating temperature of heating element 54, and thereby adjust the temperature of solid mixture 62 as well as at least a portion of collection vessel 58. The control unit 76 may receive temperature data of collection vessel 58. The ability to adjust and maintain temperature of the different components, e.g., solid mixture 62 or collection vessel 58, during sublimation may help prevent or minimize the formation of a zinc-68 “plug” in the lower half of collection vessel 58 before a significant portion of zinc-68 metal is sublimed from solid mixture 62. Accordingly, the temperature of the various components may be controlled so to control the condensation rate of the metal vapor within collection vessel 58. FIG. 5 illustrates a cross-sectional representation of collection vessel 58 that may be used as a receptacle for collected condensed vapors of zinc-68 from solid mixture 62 when heated by heating element 54 to an appropriate temperature and environment of reduced pressure within sublimation vessel 52. Collection vessel 58 is depicted as a cylindrical form that fits within sublimation vessel 52, which happens to have a cylindrical form. Although both the collection vessel 58 and the sublimation vessel 52 are depicted as cylindrical in shape in FIG. 5, one vessel shape may be independent of the other vessel shape. Stated another way, collection vessel 58 and sublimation vessel 52 need not have identical nor similar shape, and may have very different shape forms that define different internal geometric volumes. As shown, collection vessel 58 may have an internal volume 80 with internal sidewall 84, on which sublimed metal vapor may condense. As a metal vapor flows into the internal volume 80, the metal vapor passes along the vessel and eventually contacts a relatively cool internal sidewall 84 of collection vessel 58 and condenses along internal sidewall 84 to form sublimed metal 82. During the heat stage of the sublimation process, collection vessel 58 is positioned within both heat volume portion 86 and warm volume portion 87 of sublimation vessel 52. The temperature of the heating element 54 is maintained at a temperature that may allow the metal vapor to condense within collection vessel 58 at a rate that does not clog a lower portion of collection vessel 58 with sublimed/condensed metal 82 with significant amounts of target metal yet to be sublimed in solid mixture 62. Accordingly, there exists an optimal temperature at which to maintain heating element 54, and, the temperature along the length of internal volume 80 of collection vessel 58. Condensed metal 82 may further be prevented from traversing downward on the internal sidewall 84 via adhesion with internal sidewall 84 and tension within the condensed metal 82. As shown, collection vessel 58 may include an upper end 88 and opposite, open end 89, with internal sidewalls 84 extending from upper end 88 to opposite open end 89, thereby forming internal volume 80 of collection vessel 58, wherein vapor of the target metal may condense upon internal sidewall 84. For example, collection vessel 58 may have internal volume 80 of sufficient size to hold 10 g to 1 kg of zinc metal. Also shown, open end 89 may be configured to engage and fit with an open end of crucible 60. The configured fit may not necessarily have to be a tight or a sealed fitting between open end 89 and crucible 60. However, the snugness of the fit may minimize the escape of sublimed metal 82 into a volume of sublimation vessel 52 before the vapor has an opportunity to move up collection vessel 58 and condense within internal volume 80. In metal-metal separations using the described sublimation process, at least one metal to be sublimed from a solid mixture may have an appreciable vapor pressure at a temperature from 300° C. to 700° C. at a nominal pressure of 10−5 mbar within sublimation vessel 52. Collection vessel 58 may have internal sidewalls 84 that are cylindrical, as depicted in FIG. 5, or may be shaped as a truncated cone, or flat or scalloped elongated segments that combine to form geometric volume 80. Collection vessel 58 may be made of any material that is thermally stable to temperatures of at least about 800° C. For example, suitable materials may include a metal such as titanium, a ceramic oxide that is stable at temperatures greater than 600° C., or graphite. In one embodiment, graphite may be a material of particular interest because of its inherent lubricity and thermal stability. One advantage of collection vessel 88 being made of graphite is that condensed metal 82, and in particular, condensed zinc-68, may more easily be removed from internal volume 80. Particularly, the condensed zinc-68 may be recovered from collection vessel 58 by sliding collection vessel 58 off of condensed metal 82, thereby leaving a zinc-68 target slug that is easily refitted, e.g., by melting into a crucible and positioned back into a target assembly for irradiation. Following irradiation, the sublimation process may be repeated and again the condensed zinc-68 slug may be returned to the target assembly 10 or 10B. To make the final conversion of zinc-68 to copper-67 more efficient, any number of repeated irradiation and recovery cycles are possible, which may make the process efficient in terms of final conversion of zinc-68 to copper-67. Crucible 60 may be made of materials that are stable at high temperatures. For example, suitable materials for crucible 60 may be materials stable at temperatures to at least about 900° C. including, but not limited to, a ceramic oxide, a metal, or graphite. Crucible 60 may also be used to shape the enriched zinc-68 shot into a select geometric form of the zinc-68 target. For instance, the commercial shot of zinc-68 may be placed in crucible 60 and therein positioned in a melt furnace, or an alternative environment that may be purged of trace amounts of oxygen to minimize the formation of zinc oxide during the melt stage. As shown in FIG. 5, crucible 60 is a high temperature stable cup with an open end and an opposite closed end. Like a cup, crucible 60 may adopt a cylindrical form or any geometric form including a truncated cone form. In many instances, the geometric volume form of the zinc-68 target may adopt the interior geometric volume form of crucible 60, if the same crucible is used to both prepare the target and contain the zinc-68 target in a target assembly. Further, in some instances, crucible 60 may include any number of exterior cooling fins (not shown) to facilitate the cooling of the zinc-68 target during the irradiation. Following the sublimation of zinc-68 from solid mixture 62, the copper-67 residue that remains in crucible 60 may be isolated from other trace metals by dissolution in an acid (e.g., a mineral acid such as sulfuric acid, hydrochloric acid, phosphoric acid, nitric acid, or a combination of mineral acids), followed by ion exchange with a selective copper ion exchange resin (e.g., a quaternized amine resin) or a chelating agent immobilized on an ion exchange resin or silica substrate. In one embodiment, the copper-67 residue may be dissolved in hydrochloric acid and the resulting aqueous solution passed through a quaternary amine ion exchange resin. The non-copper trace metals in the acid solution may pass through the column at a very low pH. After passing through a low pH aqueous solution, the pH of the flush solution is raised to release the copper from the exchange resin. The collected solution is thereafter evaporated to dryness, leaving a copper-67 radioisotope. In one embodiment, the copper-67 left may be suitable for shipment, or for molecular complexation as a radiopharmaceutical for medical or research applications. Illustrated in FIG. 6 is a cross-sectional representation of collection vessel 90 that includes two separable portions including a first portion 92 having a first upper end 93, which may be a closed end, and an opposite, first open end 94. First portion 90 may include first internal sidewall 96 extending from first upper end 93 to opposite, first open end 94. Second portion 98 of collection vessel 90 may include a second, upper open end 100 configured to engage and fit with opposite, first open end 94 of first portion 92. Second portion 98 may also include an opposite, second open end 101, and second internal sidewall 102 extending from second, upper open end 100 to second, opposite open end 101. If combined in an elongated manner, first internal sidewall 96 and second internal sidewall 102 may define first internal volume 104a and second internal volume 104b, respectively, forming collection vessel 90. A multiple-piece collection vessel 90, as shown in FIG. 6, may have any number of divided portions. For example, a collection vessel may have six portions that extend from a portion proximate to the crucible and a portion proximate to the closed end of the sublimation vessel. FIG. 6 exemplifies and describes a two-piece collection vessel. In one embodiment, the advantage of dividing collection vessel 90 into at least two portions may be to facilitate the removal of the condensed metal that forms within the internal volume 104a and 104b of multiple-piece collection vessel 90. The following examples, while, in addition to referring to the subsequent figures, are put forth so as to provide a complete disclosure and description of how the articles and methods described and claimed are made and evaluated. They are intended to be purely exemplary and are not intended to limit the scope of what the inventors regard as their invention. Efforts have been made to ensure accuracy with respect to numbers (e.g., amounts, temperature, etc.) but some errors and deviations should be accounted for. Unless indicated otherwise, parts are parts by weight, temperature is in ° C., or if not stated, the temperature at which the experiment or measurement is conducted is about room temperature. Pressure is at, or near, atmospheric unless stated otherwise. There are numerous variations and combinations of reaction conditions, e.g., component concentrations, desired solvents, solvent mixtures, temperatures, pressures and other reaction ranges and conditions that may be used to optimize the product purity and yield obtained from the described process. Referring to FIG. 11, which shows a method 1100 of producing an irradiated metal target. For example, to prepare a sample zinc-68 target, about 40 g of about 98% to about 99% enriched zinc-68 may be melted into an alumina (Al2O3) crucible, as an example of step 1102. Alumina is one of many select materials that may be used as a crucible material because of its hardness, low porosity, and high melting point (2072° C.). The crucible may thereafter be positioned within a tube furnace having an 8 inch hot zone with a 2 inch diameter alumina casing for fully enclosing a slightly smaller diameter furnace tube. An inert gas line may connect to one end of the furnace tube. During the heat or melt cycle of the enriched cycle, the crucible may be blanketed in argon to minimize oxidation of zinc. The target may therein take the form of the crucible. The furnace tube may be cycled under vacuum to 20 to 30 mbar and flushed with argon for at least two cycles prior to heating the furnace tube. A high quality zinc target (i.e., natural refined zinc to remove metal contaminants to 99.9999% zinc or better) or a 98% or better enriched zinc-68 target may be melted into an approximately 7 mL volume alumina (98% Al2O3 by weight) crucible with an approximate outside diameter of 25 mm. In one embodiment, it may be advantageous if the resulting zinc “slug” inside the crucible is essentially free of voids or there is little or no zinc oxide (ZnO) coating or ZnO embedded in the slug. To minimize the amount of ZnO in the target crucible, the zinc melt may be poured into the crucible through a specially designed funnel at a temperature between about 500° C. and about 550° C. Pouring may also occur in an argon blanket environment. The funnel may be made of graphite and may be about 7 cm long, 3.5 cm in diameter, and have an orifice of approximately 5 mm in diameter. The design of the funnel, and in particular the orifice, may minimize the incorporation of ZnO from entering the target crucible. Instead, the ZnO may float on top of the zinc melt in the funnel as the zinc slowly fills the crucible from the orifice. Thereafter the ZnO may be collected on the angular surface of the funnel, removed, and may be refined by sublimation to recover additional zinc-68 material. After being positioned in a target assembly, step 1104, and directing an electron beam at converter plates, step 1106, the metal target may be irradiated using a 48 MeV, 10 kW electron linac in step 1108. The alumina crucible containing the zinc-68 target obtained from the melted shot may be positioned a few centimeters, e.g. from 3 cm to 6 cm, from the last converter plate. The converter plates may comprise three water-cooled tungsten plates in series, each having a thickness of 1.5 mm and separated from one another by 3 mm. See FIG. 1, for example. Accordingly, cooling the converter plates in step 1110 may help remove heat generated during irradiation. Fast moving electrons from the linac may then strike the converter plates and produce Bremsstrahlung photons as the electrons decelerate within the series of converter plates. The zinc-68 target may irradiate with Bremsstrahlung photons for 1 hour to 260 hours, 1 hour to 180 hours, 1 hour to 80 hours, or 5 hours to 60 hours. Following the irradiation of the target, in a further embodiment, the target may be retained in the target assembly for a time sufficient for the short-lived radioisotopes to decay to background level so that the radiation exposure to working personnel is in agreement with safety limits. To determine if sufficient copper-67 is produced within the zinc-68 target, the irradiated target may be analyzed for activity of various radioisotopes using a photon spectrometer. Different dimensions and masses of a cylindrical zinc target may be investigated. The radius to length ratio of the cylinder that produces the highest activity yield of copper-67 was assessed by comparing results for all possible values of radius. FIG. 7 illustrates a plot that describes the radius to length ratios for 40, 60, 80, and 100 g zinc-68 targets and corresponding copper-67 activity yields. As illustrated, the optimized radius and length for a 40 g zinc target may be found to be 0.8 cm and 2.8 cm respectively. Moreover, the optimal radius to length ratio for a cylindrical target may be about 0.18 to about 0.32 for many mass targets at the given electron energy and electron beam energy of 40 MeV and 1 kW, respectively. For example, it may be advantageous to prepare a zinc-68 target with a radius to length ratio in the range of 0.18 to 0.25. Alternatively, the optimal radius to length ratio may change with a corresponding change in electron beam operational parameters. In one embodiment, for the electron linac used in the photo-production of copper-67, an optimum electron beam energy may be from about 38 MeV to about 42 MeV. In order to maximize the Bremsstrahlung photon yield, while minimizing the zinc target heating, different converter thicknesses may be used. The converter geometry may be found to have three tungsten discs, each with a thickness of 1.5 mm and separated from one another by 3 mm. For example, for a 40 g zinc-68 target cylinder having a radius of 0.8 cm and a length of 2.8 cm in length and an electron beam of 38 MeV, the estimated activity of copper-67 is 16 μCi/g-target-kW-hr. Experimentally, for a cylindrical zinc target of radius 0.9 cm and length of 2.2 cm, the measured copper-67 activity may roughly be 12.4 μCi/g-target-kW-hr. Generally, it may be discovered that the experimentally measured values for the activity of copper-67 is about 20-30% less than the Monte Carlo simulated values. Following the irradiation of the zinc-68 target described herein, (as in method 1100 of FIG. 11) to allow high energy short-lived radioisotopes to decay to safe levels for handling, the target holder may be removed from the target assembly and placed in a lead pig. A sheet of lead glass may be positioned in front of the lead pig. After a minimum of 30 minutes, using laboratory tongs, the target holder containing the target crucible may be removed. Shown in FIG. 12, and method 1200, the target crucible may thereafter be removed and positioned on the support stand of a sublimation apparatus in step 1202. Once a two-piece collection vessel is positioned atop the target crucible, and subsequently a crucible stand, a crucible, collection vessel, and a stepper motor controller may be used to lower the heating element and an attached quartz sublimation vessel over the collection vessel and supported crucible, as in step 1204. The sublimation vessel may be lowered until a lower open end of the vessel contacts an O-ring on the vacuum assembly. A mechanical vacuum pump may be used to achieve a vacuum of about 2 mbar, at which time, a vacuum turbo pump may be used to lower the pressure to less than 1×10−4 mbar, as in step 1206. The vacuum system may be checked for leaks by closing the inlet valve to the turbo pump and verifying the vacuum leak rate is less than a certain rate, for example, 1×10−2 mbar/min. A furnace control program controlling the heating element, may thereafter be initiated to bring the furnace temperature to 150° C., as in step 1208. While the furnace temperature is maintained at 150° C., the vacuum valve to the turbo pump is closed, and the sublimation vessel is purged with argon. The argon valve is then closed and the vacuum opened to reapply a vacuum of about 2 mbar to the sublimation vessel. The argon/vacuum process may be cycled 3 times to remove trace amounts of oxygen gas within the sublimation vessel. Following the third purge cycle, or subsequent cycles, the valve to the turbo pump is opened and the pressure is reduced to less than 1×10−4 mbar. Again, the leak rate of vacuum may be verified to be less than 5×10−3 mbar/min. Additionally, it is contemplated that pressure may be monitored within the collection vessel and/or the sublimation vessel. Thereafter the furnace control program brings the temperature of the furnace to 600° C. and maintains such temperature during the heat cycle. For example, the sublimation run may increase the temperature from 150° C. to approximately 600° C. at a ramp rate of 6 degrees per minute. The rate may be adjustable so as to not crack the components of the apparatus depending on the choice of materials. For example, increasing temperature at high rates may crack materials made of alumina. However, in one embodiment, materials made of quartz may not be as susceptible to higher rates. The total time of the sublimation heat cycle at 600° C. time is approximately 2 hours and 15 minutes, followed by a rapid cool down, as in step 1210. Starting from a 40 g zinc target, the amount of zinc remaining in the crucible may be less than 20 mg after sublimation. Accordingly, a majority of the zinc may condense within the collection vessel during the heating stage. Additionally, throughout the heat cycle, the vacuum may be monitored to ensure the vacuum is less than 1×10−4 mbar. The majority of the zinc may be captured in the collection vessel and thus, may be used to repeat the irradiation process. Afterward, the vacuum valve is closed and the sublimation vessel is slowly back-filled with argon to raise the system pressure to 0 psi, as in step 1212. In turn, the stepper motor controller of the translation stage may be used to raise the quartz tube above the collection vessel. After removing the collection vessel, it may be set aside for further zinc recovery and copper-67 production runs. The crucible containing the copper-67 may then be removed from the sublimation apparatus. Approximately 8 mL of concentrated HCl may be added to the crucible and stirred, as in step 1214. After 30 minutes, for example, the concentrated HCl is pumped onto a 10 mL, 1×8 anion exchange column. The column may be washed with approximately 10 mL of 6M HCl to elute non-copper metal ions from the column. The copper ion may then be eluted with 10 mL of 2M HCl. The 2M eluent is dried in a glass shipping vial to less than a 1 mL volume using a heated flowing flow of nitrogen gas. As indicated, the sublimed zinc-68 may be removed from the collection vessel and may be melted into a new crucible for subsequent photo-generation of copper-67, sublimation, and recovery, for example, by repeating method 1100 and 1200 of FIGS. 11 and 12, respectively. However, each sublimation and crucible re-filling may create some waste zinc that is caught in the pour funnel or escapes due to evaporation onto cooler parts of the melt furnace. The waste zinc may be collected and packed into a crucible or may also be sublimed, melted, and reused in a subsequent irradiation process, for example, method 1100 of FIG. 11. With this recovery process, the total lost zinc per each irradiation/separation/recovery process cycle may be less than 100 mg of zinc-68. Illustrated in FIGS. 8 and 9, experimental production runs may be conducted for the irradiation of a natural zinc target and a zinc-68 target. Particularly, FIG. 8 depicts the isolation yields of the separation stage, i.e., the combined sublimation and chromatography process steps, for the recovery of the copper-67 from the copper-67 in a solid mixture. With the exception of isolated technical issues, the recovery yields may be consistently greater than 90% because the implementation of the funnel fill system with production run number 13 to remove zinc oxide and tight control to prevent contaminants, such as chlorine contacting the zinc. The 60% yields obtained with production runs 20 and 24 may be caused by improper sublimation program or a failed vacuum gasket, respectively. FIG. 9, shows the rate of activity per unit of mass and power for different productions and type of target irradiated (natural or enriched). The plot shows the expected differences between a natural zinc target and a zinc-68 target. Also, the variation in copper-67 activity created from the zinc-68 target runs may be attributed to changes in distance between the converters whereby the target with longer spacing creates lower activity. For waste zinc collected from previous sublimation runs, for example in step 1210 of FIG. 12, the zinc-68 may be heated under an argon atmosphere in a tube furnace to form a melt that is poured into an alumina crucible using the described graphite funnel, as in step 1102 of FIG. 11. Specifically, the amount of zinc-68 in the crucible was 36.52 g. The crucible may therein be positioned in the target assembly and irradiated with a calculated average power of 4.125 kW electron beam for 1 hour, as in steps 1104 and 1108 of FIG. 11. Irradiation was stopped, the crucible was placed in a lead pig, and thereafter the crucible was positioned in a sublimation apparatus the next day (about 23 hours later), though a wait time of about 3 to 5 hours may be typical. The sublimation temperature was brought from room temperature to 150° C. in about 8 minutes, the temperature being measured, for example, with a thermocouple positioned on the exterior of the quartz sublimation vessel within the heat zone of the heating element. The temperature of the sublimation vessel was increase by 6° C./min and the solid mixture sublimed at 600° C. for about 2.5 hours, as in step 1208 of FIG. 12. The measured vacuum (dynamic) within the sublimation vessel was maintained at 1.9×10−5 mbar and the final leak test revealed a leak rate of 6×10−5 mbar/min, as in step 1206 of FIG. 12. After the cool down period, the collection vessel may be removed from atop the crucible. Correspondingly, the sublimed zinc-68 may be removed and collected. The collected zinc sublimate had a measured copper-67 count of 47 cps. The crucible with the retained copper-67 had a measured count of 3150 cps or a copper-67 activity of about 9000 μCi. Thereafter, 6 mL of 10M HCL may be added to the solids remaining in the crucible and after stirring for 30 min, the HCl solution may be placed atop an ion exchange column, as in step 1214 of FIG. 12. The crucible may then be washed with an additional 6 mL of 6M HCl solution and the wash solution added atop the column. A 2M HCl solution (12 mL) may be used to elute non-copper metals from the column. Additionally, a 0.001M HCl solution (12 mL) may be used to elute the copper on the column. After drying the eluent solution with a warm nitrogen stream flow, the residue had a measured copper-67 activity of about 9080 μCi, indicating minimal loss from the column purification stage. Using the irradiation and separation processes described in FIGS. 11 and 12, respectively, FIG. 10 shows the total copper-67 activity for different production runs using different average electron beam power (kW-hr) using either a natural zinc target or a zinc-68 target. As shown, several production runs may be conducted using an approximate 36 g zinc-68 target at an average electron beam power of 4.1 kW for one hour. When using a zinc-68 target, the total activity of copper-67 may range from about 4000 μCi to about 9000 μCi. Also, when using a natural zinc target (about 19% Zinc-68), the total activity of copper-67 may range from about 1200 μCi to about 2100 μCi. In an additional production run, again using an approximate 36 g zinc-68 target, the irradiation time may be increased to two hours, essentially doubling the power or energy used to irradiate the target. As shown, when the electron beam power was increased to about 8.2 kW-hr, a total copper-67 activity of about 17,400 μCi may be obtained. It should be noted that the endpoints of each of the ranges are significant both in relation to the other endpoint, and independently of the other endpoint. It is also understood that there are a number of values disclosed herein, and that each value is also herein disclosed as “about” that particular value in addition to the value itself. For example, if the value “10” is disclosed, then “about 10” is also disclosed. It is also understood that throughout the application, data is provided in a number of different formats, and that this data, represents endpoints and starting points, and ranges for any combination of the data points. For example, if a particular data point “10” and a particular data point “15” are disclosed, it is understood that 10 and 15 are considered disclosed. It is also understood that each unit value between two particular unit values are also disclosed. For example, if 10 and 15 are disclosed, then 11, 12, 13, and 14 are also disclosed. Although several embodiments have been described in language specific to structural features and/or methodological acts, it is to be understood that the claims are not necessarily limited to the specific features or acts described. Rather, the specific features and acts are disclosed as illustrative forms of implementing the claimed subject matter. |
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061817597 | abstract | A method is provided for determining the closeness to criticality of a nuclear reactor during start-up, comprising the steps of completing a control rod withdrawal step, thereby generating a change in an output signal of a neutron detector; measuring the output signal after the completion of the control rod withdrawal step and during a transient portion of the output signal; calculating the effective neutron multiplication factor (K.sub.eff) based upon the measured output signal and elapsed time between output signal measurements; and determining the closeness to criticality of the nuclear reactor based upon the calculated value of the effective neutron multiplication factor (K.sub.eff). The invention also encompasses apparatus used to perform the above method. |
045267448 | claims | 1. A fuel assembly for a boiling water reactor, comprising: a base provided with a downwardly directed inlet opening for coolant; a fuel channel fixedly attached to and extended upward from said base, said fuel channel having a vertical longitudinal axis, said fuel channel having lifting lugs attached to said fuel channel and adapted for lifting the entire fuel assembly, and said fuel channel being divided into four quadrants; a lifting plate laterally positioned within said fuel channel at an upper portion of said fuel channel, said lifting plate comprising a handling member to be grasped during a lifting operation; four bottom grids positioned within said fuel channel and supported by said base; a plurality of vertical fuel rods positioned within said fuel channel and supported at their lower ends by said bottom grids, said fuel rods comprising four separate, partial bundles of fuel rods with a partial bundle for each of said four bottom grids, each partial bundle being surrounded by one of said quadrants, the majority of the fuel rods in each partial bundle being supported on a bottom grid with freedom of movement in a direction vertically upwards, at least one of the fuel rods in each partial bundle being a tie rod attached to a bottom grid with the ability to transmit tensile force to the bottom grid; a plurality of spaces between said four partial bundles, said spaces being each provided with channel walls extending along a length greater than half the longitudinal length of said fuel rods, said channel walls extending below said lifting plate, said channel walls defining vertical flow paths for moderator water while preventing horizontal flow of said moderator water between said vertical flow paths and the interiors of said four bundles; four top grids positioned within said fuel channel and engaged with the uper ends of all said fuel rods all of all said partial bundle,said majority of fuel rods in each said partial bundle also being engaged with one of said four top grids with freedom of movement in a direction vertically upwards, said at least one tie rod being attached to a top grid with the ability to transmit tensile force from the top grid to the bottom grid; and means accessible from above the fuel assembly for releasably attaching said lifting plate to said top grids whereby said lifting plate is attached to all of said top grids, whereby upon application of tensile force to said handling member, all of said four bottom grids, fuel rods, four top grids and lifting plate can be removed as a unit from said fuel channel; or whereby upon removal of said lifting plate by releasing said means for attaching, said four partial bundles and their associated bottom grids and top grids can be removed as separate units from said fuel channel. a base provided with a downwardly directed inlet opening for coolant; a fuel channel fixedly attached to and extended upward from said base, said fuel channel having a vertical longitudinal axis, said fuel channel having lifting lugs attached to said fuel channel and adapted for lifting the entire fuel assembly, and said fuel channel being divided into four quadrants; four bottom grids positioned within said fuel channel and supported by said base; a plurality of vertical fuel rods positioned within said fuel channel and supported at their lower ends by said bottom grids, said fuel rods comprising four separate, partial bundles of fuel rods with a partial bundle for each of said four bottom grids, each partial bundle being surrounded by one of said quadrants, the majority of the fuel rods in each partial bundle being supported on a bottom grid with freedom of movement in a direction vertically upwards, at least one of the fuel rods in each partial bundle being a tie rod attached to a bottom grid with the ability to transmit tensile force to the bottom grid; a top grid positioned within said upper portion of said fuel channel and engaged with the upper ends of all said fuel rods of all of said partial bundles, said majority of fuel rods also being engaged with said top grid with freedom of movement in a direction vertically upwards, said at least on tie rod being attached to said top grid with the ability to transmit tensile force from said top grid to said bottom grid, said top grid comprising a handling member to be grasped during a lifting operation; a plurality of spaces between said partial bundles, said spaces being each provided with channel walls extending along a length greater than half the longitudinal length of said fuel rods, said channel walls extending below said top grid, said channel walls defining vertical flow paths for moderator water while preventing horizontal flow of said moderator water between said vertical flow paths and the interiors of said partial bundles; and means, accessible from above the fuel assembly, for releasably attaching said tie rods to said top grid, whereby upon application of tensile force to said handling member, all of said four bottom grids, fuel rods and top grid can be removed as a unit from said fuel channel; or whereby upon removal of said top grid by releasing said means for attaching, said four partial bundles and their associated bottom grids can be removed as separate units from said fuel channel. 2. A fuel assembly accordinglyto claim 1, wherein said fuel rods comprise upper end pins; each top grid is disposed adjacent to the lower side of said lifting plate; and said lifting plate is provided with cylindrical recesses adapted to receive said uper end pins. 3. A fuel assembly according to claim 1, wherein said tie rods comprise upper end pins; each of said top grids is formed with sixteen fuel rod positions arranged in a square lattice and comprises sixteen corresponding substantially hollow-cylindrical portions as well as a plurality of connecting portions which connect said hollow-cylindrical portions with each other and together with external surfaces thereof define a plurality of passageway openings for reactor coolant provided in said top grids, one of said connecting portions comprising a central portion directly connected to four hollow-cylindrical portions located nearest the mid-portion of said top grid, two of said four hollow-cylindrical portions lying with their axes in one and the same vertical plane through said mid-portion, said two hollow-cylindrical portions receiving said upper end pins; and said central portion being mechanically connected to said means for releasably attaching. 4. A fuel assembly according to claim 3, wherein said means for releasably attaching comprises a bolt arranged with its head in contact with an upwardly-facing surface of said lifting plate, said bolt being extended through said lifting plate and screwed with its lower end into a threaded hole provided at the center of said central portion. 5. Fuel assembly according to claim 1, wherein said lifting plate comprises a frame-like, substantially square portion, which fits with no mentionable play within and is surrounded by said upper portion of said fuel channel, said frame-like portion comprising a water channel. 6. A fuel assembly for a boiling water reactor, comprising: 7. A fuel assembly according to claim 6, wherein said top grid is formed with a plurality of hollow-cylindrical portions each corresponding to a fuel rod position; and each tie rod comprises an upper end pin which extends through one of said hollow-cylindrical portions and is secured with a nut arranged above said top grid. |
claims | 1. Apparatus for compensation of three-dimensional movements of a target volume (1) on a patient support apparatus (2) during ion beam irradiation using a raster scanning apparatus (3),wherein a compensation apparatus comprises:a position location and tracking system (4) which detects the three-dimensional movements of the target volume (1), anda depth modulator (6) which re-adjusts the depth of penetration w of the ion beam (5),the raster scanning apparatus (3), which deflects the ion beam (5) transversely and which is in operative connection with a location measurement, control and read-out module (SAMO) and a module for changing the beam excursion (SAMS), whereinthe position location and tracking system (4) is in operative connection with a movement measurement, control and read-out module SAMB;and the depth modulator (6) is in operative connection with the movement measurement, control and read-out module SAMB,and the movement measurement, control and read-out module (SAM B) comprising a microprocessor having a memory,and the memory comprising data of a model of a structure of healthy tissue that covers the target volume (1) in the upstream direction of the beam, and the microprocessor comprising computational components which break down the detected movements of the target volume (1) vectorially into longitudinal and transverse components relative to the ion beam (5) and which compare the longitudinal components with the stored model for correction of the depth of penetration w of the ion beam (5). 2. Apparatus according to claim 1, wherein the raster scanning apparatus (3) comprises two raster scanning magnets (7, 8), which deflect an ion beam (5) orthogonally in relation to a coupling-in direction into the raster scanning magnets (7, 8), in two directions preferably arranged orthogonally relative to one another, which are in turn arranged perpendicular to one another, for scanning the area of the target volume (1) slice-wise. 3. Apparatus according to claim 2, wherein the raster scanning magnets (7, 8) are controlled by fast-reacting power supply units. 4. Apparatus according to claim 1, wherein the apparatus comprises ion acceleration elements by means of which the energy of the ion beam (5) can be adjusted so that the target volume (1) can be irradiated slice-wise, staggered in terms of depth of penetration w. 5. Apparatus according to claim 1, wherein the depth modulator (6) comprises two ion-braking plates (9, 10) of wedge-shaped cross-section which cover the entire irradiation zone of the scanned ion beam (5). 6. Apparatus according to claim 5, wherein the ion-braking plates (9, 10) are mounted on linear motors (11, 12). 7. Apparatus according to claim 5, wherein the ion-braking plates (9, 10) are arranged on electromagnetically actuatable carriages. 8. Apparatus according to claim 5, wherein the ion-braking plates (9, 10) are displaceable relative to one another with their wedge-shaped crosssections overlapping in the region of the ion beam (5). 9. Apparatus according to claim 1, wherein the position location and tracking system (4) has at least one precision video camera (13) and/or X-ray detection means and/or ultrasound detection means, which are in operative connection with an image evaluation unit in the movement measurement, control and read-out module SAMB. 10. Apparatus according to claim 1, wherein an ionisation chamber (14, 15) having a fast read-out for monitoring the intensity of the ion beam stream is arranged as a transmission counter in the beam path of the ion beam (5). 11. Apparatus according to claim 1, wherein the ionisation chamber (14, 15) is arranged between the raster scanning apparatus (3) and the depth modulator (6). 12. Apparatus according to claim 11, wherein a multiwire proportional chamber (16, 17) is arranged as a location-sensitive detector in the beam direction upstream of the depth modulator (6). 13. Apparatus for modifying the depth of penetration of an ion beam in dependence upon movement of a patient on a patient support apparatus of a therapy facility, comprisinga position location and tracking system (4) for monitoring movements of the patient,a depth modulator for adjusting the depth of penetration of the ion beam into the patient, anda movement measurement and control unit which is connected to the position location and tracking system (4) and to the depth modulator and which receives information relating to the movement of the patient from the position location and tracking system (4) and controls the depth modulator for modifying the depth of penetration,wherein the movement measurement and control unit comprises a microprocessor having a memory, and the memory comprising data of a model of a structure of healthy tissue that covers the target volume in the upstream direction of the beam, and the microprocessor, with the aid of the model and the information relating to the movement of the patient, so controlling the depth modulator that the depth of penetration of the ion beam is adjusted to a target volume element in the patient irrespective of the movement of the patient, especially irrespective of the movement of the healthy tissue relative to the target volume. 14. Apparatus according to claim 13, wherein the energy absorption of the tissue that the beam passes through and, as a result, the change in the range of the ion beam in dependence upon the tissue that the beam passes through can be calculated from the model. 15. Apparatus according to claim 13, wherein the tissue through which the beam is to pass can be determined from the information relating to the movement of the patient and the model. 16. Apparatus according to claim 13, wherein the model correlates changes in the electron density distribution in the healthy tissue (for example, obtained by means of multidimensional projection radiographs or from time-resolved CT data sets) with movement states of the body. 17. Apparatus according to claim 13, wherein the depth modulator for modifying the depth of penetration includes an apparatus for modifying the kinetic energy of the ions. 18. Apparatus according to claim 13, wherein the apparatus additionally comprises means of obtaining location information relating to the location of the ion beam relative to the patient, the movement measurement and control unit so controlling a raster scanning apparatus on the basis of the location information together with the aid of the model and the information relating to the movement of the patient that the ion beam follows a movement of the target volume in a transverse direction to the ion beam. 19. Apparatus according to claim 13, wherein the raster scanning apparatus (3) comprises two raster scanning magnets (7, 8), which deflect an ion beam (5) orthogonally in relation to a coupling-in direction into the raster scanning magnets (7, 8), in two directions preferably arranged orthogonally relative to one another, which are in turn arranged perpendicular to one another, for scanning the area of the target volume (1) slice-wise. 20. Apparatus according to claim 19, wherein the raster scanning magnets (7, 8) are controlled by fast-reacting power supply units. 21. Apparatus according to claim 13, wherein the apparatus comprises ion acceleration elements by means of which the energy of the ion beam (5) can be adjusted so that the target volume (1) can be irradiated slice-wise, staggered in terms of depth of penetration w. 22. Apparatus according to claim 13, wherein the depth modulator (6) comprises two ion-braking plates (9, 10) of wedge-shaped cross-section which cover the entire irradiation zone of the scanned ion beam (5). 23. Apparatus according to claim 22, wherein the ion-braking plates (9, 10) are mounted on linear motors (11, 12). 24. Apparatus according to claim 22, characterised in that the ion-braking plates (9, 10) are arranged on electromagnetically actuatable carriages. 25. Apparatus according to claim 22, wherein the ion-braking plates (9, 10) are displaceable relative to one another with their wedge-shaped cross-sections overlapping in the region of the ion beam (5). 26. Apparatus according to claim 13, wherein the position location and tracking system (4) has at least one precision video camera (13) and/or Xray detection means and/or ultrasound detection means, which are in operative connection with an image evaluation unit in the movement measurement, control and read-out module SAMB. 27. Apparatus according to claim 13, wherein a multiwire proportional chamber (16, 17) is arranged as a location-sensitive detector in the beam direction upstream of the depth modulator (6). 28. Apparatus according to claim 13, wherein the apparatus for detecting the structure of the healthy tissue covering the target volume in the upstream direction of the beam comprises X-ray and/or ultrasound detection in the preliminaries to and during ion beam irradiation. 29. Apparatus according to claim 13, wherein the raster scanning magnets (7, 8) comprise scanner magnet current power supply units for horizontal and vertical correction by means of control and read-out modules (SAMS) for the raster scanning magnets (7, 8). 30. Apparatus according to claim 13, wherein, for location measurement, a multiwire proportional chamber (16, 17) is provided by way of a location measurement, control and read-out module (SAMO), it being possible, for the purpose of transverse compensation, to compare information stored in the location measurement, control and read-out module (SAMO) of a supervisory control system relating to the desired position of an irradiation plan with the measured actual position of the beam position from the location-sensitive detector in real time taking into account the detected transverse movement component of the target volume (1). 31. Apparatus according to claim 13, wherein, for location correction in the transverse X and Y directions, the scanner magnets comprise power supply units of the raster scanning apparatus (3) comprise and longitudinal depth correction of the depth modulator (6) from beam position to beam position is provided. 32. Apparatus according to claim 13, wherein fast shut-down of the beam by the location measurement, control and read-out module (SAMO) of the location-sensitive detector in real time and/or by the movement measurement, control and read-out module (SAMB) of the depth modulator (6) is possible, if the difference between a measured value and a desired value of the transverse beam position and/or of the longitudinal depth of penetration w exceeds a threshold that can be set in the realtime software of the control and read-out modules SAMO and/or SAMB. 33. Method for compensation of three-dimensional movements of a target volume (1) on a patient couch (2) during ion beam irradiation using a raster scanning apparatus (3), the method comprising the following method steps:detecting a structure of healthy tissue covering the target volume (1) in the upstream direction of the beam;producing a digital model of the detected structure of the covering healthy tissue;storage of the model in a memory of the movement measurement, control and readout module (SAMB);positioning of the target volume (1) on a patient couch (2) in a treatment room (18);detecting three-dimensional movements of the target volume (1) in real time during the irradiation procedure by means of a position location and tracking system (4);vectorially dividing the movements into longitudinal and transverse components;compensating the transverse components of the movements by corrective control of raster scanning magnets (7, 8) of the raster scanning apparatus (3);compensating the longitudinal components of the movements by comparison with data of the stored model and comparison-based modification of the settings of a depth modulator (6). 34. Method according to claim 33, wherein detecting the structure of the healthy tissue covering the target volume in the upstream direction of the beam is carried out by means of X-ray and/or ultrasound detection in the preliminaries to and during ion beam irradiation. 35. Method according to claim 33, wherein the raster scanning magnets (7, 8) are controlled by way of scanner magnet current power supply units for horizontal and vertical correction by control and read-out modules (SAMS) for the raster scanning magnets (7, 8). 36. Method according to claim 33, wherein location measurement is registered and evaluated using a multiwire proportional chamber (16, 17) by way of a location measurement, control and read-out module (SAMO), information stored in the location measurement, control and read-out module (SAMO) of a supervisory control system relating to the desired position of an irradiation plan being compared, for the purpose of transverse compensation, with the measured actual position of the beam position from the location-sensitive detector in real time taking into account the detected transverse movement component of the target volume (1). 37. Method according to claim 33, wherein, by means of the scanner magnet power supply units of the raster scanning apparatus (3), location correction transversely in the X and Y direction and, by means of the depth modulator (6), longitudinal depth correction are carried out from beam position to beam position. 38. Method according to claim 33, wherein fast shut-down of the beam is initiated by the location measurement, control and read-out module (SAMO) of the location-sensitive detector in real time and/or by the movement measurement, control and read-out module (SAMB) of the depth modulator (6), if the difference between a measured value and a desired value of the transverse beam position and/or of the longitudinal depth of penetration w exceeds a threshold that can be set in the realtime software of the control and read-out modules SAMO and/or SAMB. |
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059129336 | claims | 1. A method for direct evaluation of an operating limit minimum critical power ratio (OLMCPR) of a boiling water reactor (BWR) using computer simulation(s) of postulated operational events, said reactor described by one or more parametric quantities indicative of design constraints and operating conditions, comprising the steps, executed by a computer, of: a) initializing said quantities to nominal values; b) determining steady state initial conditions of reactor operation; c) simulating an operational event for a plurality of fuel rods in a BWR core; d) calculating a minimum critical power ratio (MCPR) for each fuel rod simulated in step (c); e) determining a probability distribution P(MCPR) of potential MCPR values for each MCPR value obtained in step (d); f) calculating a value for a total number of fuel rods subject to boiling transition (NRSBT) by summing portions of probability distributions corresponding to values for MCPR where MCPR<1.0 for each probability distribution from step (e); g) perturbing one or more of said parametric quantities and recalculating another value for NRSBT; h) repeating steps (b) through (g) for a predetermined number of perturbations; i) developing a histogram of NRSBT values calculated in steps (f) through (g); j) calculating a nominal NRSBT value, based on a central tendency of an NRSBT distribution, from the histogram of NRSBT values compiled in step (i); k) calculating a confidence interval for the nominal NRSBT value; l) selecting an OLMCPR for the reactor as the initial minimal MCPR such that, for a simulation of the most limiting operational event the nominal NRSBT value at a prescribed level of confidence remains less than a predetermined cutoff value; and m) effecting said BWR operation by applying the OLMCPR selected in step (l) as an operational control parameter. (i) simulate an anticipated operational occurrence for the reactor; (ii) determine a minimum critical power ratio (MCPR) for each individual fuel rod simulated; (iii) calculate a probability distribution P(MCPR) of potential MCPR values for each MCPR value; (iv) calculate a value for a number of fuel rods subject to boiling transition (NRSBT) by summing portions of probability distributions corresponding to values for MCPR where MCPR<1.0 for all rods simulated, then repetitively calculate further values for NRSBT after perturbing one or more of said parametric quantities for a predetermined number of different perturbations; (v) develop a histogram of calculated NRSBT values over all perturbations; and (vi) select the OLMCPR from a minimal MCPR such that, for a simulation of the most limiting operational event, the nominal NRSBT value at a prescribed level of confidence as determined from analysis of the histogram determined from step (v) remains less than a predetermined cutoff value. first instruction sequence means for performing a multi-dimensional simulation of reactor thermal hydraulics and power during an operational event in the reactor; and second instruction sequence means, coupled to said first instruction means, for directly calculating an OLMCPR value based on a statistic evaluation of a minimum critical power ratio (MCPR) for each fuel rod, as obtained from said first instruction sequence means, and for developing a histogram of possible NRSBT (number of fuel rods subject to boiling transition) values calculated from perturbations of predetermined reactor plant state values and fuel rod modeling parameters. a) setting a minimum critical power ratio (MCPR) for each of a plurality of individual nuclear fuel rods used in the reactor; b) calculating a value for a total number of fuel rods subject to boiling transition (NRSBT); c) perturbing one or more reactor plant state value and/or fuel rod modeling parameter and recalculating a value for NRSBT; d) performing step (c) for a predetermined number of perturbations; e) developing a histogram of NRSBT values determined in steps (b), (c) and (d); and f) selecting the OLMCPR from a minimal MCPR such that, for a simulation of the most limiting operational event, the nominal NRSBT value at a prescribed level of confidence as determined from analysis of the histogram determined from step (v) remains less than a predetermined cutoff value. determining a probability distribution P(MCPR) to each MCPR value set in step (a), said P(MCPR) indicative of a range of possible MCPR values resulting from various operational and design uncertainties; and integrating the probability distribution for values of MCPR<1.0 for each fuel rod and summing the integration results for all fuel rods. a) developing in a memory of a computer system a histogram of NRSBT (number of fuel rods subject to boiling transition) values, corresponding to a plurality of computer simulations of a transient operational occurrence in a reactor, said simulations providing values for a critical power ratio (CPR) for one or more fuel rods for a multiple of different parametric quantities for said reactor; b) selecting a nominal NRSBT value, based on a central tendency of NRSBT distribution, statistically determined by said computer system from the histogram of NRSBT values developed in step (a); c) selecting a confidence interval for the nominal NRSBT value; d) selecting an OLMCPR value from a minimal CPR such that during a simulation of a limiting transient operational occurrence the nominal NRSBT value remains less than a predetermined cutoff value; and e) effecting said Boiling Water Reactor operation by applying the OLMCPR selected in step (d) as an operational control parameter. a) developing in a memory of said computer a histogram of NRSBT (number of fuel rods subject to boiling transition) values, corresponding to a plurality of computer simulations of a transient operational occurrence in a reactor, said simulations providing values for a critical power ratio (CPR) for one or more fuel rods for a multiple of different parametric quantities for said reactor; b) calculating a nominal NRSBT value, based on a central tendency of an NRSBT distribution, from the histogram of NRSBT values obtained in step (a); c) selecting a confidence interval for the nominal NRSBT value; d) selecting an OLMCPR for the reactor as the initial minimal MCPR such that during a simulation of a transient the nominal NRSBT value remains less than a predetermined cutoff value; and e) effecting Boiling Water Reactor operation by applying the OLMCPR selected in step (d) as an operational control parameter. a) programming a computer to determine an OLMCPR value for a boiling water reactor, said computer programmed at least to: b) providing said OLMPCR value to an output device for display, recordation or storage. a) using a computer to simulate transient operational occurrences which might occur during the operation of a Boiling Water Reactor; developing a histogram of NRSBT values from a computer simulation of a transient operational occurrence in a Boiling Water Reactor, the simulation providing values for a critical power ratio (CPR) for one or more fuel rods; c) determining a nominal NRSBT value, based on a central tendency of an NRSBT distribution, from the histogram of NRSBT values obtained in step (a); d) selecting a confidence interval for the nominal NRSBT value; and e) selecting an OLMCPR for the reactor as the initial minimal MCPR such that during a simulation of a transient the nominal NRSBT value remains less than a predetermined cutoff value. 2. The method of claim 1 wherein the operational event is a transient event associated with an Anticipated Operational Occurrence (AOO). 3. The method of claim 1, wherein multi-dimensional modeling of reactor thermal hydraulics and power is used to simulate an anticipated operational occurrence. 4. The method of claim 1, wherein the parametric quantities correspond to reactor plant state values and/or modeling parameters. 5. The method of claim 1, wherein step (c) the anticipated occurrence is simulated for a plurality of fuel rods simultaneously. 6. The method of claim 1, wherein said perturbing of one or more of said parametric quantities is accomplished using a Monte-Carlo statistical analysis approach. 7. A system for determining an operating limit minimum critical power ratio (OLMCPR) of a boiling water reactor (BWR), said system comprising a computer including a storage memory and I/O devices, said memory having stored therein rules for simultaneously simulating and evaluating thermal operating characteristics for a plurality of fuel rods during an anticipated operational occurrence of the reactor and a data base of one or more parametric quantities representing reactor plant operational state values and/or fuel rod modeling parameters, said computer programmed to: 8. The system of claim 7 wherein the operational event simulated is a transient event associated with an Anticipated Operational Occurrences (AOO). 9. The system of claim 7 wherein a multi-dimensional modeling of fuel rod thermal hydraulics and reactor power is used to simulate an anticipated operational occurrence. 10. The system of claim 7 wherein step said anticipated occurrence is simulated for a plurality of fuel rods simultaneously. 11. A computer program product embodied on a computer readable medium for determining an operating limit minimum critical power ratio (OLMCPR) for a Boiling Water Reactor (BWR), comprising: 12. The computer program of claim 11 wherein said means for performing a multi-dimensional simulation of reactor thermal hydraulics and power simulates a transient operational occurrence for a plurality of fuel rods simultaneously. 13. A method for evaluating the operating limit minimum critical power ratio (OLMCPR) of a boiling water reactor (BWR), comprising the steps, executed by a computer, of: 14. The method of claim 13 wherein the calculating of an NRSBT value in step (b) is accomplished by the steps of: 15. The method of claim 13 wherein the perturbing in step (c) of one or more reactor plant state value(s) and/or modeling parameter(s) is accomplished using randomly generated variations of said values and parameters. 16. For a nuclear fuel core of a Boiling Water Reactor, wherein fuel design and/or core configuration are contingent upon an operating margin for the reactor, said operating margin being determined by a process for evaluating an operating limit minimum critical power ratio (OLMCPR), an improved process for evaluating an OLMCPR and operating a Boiling Water Reactor which results in an improved operating margin for the reactor, comprising the steps of: 17. The method claim 16 wherein said histogram is developed on a data processing system using multi-dimensional modeling of transient operational occurrences, said data processing system including memory to store CPR data obtained from said simulations. 18. For a nuclear fuel core of a Boiling Water Reactor, wherein fuel rod design and/or core configuration are contingent upon an operating margin for the reactor, said operating margin being determined by a process for evaluating an operating limit minimum critical power ratio (OLMCPR), a improved process for evaluating an OLMCPR and operating a Boiling Water Reactor that results in an increased operating margin for the reactor, comprising the steps, executed by a computer, of: 19. A method for statistically demonstrating an operating limit minimum critical power ratio (OLMCPR) of a boiling water reactor (BWR) for compliance with licensing requirements, said reactor characterized by one or more parametric quantities indicative of design constraints and operating conditions, comprising the steps of: 20. For use in controlling a Boiling Water Reactor having a nuclear fueled core characterized by an operating limit minimum critical power ratio (OLMCPR) value, a method for determining said OLMCPR value comprising the steps of: |
041558073 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to nuclear reactor cores and more particularly to core fuel assemblies which include fuel rods laterally positioned by spacing structures or grids along the assembly length. 2. Description of the Prior Art Most liquid cooled nuclear reactors include fuel assemblies comprising nuclear fuel contained in a multiplicity of elongated cladding tubes to form fuel rods. The cladding is typically of stainless steel or zirconium alloy material, or other materials which have a relatively low neutron absorption cross section. The rods may be as long as 10 to 16 feet or more in length. To maintain proper lateral alignment among the plurality of parallel and coextending fuel rods in each fuel assembly, spacing devices are used. The spacing devices utilized typically include wire wrapping of the rods or grid structures positioned at preselected positions along the assembly length. Although many varying grid types exist, they generally comprise an "egg-crate" type structure, through which the fuel rods are inserted. The number of grids utilized may vary with the design of individual fuel assemblies. The spacing structures, or grids, also function to alleviate rod-to-rod contact among rods of the same assembly and also among rods of adjacent assemblies. Typical grid structures are described and shown in U.S. Pat. Nos. 3,379,617 and 3,379,619, both issued in the name of H. N. Andrews et al. There is a strong economic and efficiency incentive to minimize the lateral surface area and the grid material in any one assembly. The grids, even though typically comprised of a low neutron absorbing material, such as alloys of zirconium, do absorb neutrons and detract from reactor efficiency. Minimizing this neutron poisoning effect lowers the cost of electric power. With increasing reactor fuel technology, and increased fuel operating experience, fuel assembly designs may change to lessen or increase the number of grids incorporated in a fuel assembly in a given core. For example, fuel in operating reactors today that contains seven grids along the length of the fuel assembly may, in the near future, be designed with eight or nine grids, or less than seven grids. As the grids are spaced along the assembly length at optimum structural or reactivity locations, the elevation of grids in a seven grid assembly will necessarily be different than the elevation of at least some of the grids in an eight grid assembly. It is critical to ensure that any contact among assemblies is of a grid-to-grid variety, as opposed to rod-to-rod or rod-to-grid contact. If grid-to-grid contact is not maintained, there is a likelihood or fretting damage or coolant flow starvation at the contact point of a rod with another rod, or another grid. This flow starvation could create a local hot spot leading to local rod melting. This further could lead, in the extreme case, to formation of a hole in the rod cladding thereby allowing reactor coolant to contact the fuel and also allowing the nuclear fuel and fission products to enter the reactor coolant. Any coolant exposed to the reactor fuel creates concerns regarding radiation levels within the plant and potential environmental releases. Also, unmatched elevations of grids in adajcent assemblies can cause local coolant flow starvation and increased assembly vibration, as a result of induced coolant cross-flow. Coolant passing through an assembly tends to discharge radially upon approaching a flow restriction such as a grid. If there is no grid, or other spacing structure, at the corresponding elevation of an adjacent assembly, this may result in local flow starvation at a point just above the grid in the assembly from which the coolant is discharging radially, and, due to the high coolant velocity, also vibrate the adjacent assembly. Further, one of the most critical factors in the design of a reactor core is the spacing among the fuel rods and the fuel assemblies. Improper spacing, through improper grid-to-grid contact, could also lessen the efficiency of reactor operation or create local high power areas in the core. In addition to potential rod failure due to local hot spots, the high axial velocity of reactor coolant within the reactor core tends to vibrate the fuel rods. If proper lateral alignment is not maintained by the grid structures, the rods may sporadically contact the rods or grids of an adjacent assembly, which could lead to fretting over the period of reactor operation. The eventual result may similarly be the entrance of the fuel or fission products into the reactor coolant. The aligned grids may also perform similar functions under assumed accident conditions, such as seismic loading. For these reasons, it is imperative that grid-to-grid contact be maintained among adajcent assemblies, while minimizing the amount of neutron poisonous or parasitic material in the core. To assure this, apparatus is required during the operating cycles when the changeover from for example, seven grid assemblies to eight grid assemblies is being performed. The apparatus must be compatible with both eight grid and seven grid type assemblies, while minimizing the amount of parasitic material placed in the core. Further, it should be adaptable to existing refueling techniques. It should also minimize any effect on the manufacturing process so as not to unduly increase the cost of nuclear fuel and hence power generation. SUMMARY OF THE INVENTION This invention provides a transition nuclear fuel assembly which allows reactor operation with fuel assemblies in a core which fuel assemblies include spacing structures, such as grids, at different core elevations. It does so in a fashion that does not unduly add neutron parasitic material to the core, is consistent with current refueling requirements, maintains grid-to-grid contact between adjacent assemblies, and does not pose undue restrictions economically or technically to the fuel manufacturing process. The transition fuel assembly may be identical to the old and newer assemblies in a reactor core, differing only in the grid elevation and number. It may also include grids that are extended over a greater axial length along the assembly. The basic principle of the transition assembly is that it contains a spacing structure or extension thereof, or a partial spacing or grid-like structure affixed to the assembly structure, located at each elevation where there is a spacing structure in an adjacent assembly in the reactor core. These spacing or grid structures, therefore, may be identical to the other grids of the fuel assemblies, or may be similar with extended outer straps, or can be auxiliary grids of less material than a standard grid, affixed to the fuel assembly structure through thimble tubes and sleeves, extended outer straps, or other such means which maintain proper rod position and grid-to-grid contact. The type and number of grid or spacing structures in a transition assembly will vary dependent upon grid core elevation and number in the other fuel assemblies in a given core. |
044951361 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIG. 1, there is shown a reactor vessel 10 housing a nuclear reactor core 12. The core 12 includes a plurality of parallel and coextending bundled-rod fuel elements 14 supported vertically by a structure within the vessel 10. The vessel 10 is sealed at the top by a head 16 from which there is supported control element drive mechanism 18 which selectively position control elements 20 above and within some of the fuel elements 14. During operation a reactor coolant fluid, such as water, is typically pumped into the vessel through a plurality of inlet nozzles 22, passes downward through an annular region 24 between the vessel and a core barrel 23 and thermal shield 25 turns in the vessel lower plenum 26, passes upwardly through the core 12, and exits through a plurality of outlet nozzles 28. The heat energy which the core imparts to the coolant is transferred in heat transfer apparatus (not shown) typically for the ultimate purpose of electrical power generation. A typical fuel element 14 of the bundled-rod type is shown in greater detail in FIG. 2. It includes a plurality of parallel and coextending fuel rods 30, each of which includes nuclear fuel pellets 32 stacked within a sealed metallic cladding 34. The fuel rods 30 are primarily supported by upper 36 and lower 38 nozzles and by grid structures 40 spaced along the element length. The element is shown receiving a control element 20 of the "spider" type, including a plurality of cylindrical control rods 21, although plates, bars, singular rods, and so forth, can be used with varying element configurations. The control element 20 is comprised of a material having a high neutron absorption cross section, such as boron carbide, tantalum, a combination of silver-indium and cadmium, or many others well known in the art. It is to be understood that while an open-lattice or grid-type fuel element is shown, the teachings herein are applicable to other fuel element structures; including those referred to as ducted elements used in many reactor types, such as liquid metal cooled fast breeder reactors. FIG. 3 shows that the fuel elements 14 are disposed in core locations in a regularly patterned array. The letters A through O and numerals 1 through 15 are herein utilized to reference a given core position (A-1, B-2, etc.). Core 12 is surrounded by a core baffle plate 15 which serves to channel coolant flow. FIG. 4 is a schematic of a prior art fuel assembly having a 17 by 17 array of fuel rods 30. This invention is to provide a maximum power capability blanket (MPC) assembly for core perimeter locations. Inspection of FIG. 3 shows there to be assemblies which present one face to core baffle plate 15 (see for example locations A-8, B-6, F-2 in FIG. 3) and assemblies which present two faces to core baffle plate 15 (see for example locations A-7, B-5, E-2 in FIG. 3). The MPC blanket design is to substantially reduce neutron leakage with a minimum power peaking penalty. This accomplishment is provided through improved neutron reflection into adjacent enriched fuel within a blanket assembly of current PWR size which has both fertile zones and enriched zones. The MPC blanket assembly is placed at the periphery and remains at the periphery throughout life. To improve the utilization of the enriched fuel in this peripheral area, the enriched zone has a high H/U lattice, where this H/U symbol is defined as the ratio between the hydrogen to the uranium characteristic of the rod array. The improved reflection is provided by low H/U fertile blanket zones. The reduction of water (H) in the fertile zone maximizes the flux of reflected neutrons in the adjacent enriched fuel. The fertile material also performs the role of a traditional blanket, i.e., capturing escaping neutrons with fertile captures and subsequently producing power within the fertile zone. One likely reason that multiple zone enrichment concepts have not been employed as radial blankets in the past is that one inevitably traps enriched fuel in an area of low utilization. The MPC blanket improves fuel utilization through appropriate H/U zoning. This same utilization improvement also provides a general flattening of power throughout the core which tends to offset the usual radial power peaking associated with radial blankets. Specific first and second embodiments appropriate to use with non-blanket fuel of the type shown in FIG. 4 are shown in FIGS. 5 and 6. FIG. 5 shows the embodiment for location in "one-face-to-baffle" locations while FIG. 6 is the embodiment for "two-faces-to-baffle" locations. In these designs there are only two enrichments and two fuel rod sizes. The smaller fuel rod 34 is the same fuel rod 30 used in the rest of the core, and the presence of these smaller fuel rods 30 defines the area of the "enriched zone", which has a high H/U ratio. A low H/U in the fertile zone is obtained through oversized fertile fuel rods 31, the presence of which define the fertile zone. The H/U ratio in the enriched zone is accomplished by removing fuel rods 33 near the boundary of the fertile zone. The importance of the neutron reflection from the fertile zone can be deduced by comparing the results of cores using this configuration with and without an H/U adjustment in the fertile zone. The oversized rods by themselves represent a drop in cell reactivity of about 10% .DELTA..rho. compared to the smaller fuel rod cell. The impact of such a change by itself would be to lower core reactivity about 0.5% .DELTA..rho. and shift power substantially toward the middle of the core. The reflective benefits of the low H/U zone not only overcome this significant disadvantage but raise core reactivity about 0.1% .DELTA..rho. and draw power away from the core center to the adjacent enriched fuel zone. It should also be noted that the configurations in FIGS. 5 and 6 have incorporated variable enrichment and H/U zoning without a major structural redesign of the reference design shown in FIG. 4. There has been no change to the non-blanket assemblies and no change to the structure or arrangement of core internals. A side benefit of the configuration shown here is a reduction in fluence to the core vessel of about 50% over current fuel designs, since neutron leakage is reduced by approximately this amount. In FIGS. 4, 5 and 6 circles 37 represent control rod guide thimble locations. |
045335133 | abstract | A concrete pressure vessel (1) for a nuclear reactor is formed with an arch (6) arranged above the cover (4) of the pressure vessel, said arch having a horizontally directed opening permitting horizontal transport of the cover. The cover (4) is retained by a plurality of compressive force transmitting elements (11) arranged between the cover and the arch (6). |
claims | 1. In a nuclear reactor, cooled by pressurized water, an apparatus for replacing a section of a primary circuit primary pipe that interconnects first and second components of a primary circuit, the apparatus comprising:means for cutting out a section of the primary pipe;means for removing the cut primary pipe;means for supporting the end parts of a new replacement section in a position held by the removed pipe;means for bevel welding the end parts of the new replacement section of the primary pipe to confronting corresponding ends of remaining primary pipe;means for working within the pipe along the entire length of the interior surface of the primary pipe; andmeans for introducing the working means through one of first and second components of the primary circuit to a location within the welded replacement pipe section. 2. An apparatus according to claim 1, wherein the working means comprises:an anthropomorphic robot arm, a support, means for securing the robot arm and a carriage to the support for moving the support and the robot arm within the primary pipe and two sets of wheels and drive motors providing a rotational drive to at least one wheel in each set of wheels of the carriage for moving the carriage within the primary pipe. 3. An apparatus according to claim 2, wherein the support comprises a structure for supporting the robot arm, two supporting shoes controlled by jacks and one locking shoe controlled by a further jack which bears against opposite parts of the inner surface of the pipe. 4. An apparatus according to claim 1, wherein the means for introducing the working means comprise a transfer surface for lateral movement in a horizontal plane on which there is movably mounted, in a direction in the horizontal plane, a supporting table and a lift for moving a supporting plate for the working means borne on the supporting table in a vertical direction, the transfer surface being secured above a horizontal surface of the first or second components of the primary circuit through which the working means is introduced into the primary pipe through a vertical axis opening providing access to an internal part of a component of the primary circuit communicating with the interior surface of the primary pipe. 5. An apparatus according to claim 1, wherein an access gangway communicates with the interior surface of the primary pipe. 6. An apparatus according to claim 1, wherein the first or second component through which the working means is introduced is chosen between either a reactor vessel or a primary pump of the nuclear reactor. 7. An apparatus according to claim 1, wherein the working means comprise means to carry out at least one operation of machining, inspecting or welding an inner surface of the bevel welded end parts of the new replacement section and the ends of the remaining parts of the primary pipe. 8. An apparatus according to claim 1, wherein the means for bevel welding weld the end parts of the new replacement section to the ends of the remaining parts of the primary pipe while the supporting means support said end parts in position. 9. An apparatus according to claim 2, wherein the arm incorporates an end part bearing an attachment device for automatic tools. 10. An apparatus according to claim 2, wherein the arm has six axes of motor-driven rotational movement. |
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060841494 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention Melting incineration apparatus for detoxifying hazardous substances by using high frequency waves as a heat source and removing hazardous gas and hazardous inorganic compounds that are generated during melting by utilizing chain reactions. 2. Description of Prior Art The concept of using the autoclave theory and high-temperature and high-pressure melting incinerator for removal of hazardous substances from waste material has been disclosed in the prior art. Conventionally, melting furnaces have been used mainly for manufacture of new alloys, extraction of specific metal oxides from ore, and sterilization, etc. However, nothing has been made available for incineration of rubbish and waste based on the autoclave theory. Occasionally, substances whose water content is very low, (for instance, once incinerated substances or the like), can be incinerated in such a melting furnace again, or undergo secondary incineration, the technology being used only in a simple way for recycling residue (e.g. making ashes into bricks, etc.) SUMMARY OF THE PRESENT INVENTION The present invention utilizes a melting furnace that incorporates the autoclave theory and is equipped with a high-frequency (horizontal wave) device for incinerating material and for thermal decomposition and recrystallization of organic and inorganic hazardous compounds contained in it and to detoxify hazardous substances that have been recrystallized or thermally decomposed by using specific solvents. A melting furnace to accomplish the above is provided and hazardous gas that is generated during the incineration process at high temperature and under high pressure in the furnace is treated. In other words, the present invention is directed to providing a furnace that is equipped for the function of removing hazardous substances such as crystallization of halogen elements including chlorine. Safety in handling and treating organic hazardous metals by the fixation method, including organic phosphorus, arsenic, antimony, and other elements (the 15th group in the periodical table) contained in the ashes generated from incineration by the melting furnace is also a concern of the present invention. In order to make use of the high-frequency device with utmost efficiency and economy, water content in the waste is a potential problem. In the present invention, prior to the melting by high frequency waves, a facility for dehydration such as thermal dehydration by far infrared (wavelength 4 .mu.m or longer) and generic drying methods is utilized. The reason far infrared rays are chosen source is to suppress generation of secondary pollutants (CO, CO.sub.2, etc.) typically generated by other heat sources. In this manner, a series of functional sections for the dehydration of waste, the incineration by a high-frequency melting furnace, safe handling and treatment of hazardous substances in the incinerated ashes, and treatment of organic gas during the melting process are provided as part of the apparatus of the present invention. Apparatus equipped with a high-frequency melting incinerator as part of a series of treatment processes is not available in the prior art. By treating hazardous substances applying the methods stated in the invention, secondary pollution is prevented from expanding to the air and the ground, and the inventive apparatus can be thus used for the prevention of environmental pollution. |
043604951 | abstract | A target arrangement for spallation-neutron-sources, according to which tet material is continuously present at the point of incidence of a proton beam. The target material is arranged at the periphery of a rotary wheel which is internally cooled. |
description | This application is a division of and claims the benefit of priority under 35 U.S.C. §120 from U.S. Ser. No. 13/139,424 filed Jul. 7, 2011, the entire contents of which is incorporated herein by reference. U.S. Ser. No. 13/139,424 is a National Stage of PCT/JP09/070705 filed Dec. 10, 2009 which was not published under PCT Article 21(2) in English and claims the benefit of priority from Japanese Application No. 2008-317128 filed Dec. 12, 2008. The present invention relates to a reactor internal structure constituting a boiling water reactor and a method of manufacturing the same, and particularly, to a reactor internal structure and a method of manufacturing the same which can suppress deposition of crud on the reactor internal structure. In a boiling water reactor, a jet pump system is adopted to increase power density. The jet pump system forcibly circulates reactor coolant as cooling water and includes an external recirculating system and an internal recirculating system as systems for forcibly circulating reactor coolant through a core portion of a reactor pressure vessel. The external recirculating system includes a plurality of jet pumps in a reactor pressure vessel and a recirculating pump outside the reactor pressure vessel. Cooling water fed from the recirculating pump is jetted by the jet pumps and reactor water around the jet pumps is drawn and forcibly fed into a core portion from a core bottom plenum disposed under the core portion, so that the reactor coolant is forcibly recirculated in the reactor pressure vessel. FIG. 1 is a vertical cross-sectional view schematically showing a configuration of a boiling water reactor in which a jet pump system of the external recirculating system is adopted. A reactor pressure vessel 1 contains reactor coolant 2 and a core 3. The core 3 includes a plurality of fuel assemblies and control rods, not shown, and is housed in a core shroud 10. The reactor coolant 2 passes through the core 3 upward and is simultaneously heated by nuclear reaction heat of the core 3 and then becomes a two-phase flow of water and steam. The coolant 2 in the two-phase state flow into a steam separator 4 installed above the core 3 and is separated into water and steam. The steam is introduced into a steam dryer 5 above the steam separator 4 to obtain dry steam, and the dry steam is transferred into a steam turbine, not shown, through a main steam line 6 and is used for power generation. A downcomer 7 between the core shroud 10 and the reactor pressure vessel 1 contains a plurality of jet pumps 11 spaced at regular intervals in a circumferential direction. The water separated by the steam separator 4 is pressurized through a recirculation system, not shown, is introduced into the jet pumps 11 from recirculation inlet nozzles 13, and flows under the core 3 through the jet pumps 11. FIG. 2 is an enlarged perspective view showing a principle part of the jet pump 11 of FIG. 1. As shown in FIG. 2, the jet pump 11 includes a vertical riser tube 12 that introduces the coolant 2, which has been supplied from the recirculation inlet nozzle 13 of a recirculating pump, not shown, as an upward flow inside the reactor. The upper part of the riser tube 12 is connected to a pair of elbows 15 via a transition piece 14. The elbows 15 split the coolant into two downward flows. The elbows 15 are each connected to an inlet throat 17 via a mixing nozzle 16. The mixing nozzle 16 discharges the coolant 2 and surrounding reactor water is drawn with the coolant 2. The discharged coolant 2 and the drawn reactor water are mixed in the inlet throat 17. The inlet throats 17 are each connected to a diffuser 18 that feeds the coolant below the core. The elbow 15, the mixing nozzle 16, and the inlet throat 17 are integrated into a single unit called inlet mixer 51. In the case of jet pumps constituting a boiling water reactor, unfortunately, crud of iron oxide in the reactor water is deposited and builds up on surfaces of jet pump members constituting the jet pump, which increases a pressure loss and reduces a flow rate, resulting in lower circulation efficiency. The components of the reactor internal structure provides like or similar problem. For example, crud (CRUD: Chalk River Unclassified Deposit) is considerably deposited and builds up on the jet pump members constituting the inlet mixer exposed to a high flow rate of hot water. This matter has been dealt with at present by increasing the speeds of recirculating pumps (PLR pumps), which however has caused a large energy loss. Further, although a water jet cleaning method has been also proposed to remove the deposited crud, this involves extremely high cost, thus being not practical. Moreover, formation of a coating on surfaces of jet pump members has been proposed to suppress deposition of crud on reactor internal structures including the jet pump members. For example, in methods proposed in specifications of Japanese Patent Laid-Open No. 2002-207094 (Patent Document 1) and U.S. Pat. No. 6,633,623 (Patent Document 2), coatings of oxides including TiO2, ZrO2, Ta2O5, and SiO2 are formed on surfaces of the jet pump members by a CVD (chemical vapor deposition) method or process. Further, in methods proposed in specifications of Japanese Patent Laid-Open No. 2007-10668 (Patent Document 3) and U.S. Patent Application Publication No. 2007/0003001 (Patent Document 4), coatings of platinum, rhodium, iridium, palladium, silver, and gold or metal alloys thereof are formed on surfaces of component parts such as jet pump members by methods or means of, e.g., plasma spray coating, HVOF, CVD, PVD, electroplating, and electroless plating. As mentioned above, in reactor internal structures such as jet pumps of a boiling water reactor, the crud in reactor water is deposited and builds up on, e.g., surfaces of jet pump members constituting the reactor internal structures, which might increase a pressure loss and a flow rate, resulting in lower circulation efficiency. In order to improve this matter, it has been proposed, in a conventional technology, to form coatings on the surfaces of the jet pump members to thereby suppress adhesion of deposited crud such as disclosed in the related art (Patent Documents 1 to 4). In these proposals, however, deposition of crud cannot be sufficiently suppressed by forming the coatings. Moreover, the formation of the coatings requires an expensive apparatus, and size and shape of members to be coated are limited. The present invention has been conceived to solve the defective matters described above, and an object of the present invention is to provide a reactor internal structure that can sufficiently suppress deposition of crud on a reactor internal structure of a boiling water reactor. Another object of the present invention is to provide a method of inexpensively manufacturing a reactor internal structure that can sufficiently suppress deposition of crud with a simple manufacturing process and is applicable to a complexly shaped member or a large-sized member. The inventors of the present invention have earnestly studied suppression of deposition of crud on a reactor internal structure of a boiling water reactor, and as a result, the inventors found that deposition of crud can be suppressed by forming a coating of niobium oxide, zirconium titanate, or nickel titanate and also found that a high-quality coating of niobium oxide, zirconium titanate, or nickel titanate can be inexpensively formed by so-called chemical solution deposition including the steps of: applying a solution containing a compound of these metals to the surface of the reactor internal structure; and forming a coating by heat-treating the reactor internal structure coated with these solutions. Thus, the present invention has been completed. A reactor internal structure according to the present invention is a reactor internal structure constituting a boiling water reactor, the reactor internal structure having a surface at least partially coated with niobium oxide, zirconium titanate, or nickel titanate. A method of manufacturing the reactor internal structure according to the present invention includes the steps of: applying a solution containing a niobium compound to at least a part of a surface of the reactor internal structure constituting the boiling water reactor; and forming a coating of niobium oxide by heat-treating the surface of the reactor internal structure coated with the solution. A method of manufacturing a reactor internal structure according to the present invention includes the steps of: applying a titanium-zirconium compound solution to at least a part of a surface of the reactor internal structure constituting a boiling water reactor; and forming a coating of zirconium titanate by heat-treating the surface of the reactor internal structure coated with the solution. A method of manufacturing a reactor internal structure according to the present invention includes the steps of: applying a titanium-nickel compound solution to at least a part of a surface of the reactor internal structure constituting a boiling water reactor; and forming a coating of nickel titanate by heat-treating the surface of the reactor internal structure coated with the solution. According to the present invention, it is possible to suppress deposition and buildup of crud on a surface of the member of the reactor internal structure constituting the boiling water reactor, thereby keeping initial performance of coolant passing through the reactor. Moreover, according to the manufacturing method of the present invention, the reactor internal structure capable of sufficiently suppressing deposition of crud can be manufactured with a simple manufacturing process at low manufacturing cost. In the following, there will be described an example in which an embodiment of the present invention is applied to a jet pump serving as a reactor internal structure of a boiling water reactor. In the present disclosure, terms representing directions, such as “upper”, “lower”, “right”, “left” and so on, represent directions are used with reference to the illustration in the drawings or in an actual installation state of the reactor. As described above, FIG. 2 is an enlarged perspective view showing an essential portion of a jet pump 11 of the boiling water reactor. In order to suppress deposition of crud on the jet pump 11, a coating of niobium oxide, zirconium titanate, or nickel titanate is formed on at least a part of a surface of a jet pump member constituting the jet pump 11, particularly, on a portion having much deposition of crud. Thus, it is possible to suppress the deposition and the build-up of the crud in the reactor water on the surface of the jet pump member, thereby keeping initial performance of the jet pump 11 for an extended period. Although the deposition and build-up of the crud on the surface of the jet pump member can be suppressed by forming the coating, it is not clear whether such effects can be achieved by every mechanism or not, and the mechanism is assumed as follows. First, a coating of niobium oxide, zirconium titanate, or nickel titanate is formed on at least a part of the surface of the jet pump member, so that the surface of the jet pump member has a negative surface potential. Meanwhile, iron oxides such as hematite (Fe2O3) and magnetite (Fe3O4) in the crud in the reactor water also have a negative surface potential, so that it is expected that an electrical repulsive force is generated between the surface of the jet pump member and the crud in the reactor water, and the deposition and build-up of the crud can be suppressed on the surface of the jet pump member. The coating of niobium oxide, zirconium titanate, or nickel titanate is stabilized and is not melted in reactor water of an actual nuclear power plant, and moreover, oxidation resistance of a metal substrate is expected to improve in addition to the suppression of the deposition and buildup of the crud. Moreover, a coating having high adhesive strength to the metal substrate can be formed by so-called chemical solution deposition. It is preferred that the coating has a thickness of 0.01 μm to 10 μm. The thickness of the coating is set at 0.01 μm to 10 μm for the following reason: That is, in the case where the thickness of the coating is smaller than 0.01 μm, the coating cannot evenly cover the substrate and the substrate is partially exposed, so that the oxidation resistance of the substrate rapidly decreases. On the other hand, in the case where the thickness of the coating is larger than 10 μm, the adhesive strength of the coating to the substrate decreases, so that cracks may occur on the coating, the substrate becomes less resistant to oxidation, and the coating may be peeled off from the substrate. In an actual nuclear power plant, the crud to the jet pump is considerably deposited and builds up on an inner surface of an inlet mixer 51 that is exposed to a high flow rate of hot water. Accordingly, the formation of the coating is particularly effective on the inner surfaces of the jet pump members constituting the inlet mixer 51, for example, a mixing nozzle 16 and an inlet throat 17. FIG. 2 schematically shows the coating 19 formed on the inner surface of the mixing nozzle 16 and on the inner surface of inlet throat 17. Hereunder, a method of manufacturing the jet pump members according to the present invention will be described. In order to form the coating on the surfaces of the jet pump members, first, a solution containing a niobium compound, a titanium-zirconium compound solution, or a titanium-nickel compound solution is applied to the surfaces of the jet pump members. Next, the jet pump members coated with these solutions are heat-treated to form a coating of niobium oxide, zirconium titanate, or nickel titanate. In this case, the solution containing the niobium compound, the titanium-zirconium compound solution, or the titanium-nickel compound solution is, for example, a solution containing a complex of these metallic elements, a solution containing an alkoxide compound of these metallic elements, a solution containing salts of these metallic elements, and zol generated by hydrolysis on compounds of these metallic elements. Solvents of these solutions include water, alcohols such as butanol and isopropyl alcohol, other organic solvents, and mixtures of these solvents. The complex, the alkoxide compound, and the salts of these metallic elements are not particularly limited as long as the complex, the alkoxide compound, and the salts are soluble in the solvents. The compounds of metallic elements for generating the zol by hydrolysis include alkoxide compounds and salts. The compounds are not particularly limited as long as the compounds are soluble in the solvents. These solutions are applied to the surfaces of the jet pump members by, for example, dipping, spraying, spin-coating, roll-coating, bar-coating and the like method. Optimal one of the methods may be adopted according to dimensions and shapes of the jet pump members to be coated. Subsequently, the jet pump members coated with the solutions are heat-treated. The jet pump members coated with the solutions may be kept in an electric furnace and then entirely heated. Alternatively, only the surfaces of the jet pump members may be heated by infrared radiation or any other radiation. The heating method is not particularly limited to such heating methods, and other known heating methods may be used instead. The jet pump members are preferably heat-treated at 80° C. to 600° C. A heat-treatment temperature lower than 80° C. causes problems such as insufficient thermolysis of a niobium compound, a rough coating, and an unstable coating leading to aging and exfoliation. On the other hand, a heat-treatment temperature higher than 600° C. changes a structure of a metal serving as a substrate of the jet pump member, thereby deteriorating properties such as fatigue strength and creep strength. A heat-treatment atmosphere contains oxygen in air. The coating of niobium oxide, zirconium titanate, or nickel titanate is formed by the heat treatment on the surfaces of the jet pump members. The method of manufacturing the jet pump members according to the present invention is so-called chemical solution deposition which is a highly practical method inexpensively applicable to large jet pump members or complexly shaped jet pump members with a simple process without the need for an expensive apparatus. Another advantage of the manufacturing method is that a coating can be evenly formed and surface roughness of the jet pump members hardly changes in a coating operation, thereby eliminating the need for processing after the coating operation. In this example, although the present embodiment is applied to the jet pump, the present embodiment may be applied to reactor internal structures including an inner surface of a core shroud, a stand pipe of a steam separator, and a corrugated plate of a steam dryer. Further, in this case, substantially the same effects are obtainable as those attained by the described embodiment. As a test piece, there was prepared SUS304L stainless steel worked into a rectangular test piece of 40 mm×5 mm×1 mm. A 5-wt % butanol solution of niobium alkoxide was applied to a surface of the test piece by dipping and then the test piece was heat-treated at 400° C. in atmosphere for ten minutes to form a coating. This process was repeated three times to adjust a thickness of the coating. The coating formed on the surface of the test piece had a thickness of about 1 μm and contained amorphous niobium oxide. A crud deposition characteristic test that was a simulation of an actual nuclear power plant was performed to the test piece having the coating. In the crud deposition characteristic test, the test piece is immersed and contained in water at 280° C. and 7 MPa and is kept therein for 300 hours. The water contains crud of 60 ppm which is obtained by mixing hematite (Fe2O3) and magnetite (Fe3O4) in a ratio of 1 to 1. A crud deposition characteristic is evaluated by measuring a change in a weight of the test piece before and after the test. The test piece including the coating of amorphous niobium oxide formed with a thickness of about 1 μm hardly varied in weight before and after the test. A coating was formed by the same method under the same conditions as in the first example except for use of an isopropyl alcohol solution containing 5 wt % of titanium-zirconium alkoxide in a one-to-one atomic ratio of titanium to zirconium. The coating formed on a test piece contained amorphous zirconium titanate. The test piece having the coating of zirconium titanate underwent a crud deposition characteristic test by the same method as in the first example. As a result, the test piece hardly varied in weight before and after the test. A coating was formed by the same method under the same conditions as in the first example except for use of a butanol solution containing 5 wt % of titanium-nickel alkoxide in a one-to-one atomic ratio of titanium to nickel. The coating formed on a test piece contained amorphous nickel titanate. The test piece having the coating of nickel titanate underwent a crud deposition characteristic test by the same method as in the first example. As a result, the test piece hardly varied in weight before and after the test. In a first comparative example, a crud deposition characteristic test was performed to an uncoated test piece of a SUS304L substrate by the same method as in the first example. As a result, large crud deposition was observed on a surface of the test piece by a visual check or microscopy and a considerable weight gain was recognized. As described above, it was confirmed that in the case where the reactor internal structures including the jet pump members of the foregoing examples are coated with niobium oxide, zirconium titanate, or nickel titanate, deposition of crud can be effectively suppressed. Further, in the method of manufacturing the reactor internal structures including the jet pump members of the foregoing examples, a high-quality coating can be inexpensively formed by chemical solution deposition regardless of a shape and size of the reactor internal structure. According to the present invention, it is therefore possible to suppress an increase in a pressure loss of a channel of a reactor internal structure, e.g., a jet pump of a boiling reactor, and to hence stably maintain initial performance for an extended period, thereby remarkably contributing to safety of nuclear power plants. |
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description | Hereinafter, embodiments of the present invention will be described with reference to the drawings. First Embodiment A first embodiment of the present invention will be described with reference to FIGS. 1 to 6. A vertical sectional view showing the structure of a fuel assembly in this embodiment is shown in FIG. 2, and a transverse sectional view taken on line Axe2x80x94A of FIG. 2 is shown in FIG. 3. Referring to FIGS. 2 and 3, a fuel assembly 1 includes 74 fuel rods 2, filled with fuel pellets (not shown), which fuel rods are arranged in a square lattice array of 9 rowsxc3x979 columns; two water rods 3 arranged in a region in which seven of the fuel rods 2 are arrangeable; fuel spacers 4 for holding the fuel rods 2 and the water rods 3 with mutual radial intervals thereof kept immovable; an upper tie plate 5 and a lower tie plate 6 for holding the upper end portion and the lower end portion of a fuel bundle composed of the fuel rods 2 and the water rods 3, respectively; and a fuel channel box 8 for covering the outer peripheral portion of the above structure. The fuel rods 2 include normal-length fuel rods 2a, each having a normal fuel active length (filling length of fuel pellets), and short-length fuel rods 2b each having an effective length shorter than that of the normal-length fuel rods 2a. The short-length fuel rods 2b include four first short-length fuel rods 2b1 arranged in the outermost peripheral region of the square lattice array, and two second short-length fuel rods 2b2 arranged in a region adjacent to the water rods 3. The fuel spacers 4 are provided at a plurality of positions arranged in the axial direction. As shown in FIG. 3, there are 74 fuel rods 2, including the short-length fuel rods 2b, which are short in fuel active length and accordingly, the lattice positions, which are filled with the shortlength fuel rods 2b in the fuel spacer 4a positioned on the lower portion of the fuel assembly 1, become empty of fuel rods in the fuel spacer 4b positioned on the upper portion of the fuel assembly 1. For this reason, the structure of the fuel spacer 4b positioned above the upper ends of the short-length fuel rods 2b is designed to be slightly different from the structure of the fuel spacer 4a positioned below the upper ends of the short-length fuel rods 2b. Top views of the structures of these fuel spacers 4a and 4b are shown in FIGS. 4 and 1, respectively. Referring to FIGS. 4 and 1, each of the fuel spacers 4a and 4b includes a large number (74 cells for the spacer 4a, 70 cells for the spacer 4b) of cylindrical members (cells) 9 which are arranged in a square lattice array of 9 rowsxc3x979 columns corresponding to the square lattice array of the fuel rods 2, and these cylindrical members are welded to each other and are of a size to permit the fuel rods 2 to be inserted therein, respectively; a square-shaped band member (band) 11 which surrounds the outer periphery of the joined cells 9; water rod holding members 12, each being formed into a xcexa9-shape in transverse cross-section, which are welded to those cells, which are arranged in the innermost peripheral region of the square lattice array, of the cells 9, for holding the water rods 3 in the radial and axial directions; approximately quarter-round water rod holding members 13; and water rod holding springs 14, provided on the water rod holding members 13, for imparting pressing forces to hold the water rods 3 in position. Each cell 9, which is formed into an approximately cylindrical shape, includes two projections 9a for holding a respective fuel rod 2; and a spring supporting portion (not shown), provided at the joined portion with the adjacent cell 9, for suitably supporting a loop-shaped spring 10 to press against the fuel rod 2 inserted in the cell 9. It should be noted that the structures of the loop-shaped spring and the spring supporting portion, while not shown particularly in detail in the figures, are known for example from Japanese Patent Laid-open No. Hei 6-273560. The band 11, having a square-shape, whose four sides are welded to each other, includes a large number of flow tabs 15 each of which is bent in such a manner as to project between the adjacent ones of the cells 9 in the outermost peripheral region of the square lattice array in order to introduce the flow of a coolant; and eight bathtubs 16 provided two for each side of the square-shape of the band 11, each tub projecting on the fuel channel box 8 side so as to be brought in contact with the inner surface of the fuel channel box 8. The feature of this embodiment lies in the structure of the fuel spacer 4b. That is to say, the fuel spacer 4b shown in FIG. 1 is different from the fuel spacer 4a shown in FIG. 4 in that the cells 9 located at first lattice positions 7a associated with the first short-length fuel rods 2b1 are omitted and instead the supporting members 17 are provided at the lattice positions 7a. The supporting member 17 connects the two cells 9A and 9B, adjacently located on both the sides of the first lattice position 7a associated with the first short-length fuel rod 2b1 in the outer peripheral region of the square lattice array, to the band 11. FIG. 5 is a perspective view showing the structure of the supporting member 17. Referring to FIG. 5, the supporting member 17 is formed into a shape similar to one of two halves obtained by vertically dividing a cylinder having an octagonal cross-section. While not shown in FIG. 5 to avoid complication in the drawing, as shown in FIG. 1, the supporting member 17 includes, at the joined portion with the adjacent cell 9A, a spring supporting portion for suitably supporting the loop-shaped spring 10, which operates to hold the fuel rod 2 inserted in the cell 9A by imparting a pressing force against the fuel rod 2. It should be noted that the structure of the spring supporting portion, while not shown particularly in detail in the figures, is known for example from Japanese Patent Laid-open No. Hei 2-163695. In the fuel spacer 4b, the cells 9 located at the lattice positions associated with the second short-length fuel rods 2b2 are left as they are; however, the loop-shaped springs 10, which are unnecessary for the cells 9, because the fuel rods are not inserted in the cells 9, are removed from the cells 9. The fuel assembly in this embodiment, which is configured as described above, exhibits the following effects: (1) Reduction in Pressure Loss This effect will be described with reference to a comparative example in which a fuel spacer having the same structure as that of the fuel spacer 4a, in which all of the cells 9 are located without any being omitted at all of the lattice positions, as shown in FIG. 4, is positioned above the upper ends of the first short-length fuel rods 2b1 of the fuel assembly 1. In this comparative example, the cells 9, which are not required to be provided at the lattice positions associated with the first short-length fuel rods 2b1 because the fuel rods are not present at the lattice positions, are provided at the lattice positions, and therefore, the pressure loss is correspondingly increased. On the contrary, in the fuel spacer 4b in this embodiment, the cells 9 at the first lattice positions 7a associated with the first short-length fuel rods 2b1 are omitted, and instead the supporting members 17 each being formed into a semi-octagonal cross-sectional shape, are provided at the first lattice positions 7a, as shown in FIG. 1. As a result, since the flow resistance of water as a coolant flowing upward in the fuel assembly 1 is made significantly smaller than that in the comparative example, it is possible to sufficiently reduce the pressure loss. (2) Attainment of Structural Strength This effect will be described in detail with reference to the above-described comparative example. As described above, in the fuel spacer (having the same structure as that of the fuel spacer 4a shown in FIG. 4) in the comparative example, all of the cells 9 in contact with the band 11 surrounding the outer periphery of the spacer are continuously in contact with each other, to thereby maintain the structural strength of the entire spacer. For example, if an external force is applied to the fuel spacer via the fuel channel box 8 in case of an earthquake or in handling the fuel assembly, the load is first transmitted to eight of the bath-tubs 16 provided on the band 11. After that, the load is transmitted, via the band 11, to the cells 9 in the outermost peripheral region of the square lattice array joined to the inner side of the band 11, and then the force is sequentially transmitted to the cells 9 arranged on the inner peripheral side of the square lattice array (see FIG. 4). In this way, for the fuel spacer in this comparative example, since the cells 9 arranged in succession in the path along which the load is transmitted are integrally formed and an integrity of ensuring the structural effect as a whole is obtained, it is possible to sufficiently ensure the structural strength of the entire fuel spacer. On the contrary, for the fuel spacer 4b in this embodiment, as shown in FIG. 1, since one cell 9 between the cells 9A and 9B at each side of the square lattice array is omitted, the arrangement of the cells 9 in the outermost peripheral region becomes discontinuous at the position between the cells 9A and 9B. In this embodiment, however, since the two cells 9A and 9B are joined to each other by means of the supporting member 17, the cells 9A and 9B are rigidly fixed to each other via the band 11. As a result, when a load is transmitted from the band 11 as described above, it can be received by the joined structure composed of the fixed two cells 9A and 9B and the supporting member 17, and accordingly, it is possible for the fuel spacer 4b to provide a structural strength substantially comparable to that of the fuel spacer in the comparative example in which a cell 9 is located at the lattice position between the cells 9A and 9B. (3) Attainment of Degree of Freedom in Design of Short-Length Fuel Rod Arrangement As described in the paragraphs (1) and (2), the fuel spacer 4b in this embodiment is effective to reduce the pressure loss while ensuring the structural strength of the entire spacer. Such an effect can be obtained even if the lattice position associated with the first short-length fuel rod 2b1 is located at any position in the outermost peripheral region of the square lattice array. In other words, according to this embodiment, it is possible to ensure the degree of freedom in design. In the design of a fuel assembly including shortlength fuel rods, as described above, various arrangements of the short-length fuel rods may be considered in accordance with the nuclear characteristics required for the fuel assembly. Therefore, for example, there may be considered an arrangement of the first short-length fuel rod 2b1 at a position between the two opposed bath-tubs 16 in the outermost peripheral region of the square lattice array. In this case, for example, in the fuel space having the prior art structure disclosed in Japanese Patent Laid-open No. Hei 6-3473, since the cells 9 at all of the lattice positions between the two opposed bath-tubs 16 cannot be omitted, the pressure loss cannot be sufficiently reduced. On the other hand, if the reduction in pressure loss takes precedence, the short-length fuel rods cannot be arranged at all of the lattice positions between the two opposed bath-tubs 16, and therefore, the degree of freedom in design of the fuel assembly is correspondingly limited. On the contrary, in such a case, the fuel spacer 4b in this embodiment can be modified, in accordance with the arrangement of the lattice positions associated with the first short-length fuel rods 2b1, for example, into a fuel spacer 4bA shown in FIG. 6 in which the cells 9 located at the lattice positions between the two opposed bath-tubs 16 are omitted and instead the supporting members 17 are provided at those lattice positions. Accordingly, unlike the fuel spacer having the prior art structure, even if the lattice positions associated with the first short-length fuel rods 2b1 are located between the two opposed bath-tubs 16 in the outermost peripheral region of the square lattice array, it is possible to sufficiently reduce the pressure loss while ensuring the strength of the fuel spacer. As described in the paragraphs (1) to (3), according to the fuel spacer 4b in this embodiment, it is possible to sufficiently reduce the pressure loss of the fuel spacer 4b positioned upward from the upper ends of the short-length fuel rods 2b1 while usually ensuring the structural strength of the fuel spacer 4b irrespective of the arrangement of the lattice positions associated with the first short-length fuel rods 2b1. (4) Attainment of Degree of Freedom in Design of Spring Arrangement As described above, the known loop-shaped spring 10 for pressing the fuel rods 2 is essentially disposed between the adjacent cells 9, and it functions to generate pressing forces when the fuel rods 2 are inserted in the cells 9, respectively. Accordingly, if the means for imparting a spring pressing force is not provided on the supporting member 17, which is additionally provided at the lattice position associated with the first short-length fuel rod 2b1, the supporting member 17 side of the loopshaped spring 10 disposed at the joined portion between the supporting member 17 and the adjacent cell 9A comes into a free end, with a result that the loop-shaped spring 10 cannot achieve the function of pressing against the fuel rod 2 inserted in the cell 9A. Accordingly, to press the fuel rod 2 in the cell 9A, the loop-shaped spring 10 must be disposed between the cell 9A and the cell 9 which is adjacent to a portion, opposed to the supporting member 17, of the cell 9A. As a result, the spring arrangement in the entire spacer must be reviewed as a whole. This imposes a large limitation on the design. In this embodiment, however, the spring supporting portion provided on the supporting member 17 supports the loop-shaped spring 10 for pressing the fuel rods 2 and imparts a pressing force to the loop-shaped spring 10. As a result, since the loop-shaped spring 10 in the cell 9A is allowed to function just as in the fuel spacer 4a shown in FIG. 4, it is possible to increase the degree of freedom in arrangement of the loop-shaped springs 10 and hence to ensure a degree of freedom of design comparable to that in the fuel spacer 4a shown in FIG. 4. Second Embodiment A second embodiment of the present invention will be described with reference to FIGS. 7 to 11. This embodiment has a feature such that, in the fuel spacer positioned above the upper ends of the first short-length fuel rods 2b1, the spring arrangement and the shape of the supporting member are changed. FIG. 7 is a top view showing the structure of a fuel spacer 204b in this embodiment. In the fuel spacer 204b, parts common to those in the fuel spacer 4b described in the first embodiment with reference to FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. The fuel spacer 204b shown in FIG. 7 is different from the fuel spacer 4b shown in FIG. 1 in that the cells 9 at the second lattice positions 7b associated with the second shortlength fuel rods 2b2 are omitted and instead supporting members 218 are provided at the second lattice positions 7b. The supporting member 218 connects the two cells 9C and 9D, located outwardly from and adjacently to the second lattice position 7b in the square lattice array, to the water rod holding member 12. The supporting member 218 is formed into an approximately polygonal cylindrical shape with an unnecessary side portion in terms of structure cut off for making the pressure loss as small as possible. In the fuel spacer 204b, a supporting member 217, having a structure in which the spring supporting portion is removed from the supporting member 17 shown in FIG. 1, is used as a supporting member for connecting the two cells 9A and 9B adjacently located on both sides of the lattice position associated with the first short-length fuel rod 2b1, to the band 11. With this configuration, since a loop-shaped spring 10 is not disposed on the supporting member 217, the spring arrangement in the entire fuel spacer is changed such that the supporting member 218 has two spring supporting portions (not shown) for suitably supporting the two loop-shaped springs 10 to press against the fuel rods 2 inserted in the cells 9C and 9D to impart pressing forces thereto. To be more specific, two of the known spring supporting portions having the same structure as that of the spring supporting portions used for the cells 9 are simply provided at a joined portion between the cells 9C and 9D of the supporting member 218. The remaining configuration of this embodiment is substantially the same as that of the first embodiment. According to this embodiment, in addition to the same effect as that of the first embodiment, there can be obtained an effect of simplifying the structure because the supporting member 217 has no spring supporting portion. While the supporting member 217 formed into a semi-octagonal cylindrical shape is used in the second embodiment, the present invention is not limited thereto. For example, the supporting member 217 may be formed into another shape, for example, a semi-cylindrical shape (with a partial peripheral length portion cut off) having the same thickness as that of the cell 9. This exhibits the following effect. In general, the cell 9 is manufactured by cutting a circular tube, having a specific outside diameter and a specific thickness, into a specific length, and processing the cut piece to form the projections 9a and also cutouts for the spring supporting portion. Here, if a supporting member 217A formed into a semi-cylindrical shape having the same thickness as that of the cell 9 is used as the supporting member, such a supporting member 217A can be manufactured using the raw circular tube for forming the cell 9. This is effective to reduce the manufacturing cost by making the circular tube shareable between the supporting member 217A and the cell 9. From the viewpoint of reduction in pressure loss, it may be Desirable that the peripheral length of the cylindrical shape of the supporting member 217A be made as short as possible within a length range required for welding the supporting member 217A to the adjacent cells 9 with no problem. Further, a supporting member 217A formed into an approximately cylindrical shape similar to that of the cell 9 may be used as the supporting member. FIG. 8 is a top view showing the structure of a fuel spacer 204bA including such a supporting member 217A. The supporting member 217A connects the cells 9A and 9B, adjacently located on both sides of the lattice position 7a associated with the first short-length fuel rod 2b1, to the band 11, and also the supporting member 217A is connected to the cell 9E located inwardly from and adjacently to the first lattice position 7a in the square lattice array. In addition, the supporting member 217 is made as thin as possible within an allowable thickness range in terms of the structural strength of the fuel spacer for making the cross-sectional area smaller than that of the cell 9 thereby reducing the pressure loss. The supporting member 217A formed into an approximately cylindrical shape is manufactured using a circular tube having a specific thickness, which tube is different from the raw circular tube for forming the cell 9, or using the raw circular tube for forming the cell 9, and grinding the inner surface of the tube to increase the inside diameter (that is, decrease the thickness). In the latter case, there can be obtained an effect of reducing the manufacturing cost by making the circular tube shareable between the supporting member 217A and the cell 9. It should be noted that the cross-sectional shape of the supporting member 217A in this modification is not limited to a cylindrical shape, but may be of course a polygonal shape insofar as it satisfies the requirement that the cross-section of the supporting member 217A is smaller than that of the cell 9. Further, the supporting member 217A may be configured as a member having the same cross-sectional shape as that of the supporting member 218 except that the spring supporting portions for supporting the loop-shaped springs 10 are not provided. In the manufacture of the supporting member 218, the member 217 (replaced from the supporting member 217A) having the same cross-sectional shape as that of the supporting member 218 can be manufactured by punching or bending the same raw material as that for the supporting member 218. This is effective to reduce the manufacturing cost by making the raw material shareable between the member (replaced from the supporting member 217A) and the supporting member 218. In this sharing of the raw material, the shape of the supporting member 218 is not limited to a polygonal cylindrical shape but may be of course a thin cylindrical shape or a cylindrical shape with a partial peripheral portion cut off. Third Embodiment A third embodiment of the present invention will be described with reference to FIGS. 9 to 11. This embodiment has a feature such that, in the fuel spacer positioned above the upper ends of the first short-length fuel rods 2b1, the shape of the supporting member and the supporting structure are further changed. FIG. 9 is a top view showing the structure of a fuel spacer 304b in this embodiment. In the fuel spacer 304b, parts common to those of the fuel spacer 4b described in the first embodiment with reference to FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. The fuel spacer 304b shown in FIG. 9 is different from the fuel spacer 4b shown in FIG. 1 in that the supporting member 17 located at the first lattice position (see FIG. 3) associated with the first short-length fuel rod 2b1 is replaced with a supporting member 317. The supporting member 317 connects the two cells 9A and 9B adjacently located on both sides of the first lattice position 7a associated with the first short-length fuel rod 2b1 in the outermost peripheral region of the square lattice array, and the cell 9E located inwardly from and adjacently to the first lattice position 7a in the square lattice array, to each other. In summary, the supporting member 317, which is not joined to the band 11, joins the adjacent cells 9A, 9B and 9E to each other. With this configuration, the entire fuel spacer has a structure as shown in FIG. 9 in which each cell 9 in the second layer from the outermost periphery of the square lattice array is fixedly joined to four adjacent cells or three adjacent cells and one supporting member 317 at four positions spaced at intervals of 90xc2x0 in the circumferential direction thereof. Like the supporting member 17 in the first embodiment, the supporting member 317 is formed into one of two halves obtained by vertically dividing a cylinder having an octagonal cross-section. While not shown in detail, like the supporting member 17, the supporting member 317 includes, at the joined portion with the adjacent cell 9A, a spring supporting portion for suitably supporting a loop-shaped spring 10 to hold the fuel rod 2 inserted in the cell 9A to impart a pressing force thereto. The other configuration is substantially the same as that of the first embodiment. Like the fuel assembly in the first embodiment, the fuel assembly in this embodiment, configured as described above, has the following four effects: (1) Reduction in Pressure Loss In the fuel spacer 304b in this embodiment, the cells 9 at the first lattice positions 7a associated with the first short-length fuel rods 2b1 are omitted, and instead the supporting members 317 each being formed into a semi-octagonal cross-sectional shape are provided at the first lattice positions 7a. As a result, since the flow resistance of water as a coolant flowing upward in the fuel assembly 1 is significantly reduced, it is possible to sufficiently reduce the pressure loss. (2) Attainment of Structural Strength In the fuel spacer 304b in this embodiment, as shown in FIG. 9, since one cell 9 between the cells 9A and 9B at each side of the square lattice array is omitted, the arrangement of the cells 9 in the outermost peripheral region becomes discontinuous at the position between the cells 9A and 9B. In this embodiment, however, since the two cells 9A and 9B are connected to the adjacent cell 9E on the inner peripheral side by means of the supporting member 317, the two cells 9A and 9B are rigidly fixed to each other via the cell 9E. As a result, when a load is transmitted from the band 11, it can be received by the joined structure composed of the fixedly connected two cells 9A and 9B, the supporting member 317, and the cell 9E. Accordingly, it is possible for the fuel spacer 304b to provide a structural strength substantially comparable to that of the fuel spacer in which the cell 9 is located at the lattice position between the cells 9A and 9B. (3) Attainment of Degree of Freedom in Design of Short-length Fuel Rod Arrangement In the fuel spacer 304b in this embodiment, even if the lattice position associated with the first short-length fuel rod 2b1 is located at any position in the outermost peripheral region of the square lattice array, the cell 9 at the lattice position can be omitted and instead the supporting member 317 can be provided at the lattice position. Accordingly, unlike the fuel spacer disclosed for example in Japanese Patent Laid-open No. Hei 6-3473, even if the lattice positions associated with the first short-length fuel rods 2b1 are located between the two opposed bath-tubs 16 in the outer peripheral region of the square lattice array, it is possible to omit the cells 9 at the lattice positions, and hence to sufficiently reduce the pressure loss while ensuring the strength of the fuel spacer. (4) Attainment of Degree of Freedom in Design of Spring Arrangement In the fuel spacer 304b in this embodiment, the spring supporting portion provided on the supporting member 317 supports the loop-shaped spring 10 for pressing against the fuel rods 2 to impart a pressing force thereto. Accordingly, it is possible to ensure a degree of freedom in the spring arrangement comparable to that in the fuel spacer in which the cells 9 are located at the first lattice positions 7a associated with the first shortlength fuel rods 2b1. It should be noted that the third embodiment may be variously modified without departing from the basic configuration thereof. Some modifications will be described below. FIG. 10 is a top view showing the structure of a fuel spacer 304bA in which, like the second embodiment, supporting members 318 are provided at lattice positions associated with the second short-length fuel rods 2b2 and further the spring supporting portions are removed from the supporting members 317. As shown in FIG. 10, the supporting member 318 is provided to connect the two cells 9C and 9D, adjacently located on the outer peripheral side of the second lattice position 7b associated with the second short-length fuel rod 2b2 in the square lattice array, to the water rod holding member 12; and it is formed into an approximately cylindrical shape. Also, since supporting members 317A have no spring supporting portions, the spring arrangement in the entire spacer is changed such that the supporting member 318 has two spring supporting portions for suitably supporting the two loop-shaped springs 10 to press the fuel rods 2 inserted in the cells 9C and 9D for imparting pressing forces thereto (like the second embodiment, the known two spring supporting portions are simply provided). This modification is effective to simplify the structure because the supporting members 317 have no spring supporting portions. In the fuel spacer in the modification shown in FIG. 10, the supporting members 317A may be formed into a semi-cylindrical (with a partial peripheral length portion cut off) shape having the same thickness as that of the cell 9. In this case, the supporting member 317A can be manufactured using the raw circular tube for forming the cell 9. This is effective to reduce the manufacturing cost by making the circular tube shareable between the supporting member 317A and the cell 9. From the viewpoint of reduction in pressure loss, it may be desirable that the peripheral length of the cylindrical shape of the supporting member 317A be made as short as possible within a length range required for welding the supporting member 317A to the adjacent cells 9 with no problem. The shape of the supporting members 317A and 318A may be made similar to that of the supporting member 318. FIG. 11 is a top view showing the structure of a fuel spacer 304bB including supporting members 317B and 318A having the configuration described above. As shown in FIG. 11, the supporting members 317B and 318A, each being formed into a cylindrical shape with a partial peripheral length portion cut off, are identical in cross-sectional shape to each other. These supporting members 317B and 318A are substantially similar to each other except that the supporting member 317B has no spring supporting portions for supporting the loop-shaped springs 10 to press against the fuel rods 2 inserted in the cells 9C and 9D. As a result, in the manufacture of the supporting member 318A, the supporting member 317B having the same cross-sectional shape in this modification can be manufactured by punching or bending the same raw material element as that for the supporting member 318A. This is effective to reduce the manufacturing cost by making the raw material shareable between the supporting members 317B and Fourth Embodiment A fourth embodiment of the present invention will be described with reference to FIG. 12. This embodiment has a feature such that, in the fuel spacer positioned above the upper ends of the first short-length fuel rods 2b1, the flow tabs are partially omitted. FIG. 12 is a top view showing the structure of a fuel spacer 404b in this embodiment. In the fuel spacer 404b, parts common to those in the fuel spacer 4b described in the first embodiment with reference to FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. The fuel spacer 404b shown in FIG. 12 is different from the fuel spacer 4b shown in FIG. 1 in that, of the large number of the flow tabs 15 formed on the band 11 for introducing the flow of a coolant, those located at positions adjacent to the supporting member 17 for connecting the cells 9A and 9B to the band 11 at each first lattice position 7a associated with the first short-length fuel rod 2b1 are omitted. The remaining configuration is substantially the same as that in the first embodiment. This embodiment having the above configuration exhibits the following effects: The flow tabs 15 provided on the band 11 of the fuel spacer 404b have a function of directing the flow of a coolant in the fuel assembly 1 toward the fuel rod 2 side as much as possible, thereby improving the cooling effect of the fuel rods 2 and enhancing the critical power characteristic. The flow tabs 15, however, have an inconvenience in that, since the projecting shapes of the flow tabs 15 obstruct the flow of the coolant, the pressure loss is correspondingly increased. Incidentally, since the fuel spacer 404b is positioned above the upper ends of the short-length fuel rods 2b1, no fuel rods are present at the lattice positions associated with the short-length fuel rods 2b1 and instead the supporting members 17 are provided at the lattice positions. Accordingly, the provision of the flow tabs 15 in the vicinity of the lattice positions does not improve the critical power characteristic, but causes a problem in that it increases the pressure loss. For this reason, the flow tabs 15 adjacent to each supporting member 17 may be omitted. This is effective to eliminate an increase in pressure loss due to the projecting shapes of the flow tabs 15, and hence to further reduce the pressure loss. In the fourth embodiment, the lattice position associated with the first short-length fuel rod 2b1 is located at the intermediate position on each of the four sides of the outermost peripheral region of the square lattice array and the flow tabs 15 adjacent thereto are omitted; however, the present invention is not limited thereto. For example, the lattice position associated with the first short-length fuel rod 2b1 may be located at any position in the outermost peripheral region, preferably, except for four corners of the square lattice array. The reason for this will be described below. The effect of improving the critical power characteristic due to the flow tabs 15 becomes largest at the four corners in the outermost peripheral region of the square lattice array and becomes smaller at other positions in the outermost peripheral region. Accordingly, in the case where the lattice position associated with the first short-length fuel rod 2b1 is positioned in the outermost peripheral region except for the four corners and the supporting member 17 is provided at the lattice position, the effect of improving the critical power characteristic is not reduced so much even if the flow tabs 15 adjacent to the supporting member 17 are omitted; while the effect of reducing the pressure loss due to omission of the flow tabs 15 becomes large irrespective of the arrangement of the flow tabs 15. As a result, it may be desirable to locate the lattice position associated with the first short-length fuel rod 2b1 in the outermost peripheral region except for the four corners and omit the flow tabs 15 adjacent to the supporting member 17 located at the lattice position associated with the first short-length fuel rod 2b1. In the fourth embodiment, only the flow tabs 15 adjacent to the lattice position associated with the first short-length fuel rod 2b1 in the configuration of the first embodiment are omitted; however, the present invention is not limited thereto. For example, the flow tabs 15 at other positions may be omitted insofar as an effect exerted on the critical power characteristic due to omission of the flow tabs 15 is allowable. In this case, it is possible to further reduce the pressure loss. Further, the flow tabs 15 in each configuration of the second and third embodiments may be partially omitted. In each of the first to fourth embodiments, the present invention is applied to the fuel assembly in which the two water rods are arranged in the region in which the seven fuel rods 2 are arrangeable; however, the present invention is not limited thereto, but can be applied to a fuel assembly in which one or three or more water rods are arranged in a region in which six or less or eight or less of the fuel rods 2 are arrangeable. The present invention can be also applied to a fuel assembly including a square type water rod formed to have a square shape in transverse cross-section. FIG. 13 shows a fuel spacer 4bB of a fuel assembly including such a square type water rod, wherein the fuel spacer 4bB is modified from the fuel spacer 4b shown in FIG. 1. The fuel spacer 4bB shown in FIG. 13 is different from the fuel spacer 4b shown in FIG. 1 in that a square type water rod holding member 12A is provided at the central portion of the fuel spacer 4bB and the number of the cells 9 is correspondingly reduced to 68 cells. The remaining configuration is substantially the same as that shown in FIG. 1. The fuel spacer 4bB exhibits an effect similar to that obtained by the fuel spacer 4b. In each of the first to fourth embodiments, the present invention is applied to a fuel assembly including second short-length fuel rods 2b2 in addition to the first short-length fuel rods 2b1; however, the present invention is not limited thereto. For example, the present invention can be applied to a fuel assembly in which fuel rods each having the normal fuel active length are located at the second lattice positions in place of the second short-length fuel rods 2b2 or no fuel rods may be provided at the second lattice positions. In this case, the same effect can be obtained. In each of the first to four embodiments, the first short-length fuel rods 2b1 are dispersedly located in the outermost peripheral region of the square lattice array, that is, two or more of the first short-length fuel rods 2b1 are located so as to be not adjacent to each other in the outermost peripheral region, however, the present invention is not limited thereto. For example, even in the case where two or more of the short-length fuel rods are arranged in the outermost peripheral region as disclosed in Japanese Patent Laid-open No. Hei 6-2373, the cells at the lattice positions associated with the short-length fuel rods can be removed and instead members similar to the supporting members 17, 217 or 317 shown in the first to third embodiments can be provided at the lattice positions. Even in this case, there can be obtained a strength ensuring effect comparable to that obtained in each of the above-described embodiments. Fifth Embodiment A fifth embodiment of the present invention will be described with reference to FIGS. 14 to 20. This embodiment has a feature such that the necessary minimum number of one kind of the loop-shaped springs are reasonably arranged over an entire fuel spacer. In this embodiment, parts common to those in the first embodiment are designated by the same symbols and explanation thereof will be omitted. FIG. 14 is a top view showing the structure of a fuel spacer 504b in this embodiment. As shown in FIG. 14, the fuel spacer 504b in this embodiment is configured such that the cells 9 at the first lattice positions 7a associated with the first short-length fuel rods 2b1 and the second lattice positions 7b associated with the second short-length fuel rods 2b2 are removed, and correspondingly, the arrangement of the loop-shaped springs 10 over the fuel spacer 504b is entirely changed from that in the fuel spacer 4a shown in FIG. 4. That is to say, with respect to the cell 9C adjacent in the same row to one of the two second lattice positions 7b adjacent to the water rods 3 and the cell 9D adjacent in the same column to the above second lattice position 7b, each of the cells 9C and 9D includes, at a portion on the second lattice position 7b side, a known spring supporting portion. A loop-shaped spring 10A (10B) for pressing against the fuel rod 2 inserted in the cell 9C (9D) is provided in the above spring supporting portion provided in the cell 9C (9D). The loop-shaped spring 10A (10B) is, as will be described later, supported by an approximately cylindrical spring pressing member 19 provided at the second lattice position 7b in such a manner as to generate a pressing force applied to the fuel rod 2 inserted in the cell 9C (9D). The spring pressing member 19 is made as thin as possible within an allowable range in terms of the structural strength of the fuel spacer. That is to say, to reduce the pressure loss the transverse cross-section of the spring pressing member 19 is made smaller than the transverse cross-section of the cell 9. In addition, the spring pressing member 19 can be manufactured commonly using the raw circular tube for forming the cell 9. This is effective to reduce the manufacturing cost by making the raw material shareable between the spring pressing member 19 and the cell 9. The water rod holding member 12 formed into the xcexa9-shape in transverse cross-section is joined to each of the two spring pressing members 19. The detailed structure near the two second lattice positions 7b, which forms the largest difference between the fuel spacer 504b and the fuel spacer 4a, will be described below. (1) Fuel Spacer 4a FIG. 15 is a transverse sectional view showing the structure near the two second lattice positions 7b of the fuel spacer 4a; and FIG. 16 is a perspective view showing the structure near the joined portions between the water rod holding member 12 and the cells 9. In addition, for convenience in description, the fuel rods 2 and the water rods 3 are additionally shown in FIG. 15. Referring to FIGS. 15 and 16, the water rod holding member 12 includes two spring holding projecting pieces 12a each of which projects in the shape of tongue in the loop of the loop-shaped spring 10 for holding the loop-shaped spring 10, and a spring pressing projecting piece 12b which projects in the shape of tongue in such a manner as to be brought in contact with the loop of the loop-shaped spring 10 from the outer peripheral side. The two spring holding projecting pieces 12a and the spring pressing projecting piece 12b all project in the same direction (leftward in FIG. 16), and the two spring holding projecting pieces 12a are provided above and below the spring pressing projecting piece 12b, respectively. These spring holding projecting pieces 12a and the spring pressing projecting piece 12b are manufactured by press-working a base portion of the water rod holding member 12 to form tongue-shaped cut pieces corresponding to the projecting pieces 12a and 12b and two windows 12c. The spring pressing projecting piece 12b is finished by bending the corresponding tongue-shaped cut piece from its root and then flattening it. The windows 12c are provided to provide spaces in which the loop-shaped spring 10 is inserted when the loop-shaped spring 10 is mounted. The cell 9 to be joined to the water rod holding member 12 has two windows 9b (only one is shown for simplicity) and a projecting piece 9c which projects in the direction opposed to the projecting direction of the projecting pieces 12a. The two spring holding projecting pieces 12a are in contact with the projecting piece 9c. The contact portions of the spring holding projecting pieces 12a with the cell projecting piece 9c are inserted in the loop of the loop-shaped spring 10, so that the vertical movement of the looped spring 10 may be restricted. In addition, both ends of the windows 12c and 9b form removal preventive portions 12c1 and 9b1 (only partially shown) for preventing the removal of the loop-shaped spring 10. The spring pressing projecting piece 12b functions to restrict the horizontal displacement of the loop-shaped spring 10 due to expansion/contraction thereof, and hence to generate a pressing force applied to the fuel rod 2 inserted in the cell 9. At this time, as shown in FIG. 15, the distance d1 between the spring pressing projecting piece 12b and the adjacent fuel rod 2 is made equal to the distance d2 between the two adjacent fuel rods 2. With this configuration, a pressing force of the loop-shaped spring 10 generated when the spring 10 is pressed by the spring pressing projecting piece 12b is made equal to a spring pressing force of the loop-shaped spring 10 generated when the spring 10 is held between the two fuel rods 2. As a result, the fuel rod 2 can be suitably fixed in the cell 9 at the lattice position 7b. (11) Fuel Spacer 504b FIG. 17 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between the spring pressing member 19 and the cells 9 in the fuel spacer 504b, which is an essential portion of this embodiment. Referring to FIG. 17, the spring pressing member 19 includes two spring holding projecting pieces 19a and two spring holding projecting pieces 19b, which function as spring holding portions projecting in the shape of a tongue inserted in the loops of the loop-shaped springs 10A and 10B for holding the loop-shaped springs 10A and 10B, respectively; and two spring pressing projecting pieces 19C and 19D which function as spring pressing portions projecting in the shape of a tongue in such a manner as to be in contact with the loops of the loop-shaped springs 10A and 10B from the outer peripheral side, respectively. The structures and functions of these projecting pieces 19a (19b) and 19c (19d) are similar to those of the projecting pieces 12a and 12b of the water rod holding member 12 described in the paragraph (1), respectively. To be more specific, the two spring holding projecting pieces 19a are provided above and below the spring pressing projecting piece 19c, respectively; and the two spring holding projecting pieces 19b are provided above and below the spring pressing projecting piece 19d, respectively. These spring holding projecting pieces 19a and 19b and the spring pressing projecting pieces 19c and 19d are manufactured by press-working a cylindrical base portion 19g of the spring pressing member 19 to form tongue-shaped cut pieces corresponding to the projecting pieces 19a, 19b and 19c and 19d and four windows 19e and 19f. Each of the spring pressing projecting pieces 19c and 19d is finished by bending the corresponding tongue-shaped cut piece from its root and then flattening it. At this time, the two spring holding pieces 19a (19b) are in contact with a projecting piece 9Cc (9Dc) formed in the cell 9C (9D), and the contact portions of the spring holding projecting pieces 19a (19b), with the cell projecting piece 9Cc (9Dc) are inserted in the loop-shaped spring 11A (10B). In addition, to prevent the removal of the loop-shaped spring 10A (10B), the ends of windows 19e (19f) of the loop-shaped spring 10A (10B) have removal preventive portions 19e1 (19f1) and the ends of the windows of the cell 9C (9D) have removal preventive portions (not shown for simplicity). The spring pressing projecting piece 19c (19d) functions to restrict the horizontal displacement of the loop-shaped spring 10A (10B) due to expansion/contraction thereof, and hence to generate a suitable pressing force (similar to the pressing force described in the paragraph (1)) to the fuel rod 2 inserted in the cell 9C (9D). Here, there is a large structural difference between the projecting pieces 19a, 19b, 19c and 19d and the projecting pieces 12a and 12b of the water rod holding member 12 described in the paragraph (1) in that the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 17) and one spring pressing projecting piece 19d also projects in the same direction; however, the other spring pressing projecting piece 19c projects in the opposite direction (rightward in FIG. 17). The function of this embodiment having the above configuration will be described below. (1) Improvement in Reactivity Controllability In the fuel assembly 1 in this embodiment, since six of the short-length fuel rods 2b are included with the fuel rods 2 arranged in the square lattice array of 9 rowsxc3x979 columns, it is possible to equalize the H/U ratio by making use of a saturated water region formed on the upper side of the short-length fuel rods 2b. At this time, by arranging the short-length fuel rods 2b at the lattice positions in the outermost peripheral region of the square lattice array and at the lattice positions adjacent to the water rods, it is possible to more effectively improve the controllability of the reactivity by reducing the void coefficient as disclosed in Japanese Patent Laid-open No. Hei 5-232273. (2) Reduction in Pressure Loss Since the unnecessary cells 9 at the first and second lattice positions 7a and 7b in the fuel spacer 4b positioned above the upper ends of the short-length fuel rods 2b are omitted, the pressure loss can be correspondingly reduced. In addition, a spring pressing member 19 is provided in place of the cell 9 at each second lattice position 7b; however, since the transverse cross-section of the spring pressing member 19 is made smaller than that of the cell 9 as described above, it is possible to reduce the pressure loss. 1 (3) Reasonable Arrangement of Spring As a result of removing the cells 9 in the fuel spacer 504b for reducing the pressure loss (see the paragraph (2)), the arrangement of the loop-shaped springs 10 over the entire fuel spacer is necessarily changed such that the loop-shaped springs 10A and 10B are respectively provided on portions, on the second lattice position 7b side, of the cells 9C and 9D located at the lattice positions adjacent to each of the two second lattice positions 7b for pressing against the fuel rods 2 in the cells 9C and 9D. The usual loop-shaped spring 10 functions to generate pressing forces when the fuel rods 2 are inserted in a pair of the adjacent cells 9, and accordingly, if such usual loop-shaped springs are used for the loop-shaped springs 10A and 10B, the loop-shaped springs 10A and 10B are made free on the second lattice position 7b side and thereby cannot generate the pressing forces. Incidentally, the structure in which one loopshaped spring which is free on one side is supported at one lattice position in such a manner as to generate a suitable pressing force has been known, for example, as represented by the structure shown in FIG. 16 or the structure disclosed in Japanese Patent Laid-open No. Hei 6-273560; however, the above-described structure in which the two loop-shaped springs 10A and 10B each being free on one side are supported at one lattice position has not been known. On the contrary, in the fuel spacer 504b in this embodiment, since the spring pressing member 19, which includes the four spring holding projecting pieces 19a and 19b and the two spring pressing projecting pieces 19c and 19d, is provided at one of two second lattice positions 7b, it is possible to support the loop-shaped springs 10A and 10B so that each is free on one side and hence to generate suitable pressing forces. This makes it possible to reasonably arrange the necessary minimum number (36 pieces) of the loop-shaped springs 10 over the entire fuel spacer without increasing the kinds of springs being used, more specifically, using only one kind of the springs. (4) Attainment of Rigidity/Strength of Spring Pressing Member to Pressing Force of Spring This function will be described with reference to a comparative example. FIG. 18 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between a spring pressing member and the cells in the comparative example. In FIG. 18, parts common to those in FIG. 17 are designated by the same symbols. In the comparative example shown in FIG. 18, two of the structures shown in FIG. 16, each being similar to that disclosed in Japanese Patent Laid-open No. Hei 6-273560, are simply arranged for supporting the two loop-shaped springs 10A and 10B so that each is free on one side at one second lattice position 7b. The structure shown in FIG. 18 is different from that shown in FIG. 17 in that the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 18) and the two spring pressing projecting pieces 19c and 19d project in the same direction (leftward in FIG. 18). Such a structure as shown in FIG. 18 has the following inconvenience. That is to say, to bring the spring pressing projecting pieces 19c and 19d in contact with the loops of the loop-shaped springs 10A and 10B from the outer peripheral side, the spring pressing projecting pieces 19c and 19d are required to project on the inner side (toward the front in FIG. 18) of the spring pressing member 19 more than the spring holding projecting pieces 19a and 19b inserted in the loops. Accordingly, if the projecting direction of all of the spring pressing projecting pieces 19c and 19d is made identical to the projecting direction of the spring holding projecting pieces 19a and 19b, the size of the cutouts on both sides of the spring pressing projecting pieces 19c and 19d must be made larger, and correspondingly the width (or area) of a bridge 19g1, equivalent to the root portion of the spring pressing projecting piece 19c, of the base plate portion 19g of the spring pressing member 19 becomes smaller. This makes it difficult to ensure a sufficient strength and rigidity against the pressing force of the loop-shaped spring 10A. On the contrary, according to the configuration of this embodiment as shown in FIG. 17, since the projecting direction of one spring pressing projecting piece 19c is reversed relative to the projecting direction of the spring holding projecting pieces 19a and 19b, the width (or area) of the bridge 19g1 can be made larger. This makes it possible to ensure a sufficient strength and rigidity. As described above, according to this embodiment, the fuel assembly 1 is configured such that the shortlength fuel rods 2b are arranged in the outermost peripheral region of the square lattice array of 9 rowsxc3x979 columns adjacent to the water rods 3, and at the fuel spacer 4b, the cells 9 at the lattice positions 7a and 7b associated with the short-length fuel rods 2b are removed to reduce the pressure loss. In this fuel assembly 1, the two loop-shaped springs 10A and 10B, each being free on one side, are supported by the spring pressing member 19 at one second lattice position 7b, so that the necessary minimum number of the loop-shaped springs 10 may be reasonably arranged over the entire fuel spacer without increasing kinds of the springs being used. Since the number of the loop-shaped springs 10 is selected at the necessary minimum value, there can be obtained an effect of further reducing the pressure loss, and since the width (or area) of the bridge 19g1 can be made larger, there can be obtained an effect of ensuring a sufficient strength and rigidity against the pressing forces of the loop-shaped springs 10A and 10B. In the fifth embodiment, the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 17) and one spring pressing projecting piece 19d also projects in the same direction; while the other spring pressing projecting piece 19c projects in the opposite direction (rightward in FIG. 17); however, the structure of the projecting pieces is not limited thereto. Hereinafter, two modifications will be described with reference to FIGS. 19 and 20. FIG. 19 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between the spring pressing member and the cells according to the first modification. In FIG. 19, parts common to those in FIG. 17 are designated by the same symbols. The structure shown in FIG. 19 is different from that shown in FIG. 17 in that the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 19), while both spring pressing projecting pieces 19c and 19d project in the opposite direction (rightward in FIG. 19). With this structure, since the width (or area) of the bridge 19g1 can be made larger than that in the comparative example shown in FIG. 18, it is possible to ensure a sufficient strength and rigidity against the pressing forces of the loop-shaped springs 10A and 10B. That is to say, it becomes apparent that at least one of the spring pressing projecting pieces may project in the direction opposite to the projecting direction of the spring holding projecting pieces. FIG. 20 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between the spring pressing member and the cells according to the second modification. In FIG. 20, parts common to those in FIG. 17 are designated by the same symbols. The structure shown in FIG. 20 is different from that shown in FIG. 17 in that the opposed free ends of the spring pressing projecting pieces 19c and 19d in the structure shown in FIG. 17 are connected to the base plate portion 19g of the spring pressing member 19, that is, as shown in FIG. 19, both right and left ends of each of the projecting pieces 19c and 19d are connected to the base plate portion 19g so as to be integrally formed therewith. In the second modification, when the pressing forces of the loop-shaped springs 10A and 10B are applied to the spring pressing projecting pieces 19c and 19d, they can be supported by the base plate portion 19g connected to both sides of each of the spring pressing projecting pieces 19c and 19d. This is effective to ensure a sufficient strength and rigidity. The second modification has another effect. In the structures shown in FIGS. 17 and 19, as described above, the spring pressing projecting pieces 19c and 19d are manufactured by forming tongue-shaped cut pieces in the base plate portion 19g of the spring pressing member 19, bending the cut pieces from the roots thereof and flattening them. In contrast, in this modification, the spring pressing projecting pieces 19c and 19d can be simply manufactured merely by forming cut lines corresponding to both side lines of the projecting pieces 19c and 19d in the base plate portion 19g of the spring pressing member 19 and pressing the portions surrounded by the cut lines such that the portions project inward of the spring pressing member 19. This is effective to reduce the manufacturing cost. While the spring pressing member 19 is formed into an approximately cylindrical shape in the fifth embodiment, the present invention is not limited thereto. For example, like the supporting member 218 in the second embodiment, the spring pressing member 19 may be formed into an approximately polygonal shape from which a partial side portion is removed for reducing the pressure loss. As shown in FIG. 21, flow tabs 17 having the same function as that of the flow tabs 15 provided on the band 11 may be provided on the spring pressing member 19. The flow tabs 17 project outward from the outer periphery of the spring pressing member 19 for introducing the flow of a coolant in the projecting direction of the flow tab 17. The structure including the flow tabs 17 exhibits the following effect. The spring pressing member 19 at the second lattice position 7b, which is at the level in which no fuel rod 2 is present, is not required to be cooled. Accordingly, to direct the flow of a coolant passing through the spring pressing member 19 toward the other fuel rods 2 around the lattice position 7b as much as possible, the flow tabs 17 are provided on the spring pressing member 19. This makes effective use of the coolant and hence improves the effect of cooling the fuel rods 2. Sixth Embodiment A sixth embodiment of the present invention will be described with reference to FIG. 22. In this embodiment, a spring pressing member having a shape different from that in the fuel spacer 504b of the fifth embodiment is provided. Parts common to those in FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. FIG. 22 is a top view showing the structure of a fuel spacer 604b in this embodiment. In addition, the water rods 3 are also shown in the figure for more clearly showing the structure of the fuel spacer 604b. Like the fuel spacer 504b, the fuel spacer 604b shown in FIG. 22 is positioned above the upper ends of the short-length fuel rods 2b in the fuel assembly 1 shown in FIG. 2 or FIG. 3. In the fuel spacer 604b, the spring pressing member 19 of the fuel spacer 504b shown in FIG. 14 is replaced with a spring pressing member 619, formed into an umbrella shape in transverse cross-section, which has a function of the spring pressing member 19 in combination with the function of the water rod holding member 12. The structure of the spring pressing member 619 will be briefly described. Like the spring pressing member 19 shown in FIG. 14, the spring pressing member 619 includes spring holding projecting pieces (not shown) which are inserted in the loops of the loop-shaped springs 10A and 10B for holding the loop-shaped springs 10A and 10B; and spring pressing projecting pieces 619a and 619b which are brought in contact with the loops of the loop-shaped springs 10A and 10B from the outer peripheral side. The spring pressing member 619 also includes a water rod holding portion 619c which functions to hold the water rod 3 in the radial and axial directions, like, the water rod holding member 12 shown in FIG. 14. The remaining configuration is substantially the same as that of the fuel spacer 504b in the fifth embodiment. The fuel assembly including the fuel spacer 604b in this embodiment exhibits an effect comparable to that of the fuel assembly including the fuel spacer 504b in the fifth embodiment. Additionally, according to this embodiment, since the spring pressing member 19 and the water rod holding member 12 are replaced with the spring pressing member 619, it is possible to reduce the number of parts, and hence to lower the manufacturing cost; and also it is possible to reduce the transverse cross section of the portion adjacent to the water rods 3, and hence to further reduce the pressure loss. In the sixth embodiment, the structures of the spring pressing projecting pieces 619a and 619b and the spring pressing projecting pieces are not limited to those shown in FIG. 17, but may be similar to those shown in FIG. 19 or 20. In the fifth and sixth embodiments, description is made by way of example of a fuel assembly in which the fuel rods 2 are located in a square lattice array of 9 rowsxc3x979 columns; however, the present invention can be applied to a fuel assembly having a square lattice array of 8 rowsxc3x978 columns or 10 rowsxc3x9710 columns, or a rectangular lattice array in which the rows and columns are different in number. Even in such a fuel assembly, if two of the loop-shaped springs are required to be supported at one lattice position, the concept of the present invention can be applied thereto, with a result that the same effect can be obtained. In a fuel assembly having a square lattice array of 9 rowsxc3x979 columns in which the arrangement of the loopshaped springs 10 is different from that shown in FIG. 14 or FIG. 22, if two of the loop-shaped springs are required to be supported at one lattice position, the concept of the present invention can be applied thereto, with a result that the same effect can be obtained. Even in the case where three or more of the loop-shaped springs 10 are required to be supported at one lattice position, the concept of the present invention can be applied thereto. Such a modification will be described with reference to FIG. 23. FIG. 23 is a top view showing the structure of an essential portion of a fuel spacer 704b in this modification. Like the fuel spacers 504b and 604b, the fuel spacer 704b shown in FIG. 23 is positioned above the upper ends of the short-length fuel rods 2 in the fuel assembly 1 shown in FIG. 2 or 3. In the case of reviewing the arrangement of the loopshaped springs 10 in accordance with removal of the cells 9 to reduce the pressure loss, there may occur a requirement in which four free loop-shaped springs 710A to 710D (each having the same structure as that of the loop-shaped spring 10) need to be supported at one lattice position, other than those in the outermost peripheral region of the square lattice array and those adjacent to the water rods 3. To meet such a requirement, the fuel spacer 704b is provided with a cylindrical spring pressing member 719 for supporting the springs 710A to 710D in such a manner that the springs can generate suitable pressing forces. The structure of the spring pressing member 719 will be briefly described. Like the spring pressing member 19 shown in FIG. 19, the spring pressing member 719 includes four spring holding projecting pieces (not shown) which are inserted in the loops of the four loop-shaped springs 710A to 710D surrounding the spring pressing member 719 for holding the loop-shaped springs 710A to 710D; and four spring pressing projecting pieces 719a to 719d which are brought in contact with the loops of the loop-shaped springs 710A to 710D from the outer peripheral side. The spring pressing member 719 is also made as thin as possible within an allowable thickness range in terms of the structural strength of the fuel spacer for making the transverse cross-section thereof smaller than that of the cell 9, thereby reducing the pressure loss. The remaining configuration is substantially the same as that of the fuel spacer 504b in the fifth embodiment. Like the fifth embodiment, the fuel assembly including the fuel spacer 704b in this embodiment exhibits an effect of reasonably arranging the necessary minimum number of the loop-shaped springs 710 over the fuel spacer without increasing the number of kinds of springs by supporting the four free loop-shaped springs 710A to 710D at one lattice position by means of the spring pressing member 719. In the above modification, description is made by way of example of the case of providing four free loop-shaped springs 710A to 710D; however, the present invention is not limited thereto, but may be applied to the case of providing only three of the free loop-shaped springs. In this case, by providing three sets of the spring pressing projecting pieces 719a to 719c and the spring holding projecting pieces, an effect similar to that described above can be obtained. Since the variations in allowable arrangement of the loop-shaped springs can be further increased by making the spring pressing member 719 in the above modification in combination with the spring pressing members 19 and 619 in the fifth and sixth embodiments, this arrangement is expected to provide a more effective fuel spacer from the viewpoint of reduction in pressure loss. In each of the first to sixth embodiments, description is made by way of example of a fuel assembly in which the fuel rods 2 are located in a square lattice array of 9 rowsxc3x979 columns; however, the present invention is not limited thereto, but may be applied to another fuel assembly having a square lattice array of 8 rowsxc3x978 columns, 10 rowsxc3x9710 columns, or the like. In this case, an effect similar to that described above can be obtained. While the preferred embodiments of the present invention have been described using specific examples, such description is for illustrative purposes only, and it is to be understood that changes and variations may be made without departing from the spirit or scope of the following claims. |
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claims | 1. A method to assemble a support clamp assembly in a jet pump of a nuclear reactor vessel, the method comprising:attaching a cable to a first end of a shaft and lowering the shaft and cable into an elbow conduit and thermal sleeve of the jet pump;guiding at least one thermal sleeve anchor device down the cable to slide the anchor device on the shaft, wherein the anchor device abuts against a second end of the shaft;seating the anchor device against an inside surface of the thermal sleeve;extending the shaft from the anchor device, through an aperture in the elbow conduit wherein the aperture is in a wall of the elbow conduit;attaching a securing device to the first end of the shaft extending through the aperture in the elbow conduit, wherein the securing device abuts an outside surface of the wall of the elbow conduit to apply tension to the shaft, andthe tension to the shaft biases the thermal sleeve towards the wall of the elbow conduit. 2. The method of claim 1 further comprising extending the cable through an opening in a riser pipe transition from the jet pump, and inserting the shaft, cable and anchor device through the opening and through an inside of the riser pipe, elbow conduit and thermal sleeve. 3. The method of claim 2 wherein the anchor device includes a primary cruciform and a secondary cruciform, and the method includes sliding the cable through a base in the primary cruciform and then through a base of the secondary cruciform. 4. The method of claim 1 wherein the securing device includes a boss and a nut, and sliding the first end of the shaft through an aperture in the boss and attaching the nut to the first end of the shaft extending through the boss. |
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041522079 | summary | 1. Field of the Invention This invention relates to nuclear reactors and more particularly to a system for laterally restraining, compressing, and expanding the core of a reactor. 2. Description of the Prior Art Throughout the history of nuclear reactor design extreme care has been taken to ensure safe, reliable, and redundant shutdown and hypothetical accident controls. Typical of such controls are, for example, control rods maintained in an upper position by electromagnets which, upon loss of electrical power, drop the rods into the reactor core. Such devices have proved to be extremely reliable. They do, however, require some type of signal, or a loss of power, to position the control rods within the core. The signal can be in response to a multitude of continuously monitored plant parameters. It is particularly desirable to provide control mechanisms which respond directly to selected plant parameters. Such mechanisms can be referred to as "inherent" controls or shutdown systems. The most typical of these control mechanisms is the negative moderator coefficient experienced by many pressurized water cooled reactors upon a rising coolant temperature. This phenomenon, however, is not realized in many other reactor types. In view of the conservative design approach attendant the nuclear industry, it is desirable to provide additional inherently safe control mechanisms, applicable to a broad range of nuclear reactors. SUMMARY OF THE INVENTION This invention provides an inherently responsive control mechanism and core restraining device for fluid cooled nuclear reactors. It is based upon the nuclear characteristic that expanding the geometry of a core in response to an undesirable condition will decrease the reactivity of the core, and can place it in a subcritical configuration. The invention includes an electromagnetic circuit through ferromagnetic structure arranged to radially compress core fuel assemblies adjacent one another for normal plant operating conditions. An electromagnet can act upon the structure which compresses the core, and can also form an electromagnetically induced circuit incorporating adjacent core assemblies. Between the adjacent assemblies are desirably provided compressible Belleville-type springs which are compressed during normal operation and relaxed upon a selected signal or accident condition by decreasing the electromagnetic force. The invention further includes an inherent shutdown mechanism when a portion of the electromagnetic circuit is made of a ferromagnetic material which has a predetermined curie temperature and which is in heat transfer relation with the reactor coolant fluid exiting the core. Upon an accidental rise in the coolant temperature resulting in raising of the ferromagnetic material to a temperature at which the magnetic saturation decreases, the electromagnetic force is inherently reduced, allowing the adjacent assemblies and core to radially expand. The expanded configuration decreases the reactivity of the core, leading to a controlled shutdown. |
abstract | Strain matching of crystals and horizontally-spaced monochromator and analyzer crystal arrays in diffraction enhanced imaging systems and related methods are disclosed. A DEI system, including strain matched crystals can comprise an X-ray source configured to generate a first X-ray beam. A first monochromator crystal can be positioned to intercept the first X-ray beam for producing a second X-ray beam. A second monochromator crystal can be positioned to intercept the second X-ray beam to produce a third X-ray beam for transmission through an object. The second monochromator crystal has a thickness selected such that a mechanical strain on a side of the first monochromator crystal is the same as a mechanical strain on the second monochromator crystal. An analyzer crystal has a thickness selected such that a mechanical strain on a side of the first monochromator crystal is the same as a mechanical strain on the analyzer crystal. |
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046577254 | abstract | The invention relates to a core for a nuclear reactor cooled with water under pressure.. The core comprises a first group of zones (1, 2) extending over its entire height and over a fraction of its transverse section where the fuel rods contain mainly enriched uranium oxide, and a second group of zones (3) inserted between the zones (1, 2) of the first group where the rods contain mainly plutonium. The rods of zones (1, 2) of the first group emit neutrons and maintain the neutron reactions. The energy spectrum of the neutrons is in the thermal region in the zones (1, 2). The rods of the zones (3) of the second group are spaced at a distance which is appreciably smaller than the distance separating the rods of the zones (1, 2) so that the neutrons produced therein are in the high energy region.. The invention applies, in particular, to nuclear reactors with spectral shift control. |
042882907 | claims | 1. In apparatus for exchanging a control rod drive mechanism of a nuclear reactor of the type comprising a horizontal platform supported to be rotatable in a working chamber disposed below a reactor pressure vessel and a traveling carriage traveling on a rail assembly laid on said platform, the improvement which comprises a beam attached to said traveling carriage to be swingable between the vertical and horizontal portions and provided with a carrier for vertically moving said control rod drive mechanism, bolt mounting means for loosening and clamping bolts which are used to connect said control rod drive mechanism to a housing in the reactor pressure vessel; holding means secured to said beam for carrying one end of said control rod drive mechanism when said control rod mechanism is separated from said bolt mounting means and holding said control rod drive mechanism when said beam is swung to said horizontal position, said holding means being received in a position where said holding means is not interferred with said carrier when said carrier passes near said holding means; means for conveying said bolt mounting means in and out of the passage of movement of said carrier so that said control rod drive mechanism can easily be withdrawn vertically without being interferred with said bolt mounting means; and means for receiving said control rod drive mechanism when said beam has been moved to said horizontal position and for conveying said control rod drive mechanism into an inspection chamber for inspecting said control rod drive mechanism through a passage provided for the peripheral wall of said working chamber. 2. The apparatus according to claim 1 wherein a rail assembly for conveying said bolt mounting mean is disposed to said platform in parallel with the direction of traveling of said traveling carriage and a further carriage for conveying said bolt mounting device is movably mounted on said rail assembly. 3. The apparatus according to claim 2 wherein said rail assembly is disposed on both sides of the passage of movement of said carrier. 4. The apparatus according to claim 2 wherein said rail assembly is disposed on one side of the passage of movement of said carrier. 5. The apparatus according to claim 4 wherein said bolt mounting means is moved out of the passage of movement of said carrier by horizontally rotating said rail assembly. 6. The apparatus according to claim 2 wherein said bolt mounting means conveying carriage is rotatably provided with a frame, said frame having an engaging member at the front end adapted to engage said bolt mounting device so that when said frame is rotated, said bolt mounting device will be moved out of the passage of movement of said carrier. 7. The apparatus according to claim 6 wherein said bolt mounting means conveying carriage is connected to said traveling carriage disposed on said platform. 8. The apparatus according to claim 1 wherein said bolt mounting means is provided with wheels so that said bolt mounting means travels on a rail assembly disposed to said platform in parallel with the direction of traveling said traveling carriage. 9. The apparatus according to claim 8 wherein means for conveying said bolt mounting means is provided for said rail assembly. 10. The apparatus according to claim 1 wherein a rail assembly, on which said means for receiving and conveying said control rod drive mechanism travels, has an inclination substantially equal to that of a rail assembly laid on said passage provided for the peripheral wall of said working chamber. 11. The apparatus according to claim 1 wherein said rail assembly for guiding said control rod drive mechanism receiving and conveying means is constructed to be vertically rotatable about one end of said rail assembly. 12. The apparatus according to claim 1 wherein said working chamber located below the reactor pressure vessel is communicated with an outer working chamber through the passage provided for the peripheral wall of said first mentioned working chamber and said rail assembly for guiding said traveling carriage is connected to the rail assembly laid on said passage through the peripheral wall of said first mentioned working chamber so that said traveling carriage on which said beam is mounted will be moved from said first mentioned working chamber to said outer working chamber. 13. The apparatus according to claim 1 wherein said bolt mounting means receiving and conveying means is provided with a member adapted to be connected to said beam while it is held in vertical state and under the connected condition, said bolt mounting means can be moved into and out of the passage of movement of said carrier. 14. The apparatus according claim 1 wherein said means secured to said beam for holding said control rod drive mechanism comprises a holding arm provided with a pair of jaws adapted to firmly or loosely hold said control rod drive mechanism. 15. The apparatus according to claim 1 wherein said bolt mounting means includes a connecting member adapted to be connected to said traveling carriage or said beam when it is held in a vertical state, a first supporting member being movable towards or away from said beam when it is held in said vertical state and supporting said control rod drive mechanism, and a second supporting member being movable towards and away from said beam when it is held in said vertical state and supporting said bolt mounting means so as to convey it into and out of the passage of movement of said carrier. 16. The apparatus according to claim 1 wherein said bolt mounting means is provided with a differential gearing connected to the output shaft of an electric motor through torque limit means and a reduction gear, and wherein a multi-shaft automatic bolt clamping device having spanners on respective shafts is provided for a rotatary shaft of said differential gearing. |
description | This invention relates generally to the field of particle optics and waveguides, and in particular to devices for modifying or manipulating beams of particles and electromagnetic waves by influencing the wave properties of such beams. Particles of interest include atoms, ions molecules and charged particles such as electrons, and beam manipulations or modifications envisaged include modulation, tuning, diffraction, polarization and beam splitting. Applications of particular interest include electromagnetic waveguides and atom optics, the tunable diffraction-based spectroscopy of atoms, molecules and isotopes, gravimeters and related instrument, manipulation of electromagnetic waves, synchrotron optics, as well as non-lithographic deposition and patterning in the area of nanofabrication. Atom optics relies on the concept of providing beams of atoms sufficiently slowed down for their de Broglie wavelengths to be of manageable nanometer-scale dimensions. An ongoing challenge is to develop suitable optics devices that will allow beams of atoms, or of ions or molecules or charged particles, to be usefully employed for their wave-like properties. For example, interposition of an atomic lens can allow a beam of atoms from a diffuse source to be focused into an array of lines and dots of nanometer dimensions, a technique that can be applied as a novel form of nanofabrication. Such developments were described by R. J. Celotta, R. Gupta, R. E. Scholten and J. J. McClelland, in “Nanostructure fabrication via laser focused atomic deposition”, J. Appl. Phys. 79 (80, 15 Apr. 1996a; J. J. McClelland and R. J Celotta, in “Laser-Focused Atomic Deposition—Nanofabrication via Atom Optics”, pre-print, NIST; J. J. McClelland, “Nanofabrication via Atom Optics” in Handbook of Nanostructured Materials and Nanotechnology, Vol. 1, 335-385 (2000); M. R. Walkiewicz, “Manipulation of Atoms Using Laser Light”, PhD Thesis, University of Melbourne, (2000) 222 p.' J. J, McClelland, William R. Anderson, Curtis C. Bradley, Mirek Walkiewicz, Robert J. Celotta, Erich Jurdik and Richard D. Deslattes, “Accuracy of nanoscale pitch standards fabricated by laser-focused atomic deposition” NIST Journal of Research 108(2), 99-113 (2003) Feb. 14, 2003. The NIST researchers used a laser light tuned near an atomic transition to form an array of atom lenses for focusing a beam of atoms into an array of dots of a size as small as 30 nanometers. It is an object of this invention to provide a device useful in the field of electromagnetic waveguides and particle optics, and consequently in the manipulation of particle beams in the field of nanofabrication. The invention borrows a structure known in another branch of nanotechnology and modifies and extends it for the purposes of the present invention. The known structure is the micro or nano electrical conductor crossbar network, previously described in a range of contexts including a displacement or vibration-measuring system (international patent publication WO 00/14476), a memory system (U.S. Pat. No. 6,128,214) and a demultiplexer (U.S. Pat. No. 6,256,767). A micro or nano electrical conductor crossbar network comprises a set of two separate substrates, each having a two dimensional array of micro- or nano-wires (conductors) deposited on it and extending as an array of parallel lines on the substrates. The two substrates are separated by suitable distance. The arrays of parallel micro- or nano-conductors on the two substrates facing each other may be at an arbitrary angle with respect to each other, but of particular interest for some applications is the case where the arrays are at a right angle. Thus a crossbar network consists of a two dimensional array of micro or nanometer scale devices, each comprising a cross-over point or a junction formed where a pair of spaced conductors cross but do not touch one another. Each junction has a state, e.g. capacitance, or quantum tunnelling current conductance, that can be altered by applying a voltage across the respective conductors that cross at the junction. The most significant feature of the aforementioned U.S. patents is the presence of a connector species forming an electron donor-bridge-acceptor (DBA) molecular junction (a molecular switch) at each cross-over, while international patent publication WO 00/14476 does not include a specific connector species, not even calls for them, but instead relies on a sensitivity to quantum tunnelling current at the cross-over points, and discloses how the set of cross-over points will form an artificial scattering lattice effective to scatter electromagnetic wave or a beam of atoms directed parallel to the sandwich structure into the space between the conductor layers. Each conductor may be independently connected electrically, i.e. they have no common bias; there will then be a pixelised array which is an analogue of a two-dimensional “pinball game” for waves or atoms, with predefined scattering centres. This concept is further developed in the present application and broadened to include larger dimensions. Reference to the aforementioned patent publication and patents is not to be construed as an admission that their content, whether in whole or in part, is or has been common general knowledge. The invention provides apparatus for manipulating or modifying electromagnetic waves or a beam of particles, eg atoms, ions, molecules or charged particles, which includes a micro or nano electrical conductor crossbar network having multiple cross-over junctions that define respective scattering points for the particles of the beam, wherein at least one structural parameter of the crossbar network is selectively tuneable to obtain a desired manipulation or modification of said beam when incident on the network in a pre-determined direction. The invention also provides a method of manipulating or modifying electromagnetic waves or a beam of particles, eg atoms, ions, molecules or charged particles, including directing the beam as an incident beam into a micro or nano electrical conductor crossbar network in a predetermined direction, which network has multiple cross-over junctions that define respective scattering points for the particles of the beam and is arranged so that at least one structural parameter of the crossbar network is selectively tuneable to obtain a desired manipulation or modification of said beam, whereby the beam emerges from the network modified or manipulated with respect to the incident beam. In the context of this specification, references to a micro or nano electrical conductor are an indication that the conductor has a width in the micron to nanometer range. The conductors may conveniently be flat strips or wires of any suitable cross-section, and may typically be supported on a substrate. The method preferably includes, prior to directing the beam as described, tuning the crossbar network by tuning at least one structural parameter of the crossbar network with respect to the incident beam. Advantageously, the conductors of the crossbar network have a width in the range 1 nanometer to 300 microns. Preferably, the conductors are arranged in respective spaced layers each having a subset of multiple substantially parallel conductors, eg on a respective substrate. The spacings between the conductors plus insulating strip (pitch) within each layer may be in the range 1 nanometer to 500 microns, while the spacing between layers is, eg, in the range 0.5 nanometers to 200 microns between opposed conductor faces. The respective subsets of conductors can typically be supported in or on a respective insulating or semiconducting substrate. In certain applications, the conductors can be carbon nanotubes of arbitrary helicity or radius, either single or multi-walled. In one or more particular embodiments, there can be a connector species at some or all of the cross-over junctions in the crossbar network. The separation of adjacent layers can be determined and defined in any suitable manner, in some cases dependent on the presence and nature of the connector species of the crossbar network. For example, the gap between substrates supporting respective conductor layers may be an at least partial vacuum or may be filled with an appropriate medium. Suitable arrangements for accurately maintaining the gap include the use of buckyball (C60) nanobearings or nanotubes, or the interpositioning of a separation film of an organic medium, preferably organic liquid eg cyclohexane or soft matter spacer eg. Self Assembled Monolayers (SAMs). The apparatus preferably includes means to selectively tune said at least one structural parameter of the network. More easily tuneable parameters include the angle between the alignments of parallel conductors in respective layers of the wires (tuned by relatively rotating the layers), the potential difference at each separate cross-over point (tuned by varying the potential applied to the individual conductors), or the actual configuration of scattering points defined by cross-over junctions in the network (tuned by altering the configuration of “live” conductors—see FIG. 2). Less easily tuneable parameter includes the spacing between adjacent layers of the conductors. Selective turning of the tuneable parameter, where it is a spatial parameter, may be by mechanical adjustment means forming a nano or micro electromechanical system (NEMS or MEMS). For example, the adjustment means may include piezoelectric actuators of known type suitable for performing adjustments at nano- or micrometer scale dimensions. Tuning can also be achieved by electrical and computer means, through pre-programmed tuning or real-time modification of variables eg conductor potentials. In one application, the apparatus is a diffraction grating with respect to an incident particle beam, for splitting the incident particle beam into a plurality of parallel sub-beams, i.e. a diffraction pattern output. It should be noted that for passing of charged particles through the grid, for fixed polarization i.e. constant voltage between the grids, charges will drift in the overall field and will be deflected from the plane of the two grids towards oppositely charged grid. To counter this so overall the charged particles cannot “feel” the polarization, the device can have oscillating potential; i.e. oscillating from positive to negative charge. The frequency of such oscillating applied potential of the electric field will depend on the dynamics of incoming beam of charged particles (mass, charge, velocity) and geometrical characteristics of the grid (spatial extension, separation between the lines, and separation between the planes). Referring to the drawings there is shown an embodiment of the invention depicted in FIG. 1, which is a very simplified diagram of an electrical conductor crossbar network 10 configured as an atomic beam diffraction grating. The direction of wave propagation of the atomic beam is indicated by the arrow 15. The atomic beam is sufficiently slowed for it to exhibit wave behaviour having a de Broglie wavelength of the order of magnitude of the lattice spacing of a lattice of scattering points 20 defined by crossbar network 10, and is thereby diffracted so as to form a diffraction pattern on downstream image plane 30. In this way, incident beam 15 is manipulated or modified by crossbar network 10 whereby the beam emerges from the network manipulated or modified with respect to incident beam 15. Crossbar network 10 comprises respective spaced layers 12, 13 of elongated electrical conductors 16, 17 typically provided in or on respective insulating or semiconductor substrates, not shown here for purposes of enhanced illustration. There are a variety of techniques for forming crossbar network 10, well known and understood by those skilled in the art. In each layer, the electrical conductors 16 and 17 are parallel, and the two conductor arrays extend at 90° with respect to each other so as to define multiple cross-overs or nodes 25. The nodes 25 thereby form cross-over junctions at which, when the pair of conductors are energised, the resultant electrical fields define scattering points 20 in a scattering field pattern of electrical potential gradients. In a practical nanofabrication application, image plane 30 may be a substrate on which the atoms of beam 15 are being deposited in a pre-determined pattern constituting the diffraction pattern generated by the interaction between the atomic beam and the crossbar network. In a modification, a set of shutters may be placed perpendicularly between the crossbar network and the image plane (30) (which in turn can also be allowed to move in x and y directions). Typically, each conductor 16, 17 has an independent electrical connection so that discrete electrical potentials can be applied individually to each conductor of each planar layer. This is a normal feature of crossbar networks. In this way, each node 25 can be separately characterised and the network can be tuned by varying the actual array of cross-over points that are “on” and therefore acting as scattering points. The lattice spacing parameter, or lattice constant, and the configuration of scattering points can thereby be varied and constitute tuneable parameters of network 10. An example is provided in the first configuration of FIG. 2 where, by switching off every second “horizontal” conductor, a rectangular lattice is formed from the square array of conductors. Moreover, the lattice form factor (or scattering “atomic factor”) can be varied by altering the magnitude of the voltage bias at the crossover junctions. If conductor layers 12, 13 are independently mounted in a structure that allows their respective substrates to be relatively moved towards or away from each other, or to be relatively rotated, respective physical parameters can be tuned to vary the scattering pattern in other ways. For example, in the second configuration of FIG. 2 shows how a square network or lattice can be converted to a rhomboidal network or lattice by simply rotating one substrate and therefore one planar conductor array over the other. The spacing of the conductor layers 12, 13 is preferably in the range of 0.5 nanometers to 200 microns. If the spacing is less than approximately 10 to 15 nanometers, quantum tunnelling should dominate and will be observed at cross-over junctions or nodes 25, and will contribute to or constitute the mechanism by which the nodes become scattering points. At higher spacings, the cross-over points will form capacitances with a defined electrical field pattern. Computational and numerical analysis of the parameter space for the illustrated device is capable of providing optimised solutions for particular applications. In particular, it is possible to numerically compute and respectively tune or modulate in real-time the geometrical 2-D structure of the device in terms of selected variable parameters from those discussed above, for optimal and desired performance. Envisaged applications include, but are not limited to, nanofabrication and patterning using particle beams, atom writing and deposition, beamsplitters, spectroscopy of atoms, isotopes and molecules, gravimeters and several other instruments. It will be understood that the invention disclosed and defined in this specification extends to all alternative combinations of two or more of the individual features mentioned or evident from the text or drawings. |
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description | Pursuant to 35 U.S.C. § 119(a), this application claims the benefit of the earlier filing date and the right of priority to Korean Patent Applications No. 10-2017-0074396, filed on Jun. 13, 2017, the contents of which are incorporated by reference herein in their entirety. The present disclosure relates to a reactor cooling method, and more particularly, to power production using heat generated in a core and transferred to a reactor or reactor coolant system during a normal operation, and emergency power production and reactor cooling using heat generated in the core and transferred to the reactor or the reactor coolant system during an accident. Nuclear reactors are divided into loop type reactors (e.g., commercial reactors: Korea) in which major components (steam generator, pressurizer, pump, etc.) are installed outside a reactor vessel and integral reactors (e.g., SMART reactors: Korea) in which the major components are installed inside a reactor vessel. In addition, Nuclear power plants are divided into active plants and passive plants depending on the implementation of a safety system. An active plant is a reactor using an active component such as a pump operated by electric power of an emergency diesel generator (EDG) or the like to drive a safety system, and a passive plant is a plant using a passive component operated by gravity, gas pressure or the like to drive a safety system. A passive safety system in a passive plant may maintain the reactor in a safe manner only with a natural force built in the system without an operator action or an AC power source of safety grade such as an emergency diesel generator for more than a period of time (72 hours) required by regulatory requirements in the event of an accident, After 72 hours, using an operator action and a non-safety systems might be allowed to maintain the function of the safety systems and an emergency DC power source (battery). Unlike a general thermal power plant where heat generation is stopped when fuel supply is stopped, a reactor in a nuclear power plant generates residual heat from a reactor core for a significant period of time by a fission product produced and accumulated during a normal operation even when a fission reaction is stopped in the reactor core. Accordingly, a variety of safety systems for removing the residual heat of the core during an accident are installed in the nuclear power plant. In case of an active nuclear power plant (Conventional Nuclear Power Plant of Korea), a plurality of emergency diesel generators are provided in preparation for a case of interruption of electric power supply from the inside or outside at the time in an accident, and most active nuclear power plants use a pump to circulate cooling water, and thus a large-capacity emergency AC power source (a diesel generator) is provided due to the high power requirements of those active components. An operator action allowance time for an active nuclear reactor is estimated about 30 minutes. In order to exclude active components such as a pump that requires a large amount of electricity, a driven force such as gas pressure or gravity is introduced in a passive nuclear reactor (U.S. Westinghouse AP1000, Korean SMART) that has been developed or is being developed to enhance the safety of the nuclear power plant, and thus a large amount of power is not required other than small components such as a valve, which is essentially required for the operation of a passive safety system. However, to enhance the safety in a passive nuclear power plant, an operator action allowance time is drastically extended from 30 minutes to 72 hours or longer, and an emergency active power source (diesel generator) is excluded, and an emergency DC power source (battery) is adopted. And thus the emergency DC power source should be maintained for more than 72 hours. Therefore, the emergency power source capacity required per unit time in a passive nuclear power plant is relatively small compared to an active nuclear power plant, but it is very large in terms of the battery capacity because the emergency power should be supplied for 72 hours or more. In the other hand, a residual heat removal system (auxiliary feedwater system or passive residual heat removal system) is employed as a system for removing the heat of a reactor coolant system (the sensible heat of the reactor coolant system and the residual heat of the core) using a residual heat removal heat exchanger connected to a primary system or secondary system when an accident occurs in various nuclear power plants including an integral reactor. (AP1000: U.S. Westinghouse, commercial loop type nuclear power plant and SMART reactor: Korea) Furthermore, a safety injection system is employed as a system for directly injecting cooling water into the reactor coolant system in case of a loss-of-coolant accident to maintain a water level of the reactor core and removing the heat of the reactor coolant system (the sensible heat of the reactor coolant system and the residual heat of the core) using the injected cooling water. (AP1000: U.S. Westinghouse, commercial loop type and SMART reactor: Korea) Moreover, a reactor containment cooling system or spray system is a system for condensing steam using cooling or spraying to suppress a pressure rise when a pressure inside the reactor containment rises due to an accident such as a loss-of-coolant accident or a steam-line-break accident. Additionally, there are a method of directly spraying cooling water into the reactor containment (commercial loop type reactor: Korea), a method of inducing steam discharged in the reactor containment into a suppression tank (commercial boiling water reactor), a method of using a heat exchanger installed inside or outside the reactor containment (reinforced concrete containment building)) (APR+: Korea), a method of using a surface of the steel containment vessel as a heat exchanger (AP1000: U.S. Westinghouse), or the like. As described above, various safety systems configured with multiple trains with two or more trains are installed in each system such as a residual heat removal system and a safety injection system for cooling the reactor coolant system (including the reactor vessel) to protect the reactor core at the time of an accident. However, in recent years, there has been a growing demand for safety enhancement of nuclear power plants due to the impact of Fukushima nuclear power plant (boiling water reactor) accident and the like, and thus there is a rising demand for safety facilities against a severe accident such as an external reactor vessel cooling system even in a pressurized water reactor (PWR) with a very low risk of leakage of large amounts of radioactive materials due to employing a very large-internal-volume nuclear reactor containment. In detail, various safety facilities are provided to relieve an accident in case of the accident. In addition, each of the safety facilities is configured with multiple trains, and the probability that all systems fail simultaneously is very small. However, as a public requirement for the safety of nuclear power plants increases, safety facilities have been enhanced in preparation for a severe accident even with a very low probability of occurrence. The external reactor vessel cooling system is a system provided to cool the outside of reactor vessel during core meltdown to prevent damage of the reactor vessel, assuming that a serious damage occurs in the core cooling function and a severe accident that the core is melted occurs since various safety facilities do not adequately perform functions due to multiple failure causes at the time of an accident. (AP1000 U.S. Westinghouse) When the reactor vessel is damaged, a large amount of radioactive material may be discharged into the reactor containment, and a pressure inside the reactor containment may rise due to an large amount of steam generated by corium (melted core)-water reaction and gas formed by the core melt-concrete reaction. The reactor containment serves as a final barrier to prevent radioactive materials from being discharged into an external environment during an accident. When the reactor containment is damaged due to an increase in internal pressure, a large amount of radioactive material may be released to an external environment. Therefore, the external reactor vessel cooling system performs a very important function of suppressing radioactive materials from being discharged into the reactor containment and an increase of the internal pressure during a severe accident to prevent radioactive materials from being discharged into an external environment. The external reactor vessel cooling system which is adopted in many countries is a system in which cooling water is filled in the reactor cavity located at a lower part of the reactor vessel and the cooling water is introduced into the cooling flow path in a space between the thermal insulation material and the reactor vessel and then steam is discharged to an upper part of the cooling flow path. In addition, a method of injecting a liquid metal at the time of an accident to mitigate the critical heat flux phenomenon, a method of pressurized cooling water to induce single phase heat transfer, a method of modifying a surface of the external reactor vessel to increase the heat transfer efficiency, a method of forming a forced flow, and the like, may be taken into consideration. On the other hand, Stirling engine was developed by Robert Stirling (1816) as an external combustion engine, which tightly holds gas in a closed cylinder and drives an actuator (a type of piston) and a piston according to the strokes of heating, expansion, cooling and compression to produce power. The Stirling engine is classified into α type, β type, γ type, and dual-acting type according to the configuration of the cylinders and the pistons, and classified into a mechanically driven Stirling engine (kinematic engine) and a free piston Stirling engine (FPSE), and the like depending on the piston movement. Stirling engines may also operate with heat having a small temperature difference (e.g., 2° C.) and have a very high theoretical efficiency and low noise and low vibration compared to internal combustion engines. In addition, the Stirling engine may use various heat sources such as solar heat and geothermal heat, and the use of environmentally friendly heat sources has an advantage of low discharge of pollutants. In an external reactor vessel cooling system in the related art, since a thermal insulation material has to perform an appropriate thermal insulation function during a normal operation of the nuclear power plant, a flow path is sealed such that the inlet and outlet flow paths formed in the thermal insulating material at the time of an accident must be properly opened in a timely manner, and there is a delay time for filling the reactor cavity, and the heat removal ability may be reduced due to a critical heat flux phenomenon or the like while evaporating cooling water to form a steam layer on the external reactor vessel. In addition, there is also a research on cooling the external reactor using a liquid metal, but the liquid metal method has difficulties in the maintenance of the liquid metal. In addition, the method of cooling the external reactor using a pressurization method has difficulties in the application of a natural circulation flow, and the method of modifying the reactor vessel surface has difficulties in the fabrication and maintenance of the surface, and the forced flow method has a disadvantage in that it must be supplied with electric power. On the other hand, the large-capacity steam turbine method has a large size of the facility, thus increasing the cost when the strengthened seismic design is applied thereto. Therefore, there is a limitation in being designed to produce electric power during a normal operation and during an accident of the nuclear power plant. In addition, since an external reactor vessel cooling system in the related art is operated by an operator action at the time of an accident, various instruments and components for monitoring the accident are required for the operation, and a probability that a system in a standby mode fails to operate at the time of an accident is higher than a probability that a system being operated is stopped to operate at the time of an accident. Accordingly, the present disclosure presents a reactor cooling and power generation system in which a large-scale turbine power generation facility in the related art is maintained almost same design, and a small-scale power generation facility including the Sterling engine is additionally installed to receive heat generated and discharged from the core during a normal operation or during an accident of the nuclear power plant. An object of the present disclosure is to provide a reactor cooling and power generation system having system reliability in which safety class or seismic design are easily applicable, and reactor cooling is carried out while continuously operating during a normal operation as well as during an accident to produce emergency power. Another object of the present disclosure is to propose a reactor cooling and power generation system having enhanced safety in which residual heat of a certain scale or more is removed during a normal operation as well as during an accident. Still another object of the present disclosure is to propose a nuclear power plant having economic efficiency and safety due to the downsizing and reliability enhancement of an emergency power system of the nuclear power plant. A reactor cooling and power generation system according to the present disclosure may include a reactor vessel, a heat exchange section formed to receive heat generated from a core inside the reactor vessel through a fluid, and an electric power production section comprising a Sterling engine formed to produce electric energy using the energy of the fluid whose temperature has increased while receiving the heat of the reactor, wherein the system is formed to circulate the fluid that has received heat from the core through the electric power production section, and operate even during a normal operation and during an accident of the nuclear power plant to produce electric power. According to an embodiment, the electric power produced during the normal operation of the nuclear power plant may be supplied to an internal and external electric power system and an emergency battery. According to an embodiment, the electric energy charged in the emergency battery may be formed to supply an emergency electric power as an emergency power source during an accident. Furthermore, the electric power produced during an accident of the nuclear power plant may be formed to be supplied to an emergency power source of the nuclear power plant. According to an embodiment, the emergency power source may be formed to be supply an electric power source for the operation of a nuclear safety system or valve manipulating for the operation of the nuclear safety system or monitoring the nuclear safety system or operation of the reactor cooling and power generation system during an accident of the nuclear power plant. According to an embodiment, a seismic design of seismic category I, II or III may be applied thereto, and a safety grade of safety class 1, 2 or 3 may be applied thereto. According to an embodiment, the system may further include a first discharge section connected to the heat exchange section, wherein the first discharge section is formed to allow at least a part of the fluid excessively supplied to the electric power production section to bypass the electric power production section. According to an embodiment, the heat exchange section may be formed to surround at least a part of the reactor vessel, and may have a shape capable of cooling an outer wall of the reactor vessel formed to receive heat discharged from the reactor vessel that has received heat generated from the core. Furthermore, at least a part of the shape of the heat exchange section having a shape of cooling the outer wall of the reactor vessel may include a cylindrical shape, a hemispherical shape, a double vessel shape, or a mixed shape thereof. According to an embodiment, the system may be connected to an in-containment refueling water storage tank (IRWST) to supply refueling water to the heat exchange section having a shape capable of cooling the outer wall of the reactor vessel. Furthermore, a second discharge section may be provided in a heat exchange section having a shape capable of cooling the outer wall of the reactor vessel, and the second discharge section may be formed to discharge the refueling water supplied from the in-containment refueling water storage tank (IRWST). According to an embodiment, a coating member may further be formed on the heat exchange section having a shape capable of cooling the outer wall of the reactor vessel to prevent the corrosion of the reactor vessel. A surface of the coating member may be chemically treated to increase a surface area thereof. Furthermore, a heat transfer member may further be formed to efficiently transfer heat discharged from the reactor vessel. A surface of the heat transfer member may be chemically treated to increase a surface area thereof. According to an embodiment, the heat exchange section may be provided inside the reactor vessel, and may have a shape capable of cooling an inside of the reactor vessel formed to receive heat discharged from a reactor coolant system inside the reactor vessel that has received heat generated from the core. According to an embodiment, the system may be connected to an in-containment refueling water storage tank (IRWST) to supply refueling water to the heat exchange section having a shape capable of cooling an inside of the reactor vessel. Furthermore, a second discharge section may be provided in a heat exchange section having a shape capable of cooling the inside of the reactor vessel, and the second discharge section may be formed to discharge the frefueling water supplied from the in-containment refueling water storage tank (IRWST). According to an embodiment, the system may further include an evaporator section connected to the heat exchange section, wherein the evaporator section is formed to exchange heat with an inner fluid of the heat exchange section and an inner fluid of the electric power production section, and comprises a first circulation section formed to circulate through the heat exchange section and the evaporator section; and a second circulation section formed to circulate through the evaporator section and the electric power production section. According to an embodiment, at least one of the first circulation section and the second circulation section may be formed to circulate by a single-phase fluid. According to an embodiment, the heat exchange section may further include a core catcher, and the core catcher may be formed to receive and cool a melted core when the core is melted in the reactor vessel. According to an embodiment, the Stirling engine may include a power generation section including a cylinder having a reciprocator and a piston configured to generate motive power by heat received through the fluid that has received heat, and a power transmission section, and an electricity generation section configured to convert mechanical energy generated by the power generation section into electrical energy. According to an embodiment, the Stirling engine may include a high temperature section and a low temperature section respectively filled with working gas, and formed with spaces partitioned from each other inside a cylinder, and working gas filled in the high temperature section and the low temperature section may be formed to communicate with each other, and formed to move a reciprocator and a piston according to the communication of the working gas. According to an embodiment, the Stirling engine may further include a regenerative heat exchange section, and the regenerative heat exchange section may transfer and store heat stored in the working gas in the regenerative heat exchange section when the working gas moves from the high temperature section to the low temperature section, and transfer the heat stored in the regenerative heat exchange section to the working gas when the working gas returns from the low temperature section to the high temperature section. According to an embodiment, a fan or a pump may be provided in the low temperature section, and the fan or the pump may be formed to supply a cooling fluid to the low temperature section to exchange heat with the working gas of the low temperature section. The cooling fluid may include air, pure water, seawater, or a mixture thereof. According to an embodiment, the system may further a condensate storage section at a lower portion of the electric power production section to collect condensate generated by condensing the fluid heat-exchanged in the electric power production section. According to an embodiment, condensate in the condensate storage section may be supplied to the heat exchange section by gravity or the power of the pump. A nuclear power plant according to the present disclosure may include a reactor vessel, a heat exchange section formed to receive heat generated from a core inside the reactor vessel through a fluid, and an electric power production section comprising a Sterling engine formed to produce electric energy using the energy of the fluid whose temperature has increased while receiving the heat of the reactor, wherein the system is formed to circulate the fluid that has received heat from the core in the heat exchange section through the electric power production section, and formed to operate even during a normal operation and during an accident of the nuclear power plant to produce electric power. The reactor cooling and power generation system according to the present disclosure is formed to drive an electric power production section including a Sterling engine formed to produce electric energy using the energy of a fluid in a small scale facility. A heat exchange section and an electric power production section of the present disclosure may continuously operate not only during normal operation but also during an accident to cool residual heat and generate emergency power, thereby improving system reliability. A heat exchange section for facilitating the application of safety class and seismic design with a small scale facility, and continuously operating during an accident as well as during a normal operation and performing reactor cooling by the application of safety class or seismic design may be included therein, thereby improving the reliability of the nuclear power plant. The reactor cooling and power generation system according to the present disclosure may be designed to remove residual heat of a certain scale or more generated from the core of the reactor, and continuously operated not only during a normal operation but also during an accident to reduce a probability of actuation failure at the time of an accident, thereby improving the safety of the nuclear power plant. The nuclear power plant according to the present disclosure may improve the economic efficiency of the nuclear power plant through the downsizing of an emergency power system through the reactor cooling and power generation system. Hereinafter, preferred embodiments of the present disclosure will be described in detail with reference to the accompanying drawings, and the same or similar elements are designated with the same numeral references regardless of the numerals in the drawings and their redundant description will be omitted. In describing the present disclosure, if a detailed explanation for a related known function or construction is considered to unnecessarily divert the gist of the present disclosure, such explanation has been omitted but would be understood by those skilled in the art. The accompanying drawings are used to help easily understand the technical idea of the present disclosure and it should be understood that the idea of the present disclosure is not limited by the accompanying drawings. It will be understood that although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are generally only used to distinguish one element from another. A singular representation may include a plural representation unless it represents a definitely different meaning from the context. Terms “include” or “has” used herein should be understood that they are intended to indicate an existence of several components or several steps, disclosed in the specification, and it may also be understood that part of the components or steps may not be included or additional components or steps may further be included. Hereinafter, a reactor cooling and power generation system provided with an heat exchange sections 120, 220, 320a, 320b, 320c, 320d, 320e, 420 having a shape capable of cooling an external reactor vessel formed to receive heat discharged from a reactor vessel that has received heat generated in a core to operate even during a normal operation and during an accident of the nuclear power plant so as to produce power will be described in more detail with reference to FIGS. 1A through 1E, FIGS. 2A through 2E, FIGS. 3A through 3E, and FIG. 4. FIG. 1A is a conceptual view of a reactor cooling and power generation system 100 associated with an embodiment of the present disclosure. In an embodiment of the present disclosure, the reactor coolant system 111 may be circulated inside the reactor vessel 110. In addition, a heat insulator 116 surrounding a part of the reactor vessel 110 may be formed. Furthermore, the inside of the reactor vessel 110 may be formed to have a core 114. The core 114 refers to a nuclear fuel. The reactor vessel 110 may be a pressure vessel designed to withstand high temperatures and pressures because electric power is produced by heat generated while performing fission in the core 114. Even when a control rod is inserted into the core 114 to stop the core 114 during a nuclear power plant accident, residual heat may be generated for a considerable period of time. When it is assumed that various safety and non-safety systems do not operate at the time of an accident of the nuclear power plant though having a very low probability of occurrence, cooling water inside the reactor vessel 110 may be lost to increase the temperature of the nuclear fuel, thereby causing a core meltdown phenomenon. On the other hand, during a normal operation of the nuclear power plant, heat may be received from the reactor coolant system 111 to produce steam. The steam generator 113 may be a pressurized water reactor. Further, the steam produced by the steam generator 113 may be steam that is phase-changed by receiving water a feedwater system 10 through a main feedwater line 11 and an isolation valve 12 connected thereto. The steam produced by the steam generator 113 is passed through a main steam line 14 connected to an isolation valve 13 and supplied to a large turbine 15 and a large generator (not shown) to produce electric power while the fluid energy of the steam is converted into electric energy through mechanical energy. However, although the pressurized water reactor is illustrated in the present disclosure, the technology of the present disclosure is not limited to the pressurized water reactor. In addition, a reactor coolant pump 112 may circulate coolant that fills the inside of the reactor vessel 110. A pressurizer 115 provided inside the reactor vessel 110 may be formed to control the pressure of the reactor coolant system 111. Moreover, a passive residual heat removal system including an emergency cooling water storage section 20 and a heat exchanger 21 may be provided therein to discharge the heat of the reactor coolant system 111 to the emergency cooling water storage section 20 through the steam generator by natural circulation due to a two-phase flow received through lines 22, 23 and the opening and closing of a valve 24 during an accident. Further, when steam is generated while emergency cooling water is evaporated by heat transferred to the emergency cooling water storage section 20, the steam may be released through a steam discharge section 25 to discharge the transferred heat to the atmosphere. The reactor cooling and power generation system 100 is in an actuation state even during a normal operation and heat is continuously transferred to the reactor vessel 111 by residual heat generated from the core 114 until the temperature of the reactor vessel 110 is significantly reduced to reach a safe state, and thus the reactor cooling and power generation system 100 continues to operate. Accordingly, an operator action for the reactor cooling operation, various measuring instruments and control systems, valve operation or pump start and the opening and closing of a thermal insulation material may not be required as in a conventional method, and thus a probability of actuation failure of the reactor cooling and power generation system 100 is greatly reduced to improve the safety of the nuclear power plant. In addition, since emergency power can be stably produced by the reactor cooling and power generation system 100 until the temperature of the reactor vessel is reduced to reach a safe state during an accident, the capacity of an emergency DC battery (emergency electric power source) may be decreased to improve the economic efficiency of the nuclear power plant and improve the reliability of an emergency power system of the nuclear power plant by securing the emergency power supply means of a safety system, thereby improving the safety of the nuclear power plant. In detail, in case of a passive nuclear power plant, emergency power required during an accident is less than about 0.05% compared to the power generation capacity generated from the nuclear power plant during a normal operation, but it is designed to use battery for 72 hours or more, and thus a very large sized battery is required, having a disadvantage of increasing the cost. However, the reactor cooling and power generation system 100 may produce an appropriate level of emergency power using residual heat continuously generated from the core 114 (an amount of residual heat generated is several % (initial shutdown) to 1/several % (after 72 hours subsequent to shutdown) compared to a normal amount of thermal power). Moreover, when power is produced using the in-vessel cooling and power generation system 100, the power production amount is several tens of kWe to several MWe, and the capacity is less than 1/several % compared to the feedwater system 10 and the large turbine 15 for a normal operation of the nuclear power plant. This system 100 has almost no influence on the operation of nuclear power plant, and therefore, even when this system 100 fails during a normal operation. Than is to say, this system 100 has a capacity less than 1/several %, so it has little effect on a nuclear power plant operation. In addition, when power is produced using the reactor cooling and power generation system 100, it may be constructed in a small scale compared to the large capacity feedwater system 10 and the large turbine 15 for producing normal power, and therefore, it is easy to apply seismic design and safety class, and cost increase is not so great due to small facilities even when seismic design and safety class are applied. Besides, even in the event of an accident, it operates continuously as a normal operation without any additional valve operation, and therefore, during an accident, the probability of actuation failure of valves, pumps, and the like for the actuation of the reactor cooling system in the related art, and the probability of actuation failure or breakdown due to an error measuring instruments and control signals may be significantly reduced. Moreover, when the heat exchange section 120 and the electric power production section 130 fail due to the occurrence of a severe accident, a flow path through the in-containment refueling water storage tank (hereinafter, referred to as IRWST) 170 and the first discharge section 175 is already formed, and therefore, it may be formed to efficiently supply and discharge a flow rate of cooling water by a simple operation such as opening or closing a valve according to an operator action, and used for the cooling of the reactor coolant system 111 and core melt including the reactor vessel 110. In particular, in case of an integral reactor, a lower space of the inner and outer reactor vessel has a simple structure, and the lower or other space of the inner and outer reactor vessel is easily secured, and thus it is easier to apply the reactor cooling and power generation system 100 of the present disclosure. In addition, the reactor cooling and power generation system 100 may be used as an additional residual heat removal means that performs the role of removing the residual heat of the reactor core 114 during an accident. Hereinafter, the reactor cooling and power generation system 100 according to the present disclosure will be described in detail. The inside of the reactor containment boundary 1 (not shown) (hereinafter also referred to as a containment or reactor containment) may include a reactor vessel 110, a heat exchange section 120, and an IRWST 170. The heat exchange section 120 may be formed to receive heat generated from the core 114 inside the reactor vessel 110 through a fluid. In an embodiment, the heat exchange section 120 may be configured to enclose at least a part of the reactor vessel 110. In other words, the heat exchange section 120 may be formed to receive heat discharged from the reactor vessel 110 and cool the outer wall of the reactor vessel 110. On the other hand, the outside of the reactor containment boundary 1 includes an electric power generation section 130 and a condensate storage section 150. The electric power generation section 130 may be connected to the motors 135, 152 and the power (electric power) system 160 to supply power. The power system 160 may include an internal and external electric power system 161, a charger 162, an emergency power consuming component 164, and an emergency battery 163. However, some of the components illustrated as being installed outside the reactor containment boundary 1 may be disposed inside the reactor containment boundary 1 depending on the layout characteristics of the nuclear power plant. The reactor vessel 110 formed inside the reactor containment boundary 1 may be a pressure vessel formed to circulate the reactor coolant of the reactor coolant system 111 and formed to include the core 114 therein, and designed to withstand high pressures. The heat exchange section 120 is provided outside the reactor vessel 110 to receive heat transferred from the reactor coolant system 111 to the reactor vessel 110 at an outside of the reactor vessel 110. In detail, residual heat produced in the core 114 may be transferred to an inner surface of the reactor vessel 110 through the circulation of the reactor coolant system 111, and the transferred heat may be transferred to an outer surface of the reactor vessel 110 by the conduction heat transfer of the reactor coolant system 110 and then transferred to the heat exchange section 120, thereby allowing the heat exchange section 120 to perform cooling for the reactor vessel 110. In other words, the heat exchange section 120 may perform cooling on the reactor vessel 110 and reactor coolant inside the reactor vessel 110 during a normal operation of the nuclear power plant, and perform cooling on the reactor vessel 110, reactor coolant and core melt during a nuclear accident. In an embodiment, the heat exchange section 120 may be formed to surround a lower portion of the reactor vessel 110 and may be a heat exchange section having a shape capable of cooling the outer wall of the reactor vessel 110 using a fluid that receives heat discharged from the reactor vessel 110. In another embodiment, the heat exchange section may be provided inside the reactor vessel 110, and may have the form of a heat exchange section provided inside the reactor vessel formed to receive heat from the reactor coolant system 111 inside the reactor vessel 110 that has received heat generated from the core 114. The heat exchange section having the form of a heat exchange section provided inside the reactor vessel will be described later with reference to FIGS. 5A through 5C, 6A through 6C, and 7A through 7C. In an embodiment, when the heat exchanging section 120 is a heat exchanging section having a shape capable of cooling the outer wall of the reactor vessel 110, the shape of the heat exchange section 120 may be cylindrical. However, the shape of the heat exchange section 120 is not limited to a cylindrical shape, and at least a part of the heat exchange section 120 may include a cylindrical shape, a hemispherical shape, and a double vessel shape. In addition, the heat exchange section 120 having a shape capable of cooling the outer wall of the reactor vessel may further include a coating member 121 for preventing corrosion or increasing heat transfer efficiency. In an embodiment, the surface of the coating member 121 may be reformed in various ways, and may also be processed in an uneven shape (cooling fin) to increase the heat transfer surface area. Further, the surface of the coating member 121 may further include a heat transfer member (not shown) that can be chemically treated to increase the surface area so as to improve heat transfer efficiency. In other words, the surface of the coating member 121 and the heat transfer member may be chemically treated to increase the surface area, thereby efficiently performing heat transfer. In addition, the heat exchange section 120 is provided with a discharge pipe 122, and the discharge pipe 122 may be connected to the heat exchange section 120 and the electric power production section 130 to supply the fluid of the heat exchange section 120 to the electric power production section 130. The discharge pipe 122 may be branched to a pipe 124 passing through the valve 123 and connected to the electric power production section 130. On the other hand, the discharge pipe 122 may include a first discharge section 126 connected to the valve 125, and the first discharge section 126 may be formed to discharge at least a part of the fluid excessively supplied to the electric power production section 130 or allow it to bypass the electric power production section 130. Specifically, the first discharge section 126 may be configured discharge a part of fluid (gas, steam) when a pressure of the system rises or the fluid (liquid) is excessively supplied to a pipe for discharging the fluid (gas, steam) from the heat exchange section 120 to the outside of the reactor containment (not shown). In the present disclosure, the first discharge section 126 is illustrated to discharge a fluid to the outside of the reactor containment (not shown), but may also be formed to allow the discharged fluid to bypass the power production section 130 and then condensate the fluid for reuse according to the characteristics of the nuclear power plant. Moreover, the heat exchange section 120 may be connected to the IRWST 170 to supply refueling water through the pipe 173. Specifically, the IRWST 170 may be connected to the valve 171 and the check valve 172. As a result, a second discharge section 175 connected to the valve 174 may be provided to discharge the refueling water supplied from the IRWST 170 to the pipe 173 through the second discharge section 175 during an accident. Specifically, the second discharge section 175 is configured to cool the inside and outside of the reactor vessel 110 even when cooling and power generation using the heat exchange section 120 and the electric power production section 130 is not carried out due to a failure thereof or the like during a severe accident on a pipe for discharging the refueling water received from the IRWST 170 into the reactor containment (not shown) as a fluid (gas/steam, a mixture of gas/steam and liquid/hot water or liquid/hot water), and the like. Meanwhile, the fluid may be transferred and injected into the electric power production section 130 from the heat exchange section 120. The electric power production section 130 may be formed to produce electric energy using the energy of the fluid, and may include a Stirling engine 140. Specifically, the Stirling engine 140 may include a battery (not shown) formed to provide power for an initial engine operation. In addition, the Stirling engine 140 may have a heat exchanger provided adjacent to the Stirling engine 140. The heat exchanger includes a first heat exchange section 131 and a second heat exchange section 132. The first heat exchange section 131 of the heat exchanger may be connected to a high temperature section 141 and formed to transfer the heat of the heat exchange section 120, namely, heat generated from the core 114, to the high temperature section 141, and a fluid that has exchanged heat with the high temperature section 141 may be discharged to a pipe 139. Furthermore, the second heat exchange section 132 of the heat exchanger may be connected to a low temperature section 142 to discharge the heat of the working gas transferred from the low temperature section 142 to the outside. The Stirling engine 140 may further include a regenerative heat exchange section 133 (regenerator). The regenerative heat exchange section 133 is a component for increasing the performance and efficiency of the Stirling engine 140, and the working gas reciprocated between the high temperature section 141 and the low temperature section 142 may be located between the first heat exchange section 131 and the second the second heat exchange section 132 that receive heat (heating, cooling). The regenerative heat exchange section 133 may block heat flowing from the high temperature section 141 to the low temperature section 142 to the maximum to maintain a large temperature difference. The regenerative heat exchange section 133 may store the heat of the working gas when the working gas moves from the high temperature section 141 to the low temperature section 142, and transfer heat to the working gas again when the working gas comes to the high temperature section 141 from the low temperature section 142. Specifically, the efficiency of the Stirling engine 140 is determined by a temperature difference between the high temperature section 141 and the low temperature section 142, and thus the efficiency may be maximized when the gas returning from the low temperature section 142 to the high temperature section 141 is not reheated. In this regard, the regenerative heat exchange section 133 has a close relationship with the performance of the Stirling engine 140. On the other hand, the Stirling engine 140 includes a power generation section 143 and an electricity generation section 144. The power generation section 143 may include a cylinder 143c including a reciprocator 143a, which is a type of piston formed to generate power while working gas go through the processes of heating, expansion, cooling, and compression) by heat (heating, cooling) received through the fluid that has received heat, and a piston 143b, and a power transmission section 143d. The power transmission section 143d may transfer a reciprocating movement of the reciprocator 143a and the piston 143b inside the cylinder 143c to a connecting rod to be converted into a rotational movement. Meanwhile, the electricity generation section 144 is formed to convert mechanical energy for performing a rotational movement generated by the power generation section 143 into electrical energy. In detail, the electricity generation section 144 is connected to a power section shaft and a generator shaft to convert mechanical energy into electrical energy. In other words, the Sterling engine 140 may receive heat with a predetermined scale from the inside of the reactor vessel 110 to produce electricity in consideration of characteristics during a normal operation and during an accident of the nuclear power plant. Moreover, the Stirling engine 140 may include a high temperature section 141 and a low temperature section 142, which are respectively filled with the working gas therein, and formed with spaces partitioned from each other inside the cylinder 143c, and the working gas filled in the high temperature section 141 and the low temperature section 142 are formed to communicate with each other through the first heat exchange section 131, the second heat exchange section 132 and the regenerative heat exchange section 133, and formed to move the reciprocator 143a and the piston 143b in accordance with the communication of the working gas. In addition, the Stirling engine 140 may further include a regenerative heat exchange section 133 that transfers the stored heat to the working gas when the working gas returns from the low temperature section 142 to the high temperature section 141. Moreover, the high temperature section 141 and the low temperature section 142 are respectively formed in closed spaces inside the cylinder, and the high temperature section and the low temperature section are respectively filled with the working gas. In another embodiment, the high temperature section and the low temperature section may be formed separately inside two cylinders according to circumstances. On the other hand, as illustrated in the drawing, the high temperature section 141 and the low temperature section 142 may be formed in spaces partitioned from each other inside one cylinder 143c (beta method). The working gas may be any one of air, helium, and hydrogen. Besides, the cylinder 143c of the Stirling engine 140 may have a cylindrical shape with one side opened, and may include a piston 143b disposed at a boundary position between the high temperature section 141 and the low temperature section 142 to partition the high temperature section 141 from the low temperature section 142 inside the cylinder 143c, and a reciprocator 143a disposed inside the cylinder 143c to be spaced apart from the piston 143b. The piston 143b and the reciprocator 143a may be independently moved along the inside of the cylinder 143c by the working gas. The Stirling engine 140 may include a power transmission section 143d rotatably disposed so as to be spaced apart from the opening side of the cylinder 143c. The reciprocating linear kinetic energy of the piston 143b and the reciprocator 143a may be converted into a rotational movement to implement the continuous operation of the Stirling engine 140. The high temperature section 141 and the low temperature section 142 may be connected to each other by a connecting flow path. For example, one end portion of the connecting flow path is connected to the high temperature section 141, the other end portion of the connecting flow path is connected to the low temperature section 142, and thus the working gas flows from the high temperature section 141 to the low temperature section 142 or from the high temperature section 142 to the high temperature section 141 through the connecting flow path. In addition, the Sterling engine 140 may further include a regenerative heat exchange section 133 disposed in the connecting flow path. Moreover, the low temperature section 142 is provided with a fan 136 or a pump (not shown), and the fan 136 or the pump supplies cooling fluid to the low temperature section 142 to exchange heat with the working gas of the low temperature section 142. In detail, the fan 136 or the pump may be operated by supplying electric power produced by the electricity generation section 144 of the Stirling engine 140 to the motor 135 through a connected line 134. In an embodiment, the cooling fluid may include air, pure water, seawater, or a mixture thereof, and may not be limited to the described materials, and any material may be used without particular limitation as long as it is heat exchangeable with the low temperature section 142. The basic operation of the Stirling engine 140 by the Stirling cycle principle consists of heating, expansion, cooling and compression processes. The above processes will be described in detail as follows. Heating: When heating the high temperature section 141 in which the working gas mainly is collected, the temperature rises and the piston 143b of the high temperature section 141 is pushed out to generate work or power. Expansion: Subsequently, while the temperature of the high temperature section 141 rises, the piston 143b is pushed out, and at the same time, the working gas moves to the low temperature section 142 to push out the reciprocator 143a. At this time, the working gas stores heat in the regenerative heat exchange section 133, and at the same time, starts to be cooled while passing through a side of the low temperature section 142. Cooling: Though the reciprocator 143a is pushed out while the working gas continues to move to the low temperature section 142, the piston 143b starts to return to its original position as the working gas of the high temperature section 141 becomes insufficient. Compression: When the working gas is mainly collected in the low temperature section 142, the temperature of the working gas is lowered and gradually compressed, so that the reciprocator 143a returns to its original position while the working gas moves to the high temperature section 141 little by little. As described above, as the Stirling engine 140 moves by one cycle of the heating-expansion-cooling-compression processes, and the piston 143b and the reciprocator 143a are mechanically connected to each other, a continuous cycle operation that generates power may be achieved in the power transmission section 143d. Furthermore, some types of the Stirling engine 140 has been illustrated, but the present disclosure is not limited to the Stirling engine 140 as in a presented manner, and various types of Stirling engines may also be applied thereto. In an embodiment, electric power produced in the electric power production section 130 may be variably constructed in consideration of a heat transfer rate due to heat generated in the core 114 supplied during an accident to control a load of the power production section 130 according to the heat transfer rate. In addition, the Sterling engine 140 of the electric power production section 130 may be a small-capacity electric power production facility, which makes it easy to apply seismic design or safety class described below. Various embodiments of the Stirling engine applied to the reactor cooling and power generation system of the present disclosure will be described with reference to FIGS. 8 through 10. The electric power that can be generated by the power production section 130 has a capacity of several tens of kWe to several MWe, which is less than 1% compared to the large-capacity feedwater system 10 and the large turbine 15 for producing the normal power of the nuclear power plant, and even when the facility operates or fails, there is little influence on the operation of the large capacity feedwater system 10 and the large turbine 15 for producing normal nuclear power. In other words, the large capacity feedwater system 10 and the large turbine 15 for producing normal power are one of the biggest large-scale facilities of the nuclear power plant, and applying the seismic design and safety class above a certain scale to the whole facilities is very uneconomical because it causes a huge cost increase. On the other hand, in case of the reactor cooling and power generation system 100 to which the Sterling engine 140 is applied, a size of the system 100 is much smaller than that of the feedwater system 10 and the large turbine 15, and thus it is easy to apply seismic design or safety class thereto, and the increased cost by applying seismic design or safety class is not so large. The Sterling engine 140 may be continuously driven to supply emergency power even when it is difficult to supply power due to the occurrence of an earthquake since seismic design is applied to the reactor cooling and power generation system 100, and the Sterling engine 140 may be continuously driven to supply emergency power even when various accidents occur since safety class is applied to secure system reliability. Considering that electric power required in case of a passive nuclear power plant during an accident is several tens of kWe though the emergency power has a difference according to the characteristics of the nuclear power plant, sufficient power may be supplied with only electric power produced by the Sterling engine 140. Besides, since the emergency DC battery capacity of a passive nuclear power plant is not greater than the emergency power required by an active nuclear power plant, the DC battery may be recharged by power produced by the operation of the Sterling engine 140. The reactor cooling and power generation system 100 may be formed to have a seismic design of seismic category I to III specified by ASME (American Society of Mechanical Engineers). Specifically, seismic category I is applied to structures, systems and components classified as safety items, and should be designed to maintain an inherent “safety function” in case of a safe shutdown earthquake (SSE), and the safety function is maintained even under the operating basis earthquake (OBE) in synchronization with a normal operation load, and the appropriate allowable stresses and changes are designed to be within limits. Though not requiring nuclear safety or continuous functions, seismic category II is applied to an item in which structural damages or interactions of the items may reduce the safety functions of seismic category I structures, systems and components or result in damage to the operator. In detail, seismic category II structures, systems and components are not required to have functional integrity for a safety shutdown earthquake, but required only to have structural integrity. In addition, seismic category II structures, systems and components should be designed and arranged so as not to impair the safety-related operation of seismic category I items. Seismic category III is designed according to uniform building codes (UBCs) or general industrial standards according to the individual design function. The reactor cooling and power generation system 100 may be configured to have a safety rating of safety ratings 1 to 3 of the reactor plant specified by the American Society of Mechanical Engineers (ASME). In detail, the safety class of a nuclear power plant is largely divided into safety class 1 through safety class 3. Safety class 1 is a class assigned to a RCS (reactor coolant system) pressure-boundary portion of a facility and its support that constitute a reactor coolant pressure boundary (a portion that may result in a loss of coolant beyond a normal make-up capacity of the reactor coolant in the event of a failure). Safety class 2 may be assigned to a pressure-boundary portion of the reactor containment building and its support, and assigned only to a pressure-resistant portion of a facility and its support that perform the following safety functions while not belonging to safety class 1. A function of preventing the release of fission products or detaining or isolating radioactive materials in the containment building A function of removing heat or radioactive materials generated in the containment building (e.g., containment building spray system), a function of increasing a negative reactivity to make the reactor in a subcritical state in case of an emergency or suppressing an increase of positive reactivity through a pressure boundary facility (e.g., boric acid injection system) A function of supplying coolant directly to the core during an emergency to ensure core cooling (e.g., residual heat removal, emergency core cooling system) and a function of supplying or maintaining sufficient reactor coolant for cooling the reactor core during an emergency (e.g., refueling water storage tank) Safety class 3 is not included in safety classes 1 and 2, and may be assigned to a facility that performs one of the following safety functions: A function of controlling the concentration of hydrogen in the reactor containment building within the allowable limit A function of removing radioactive materials from a stable space outside the reactor containment building (e.g., reactor control room, nuclear fuel building) with safety class 1, 2 or 3 facilities A function of increasing a negative reactivity to make or maintain the reactor in a subcritical state (e.g., boric acid make-up) A function of supplying or maintaining sufficient reactor coolant for core cooling (e.g., Reactor coolant replenishment system) A function of maintaining a geometric structure inside the reactor to ensure core reactivity control or core cooling capability (e.g., core support structure) A function of supporting or protecting the load for safety class 1, 2 or 3 facilities (concrete steel structures not included in KEPIC-MN, ASME sec. III). A function of shielding radiation for people outside the reactor control room or nuclear power plant A cooling maintenance function of spent wet storage fuel (e.g., spent fuel vault and cooling system) A function of ensuring safety functions performed by safety class 1, 2 or 3 facilities (e.g., a function of removing heat from safety class 1, 2 or 3 heat exchangers, a safety class 2 or 3 pump lubrication function, a fuel feeding function of emergency diesel engine) A function of supplying activation electric power or motive power to safety class 1, 2 or 3 facilities A function of allowing safety class 1, 2 or 3 facilities to provide information for manual or automatic operation required for the performance of safety functions or controlling the facilities A function of allowing safety class 1, 2 or 3 facilities to supply power or transmit signals required the performance of safety functions Manual or automatic interlocking function to ensure or maintain safety class 1, 2 or 3 facilities to perform appropriate safety functions A function of providing appropriate environmental conditions for safety class 1, 2 or 3 facilities and an operator A function corresponding to safety class 2 to which standards for the design and manufacture of pressure vessels, KEPIC-MN, ASME Sec. III, are not applied On the other hand, a fluid discharged through heat exchange with the high temperature section 141 is transferred to the condensate storage section 150 along the pipe 139. In detail, the condensate storage section 150 may be disposed at a lower portion of the electric power production section 130 to collect condensate being condensed and discharged while exchanging heat with the fluid of the high temperature section 141. However, in an embodiment of the present disclosure, it may be constructed such that the condensate generated in the high temperature section 141 is transferred to the condensate storage section 150 by gravity. However, it may also be constructed such that a pump (not shown) is installed between the pipe 139 and the condensate storage section 150 according to the characteristics of the nuclear power plant to forcibly transfer the condensate. The condensate collected in the condensate storage section 150 may be circulated through the heat exchange section 120 and the electric power production section 130. Moreover, the condensate storage section 150 may be connected to the heat exchange section 120 and the pipe 156 to supply the condensate to the heat exchange section 120. The condensate may be supplied to the pipe 156 through the pipes 159 and 151 connected to the condensate storage section 150. Specifically, the condensate in the condensate storage section 150 may be supplied to the pipe 156 connected to the heat exchange section 120 through the valve 154 and the check valve 155 by the motor 152 and the small pump 153 connected to the pipe 151. Furthermore, the condensate may be supplied to the pipe 156 connected to the heat exchange section 120 by gravity through the valve 157 and the check valve 158 connected to the pipe 159 of the condensate storage section 150. The motor 152 may be supplied with electric power produced by the power production section 130 itself through a connected line 138. In addition, the motor 152 may be provided to charge electric power produced by the power production section 130 to the emergency battery 163, and receive electric power from the emergency battery 163. On the other hand, in another embodiment, a fluid exchanging heat with the high temperature section 141 through the pipe 124 may be driven by a single-phase fluid that does not undergo phase change due to heat discharged from the core 114. When the single-phase fluid circulating through the heat exchange section 120 and the electric power production section 130 is a gas, the gaseous fluid may be circulated without having the above-mentioned condensate storage section. Moreover, when the single-phase fluid is a liquid, the liquid may be circulated through the heat exchange section 120 and the electric power production section 130 without having the condensate storage section described above with the aid of a pressurizer and a pressure control section. The power system 160 may be formed to use the power produced during the foregoing normal operation of the nuclear power plant as the power of the internal and external power (electric power) system 161. In detail, the internal and external electric power system 161 may be a system for processing electricity supplied from an on-site large turbine generator, a power production section 130, an on-site diesel generator, and an external electric power grid. In addition, electric energy may be stored in the emergency battery 163 through a charger 162, which is a facility for storing alternating current (AC) electricity supplied from the on-site, the outside, or the power production section 130 or the like. The emergency battery 163 may be a battery provided in a nuclear power plant on-site to supply emergency DC power used during an accident. Further, the electric energy stored in the emergency battery 163 may be supplied to the emergency power consuming component 164 and used as an emergency power source. The emergency power source may be used as a power source for operating the nuclear power plant safety system or opening or closing a valve for the operation of the nuclear power plant safety system or monitoring the nuclear safety system during an accident of the nuclear power plant. Moreover, the electric power produced by the power production section 130 during an accident of the nuclear power plant may also be formed to be supplied to the emergency power source of the nuclear power plant. Moreover, when the heat exchange section 120 and the power production section 130 fail due to the occurrence of a severe accident, a flow path through the IRWST 170 and the first discharge section 175 is already formed, and therefore, it may be formed to efficiently supply and discharge a flow rate of cooling water by a simple operation such as opening or closing a valve according to an operator action to cool the reactor vessel 110. FIG. 1B is a conceptual view illustrating the operation of the reactor cooling and power generation system 100 during a normal operation of the nuclear power plant associated with an embodiment of the present disclosure. Referring to FIG. 1B, it is a conceptual view illustrating system arrangement and supercritical fluid flow during a normal operation of the nuclear power plant. Main feedwater (water) is supplied from the feedwater system 10 to the steam generator 113, and heat received from the core 114 by the reactor coolant circulation of the reactor coolant system 111 is transferred the secondary system through the steam generator 113 to increase the temperature of the main feedwater and produce steam. The steam produced from the steam generator 113 is supplied to the large turbine 15 along the main steam line 14 to rotate the large turbine 150 and rotate the large generator (not shown) connected through the shaft to produce electric power. The power produced through the large generator may supply electricity to an on-site or off-site from the power system. Meanwhile, feedwater supplied from the small pump 153 to the heat exchange section 120 through the pipe 156 may receive heat while moving upward along the outer wall of the reactor vessel 110 to produce steam. The steam may be supplied to the electric power production section 130 including the Stirling engine 140 along the discharge pipe 122 disposed at an upper portion of the heat exchange section 120, and the thermal energy of the steam may be converted into mechanical energy while operating the Sterling engine 140, and the mechanical energy may be converted into electrical energy by the electricity generation section 144 connected to the shaft to produce electric power. Further, electric power produced by the power production section 130 may be formed to use the electric power as the electric power of the internal and external power (electric power) system 161 through the power system 160. In addition, electric energy may be stored in the emergency battery 163 through a charger 162, which is a facility for storing alternating current (AC) electricity supplied from the on-site, the outside, or the power production section 130 or the like as emergency power. The emergency battery 163 may be a battery provided in the on-site to supply emergency DC power used during an accident. Further, the electric power may be supplied to the emergency power consuming component 164 and used as an emergency power source. In addition, a fluid discharged while exchanging heat with the high temperature section 141 of the Stirling engine 140 may be heat-exchanged with the high temperature section 141 and condensed to discharge condensate, and the condensate may be collected in the condensate storage section 150 along the pipe 139. The condensate collected in the condensate storage section 150 may be circulated through the heat exchange section 120 and the electric power production section 130. Moreover, the condensate storage section 150 may be connected to the heat exchange section 120 and the pipe 156 to supply the condensate to the heat exchange section 120. As described above, during a normal operation of the nuclear power plant, the reactor cooling and power generation system 100 may be operated simultaneously with the nuclear power generation facility. FIG. 1C is a conceptual view illustrating a forced circulation operation of the reactor cooling and power generation system 100 during a nuclear design basis accident associated with an embodiment of the present disclosure. Referring to FIG. 1C, it is illustrated a conceptual view of the operation of the reactor cooling and power generation system 100 when the operation of the small pump 153 and the electric power production section 130 is enabled during a nuclear design basis accident. Specifically, when an accident occurs in a nuclear power plant due to various causes, safety systems such as a passive residual heat removal system, a passive safety injection system and a passive containment cooling system, including the emergency cooling water storage section 20, which are installed in a plurality of trains, may operate automatically. Further, steam generated by the operation of the safety system may be discharged from the steam discharge section 25 of the emergency cooling water storage section 20. The operation of the safety system may remove residual heat generated in the reactor coolant system 111 and the core 114. In addition, safety injection water is supplied to the reactor coolant system 111 to reduce the pressure and temperature of the reactor coolant system 111, reduce the temperature of the core 114, and suppress a pressure increase inside the reactor containment (not shown) by the operation of the passive containment cooling system to protect the reactor containment. On the other hand, while the isolation valves 12, 13 provided in the main feedwater line 11 and the main steam line 14 are closed, the operation of the large turbine 15 is stopped. However, even when the reactor core 114 is stopped, residual heat is generated in the core 114 for a considerable period of time, and there is a lot of sensible heat in the reactor coolant system 111 and the reactor vessel 110, and thus the temperature of the reactor coolant system 111 and the reactor vessel 110 does not decrease rapidly. In other words, during a nuclear design basis accident, the nuclear power generation facility is stopped, but the reactor cooling and power generation system 100 continues to operate. Accordingly, emergency power supply and residual heat removal may be efficiently carried out. Accordingly, even when an accident occurs, the heat exchange section 120 and the power production section 130 may be operated in a substantially similar state as normal operation. Therefore, the power production section 130 may cool the reactor vessel 110 while continuously producing electric power. Over time, the temperature of the reactor vessel 110 may decrease as the residual heat generated in the core 114 decreases. In this case, the reactor cooling and power generation system 100 may be operated in a substantially similar manner as normal operation while reducing the power production amount generated by the power production section 130 in accordance with the reduction in the amount of heat transferred. FIG. 1D is a conceptual view illustrating a natural circulation operation of the reactor cooling and power generation system 100 during a nuclear design basis accident associated with an embodiment of the present disclosure. Referring to FIG. 1D, it is a conceptual view in which the operation of the small pump 153 is disabled due to a natural circulation operation during a design basis accident of the reactor cooling and power generation system 100. A safety system such as a passive residual heat removal system, a passive safety injection system, and a passive containment cooling system including the emergency cooling water storage section 20 installed in a plurality of trains by the related signals may be automatically operated as in the foregoing case of FIG. 10. Accordingly, the reactor coolant system 111 is cooled, and the residual heat of the core 114 is removed, and safety injection water is supplied to the reactor coolant system 111 to reduce the pressure and temperature of the reactor coolant system 111, reduce the temperature of the core 114, and suppress a pressure increase inside the reactor containment (not shown) to protect the reactor containment. On the other hand, while the isolation valves 12, 13 provided in the main feedwater line 11 and the main steam line 14 are closed, the operation of the large turbine 15 is stopped. In detail, when supply of feedwater from the small pump 153 is stopped for various reasons, the valve 157 and the check valve 158 connected to the condensate storage section 150 may be opened by the related signal or the operator's action, 159 to supply feedwater from the condensate storage section 150 through the pipe 159, and at this time, the feedwater may be supplied by natural circulation due to gravity. In other words, gravity acts on the condensate in the condensate storage section 150 to supply the condensate by natural circulation. Accordingly, the actuation state of the heat exchange section 120 and the electric power production section 130 may be operated in a state similar to that in the normal operation except for the small pump 153. When the residual heat of the core 114 gradually decreases to reduce a steam production amount over time, the heat exchange section 120 and the electric power production section 130 may be operated similarly to the normal operation while reducing a power production amount of the electric power production section 130. FIG. 1E is a conceptual view illustrating the operation of the reactor cooling and power generation system 100 during a severe accident of the nuclear power plant associated with an embodiment of the present disclosure. Referring to FIG. 1E, it is a conceptual view in which the operation of the reactor cooling and power generation system 100 is disabled due to a severe accident operation of the reactor cooling and power generation system 100. A safety system such as a passive residual heat removal system, a passive safety injection system, and a passive containment cooling system including the emergency cooling water storage section 20 installed in a plurality of trains by the related signals may be automatically operated as in the foregoing cases of FIGS. 10 and 1D. However, when the probability of occurrence is extremely low, but various safety systems and non-safety systems do not operate, it may occur an accident in which the core temperature rises and the fuel melts. For example, in order to block the discharge of radioactive materials to the outside of the reactor containment when a severe accident such as the occurrence of the core melt 114′ during a nuclear accident, the operation of the heat exchange section 120 and the power production section 130 may be stopped. Accordingly, the pipe 173 connected to the IRWST 170 may be opened by the related signal or the operator's action to receive refueling water from the IRWST 170. As a result, the refueling water may be used to cool the reactor coolant system 111 including a lower portion of the reactor vessel 110 and the reactor vessel 110 and the core melt. In addition, since a flow path through the IRWST 170 and the second discharge section 175 is already formed, the flow supply and discharge of coolant may be efficiently carried out by a simple operation such as opening and closing the valve or the like according to an operation action on the refueling water received from the IRWST 170. In detail, the second discharge portion 175 may discharge the nuclear refueling water received from the IRWST 170 into the reactor containment (not shown) as a fluid (gas/steam, a mixture of gas steam and liquid/hot water or hot water). Furthermore, when a severe accident such as damage to the reactor vessel or exposure of the reactor core 114 occurs during a nuclear accident, in addition to the occurrence of the core melt 114′ in the reactor, the operation of the heat exchange section 120 and the power production section 130 may be stopped to allow the opening of the valve 174 connected to the first discharge section 175 and the injection of feedwater through the IRWST 170. Besides, the pipe 173 connected to the IRWST 170 may be constructed to cool the reactor vessel 110 and the reactor coolant system even when cooling and power generation using the heat exchange section 120 and the electric power production section 130 are disabled due to a failure thereof. Depending on the characteristics of the power plant, a pump (not shown) may be installed in the pipe 173 connected to the IRWST 170 to forcibly inject feedwater or inject feedwater using gravity. Furthermore, according to another embodiment electric power production section 230 described below, the same or similar reference numerals are designated to the same or similar configurations to the foregoing example, and the description thereof will be substituted by the earlier description. FIG. 2A is a conceptual view of a reactor cooling and power generation system 200 associated with another embodiment of the present disclosure. Referring to FIG. 2A, the reactor cooling and power generation system 200 may further include an evaporator section 280, and the evaporator section 280 may be formed to exchange heat with an internal fluid of the heat exchange section 220 and the condensate of the condensate storage section 250. In detail, the reactor cooling and power generation system 200 may be formed to have a first circulation section to circulate the heat exchange section 220 and the evaporator section 280. Meanwhile, the reactor cooling and power generation system 200 may be formed to have a second circulation section formed to circulate the evaporator section 280, the electric power production section 230, and the condensate storage section 250. In other words, the reactor cooling and power generation system 200 may be formed to have a dual circulation loop of the first circulation section and the second circulation section. The evaporator section 280 may be formed to be a boundary between the first circulation section and the second circulation section. The first circulation section may be formed to circulate by a single-phase fluid. In detail, the single-phase fluid of the first circulation section may be a compressed gas (gas). The fluid circulating through the first circulation section exchanges heat with the evaporator section 280 through the discharge pipe 222 and the valve 281 connected to the heat exchange section 220. The fluid being heat-exchanged in the evaporator section 280 may be supplied to the heat exchange section 220 through a pipe 282, a compressor 284, a valve 285, a check valve 286 and a pipe 287. In detail, the compressor 284 or a blower (not shown) may be formed to perform the circulation of the single-phase fluid of the first circulation section. A motor 283 that operates the compressor 284 may receive electric power by a connected line 238′ branched from a connected line 238. On the other hand, the fluid circulating through the second circulation section may be supplied from a small pump 253 to the evaporator section 280 through a valve 254, a check valve 255 and a pipe 256, and a fluid converted into steam in the evaporator section 280 may be supplied to the electric power production section 230 through the discharge pipe 222′ and then cooled and condensed in the first heat exchange section 231 and stored in the condensate storage section 250, and circulated through the pipe 251. In other words, the fluid circulating through the second circulation section is supplied to the electric power production section 230 including the Stirling engine 240 while circulating the second circulation section described above, so that the thermal energy of the fluid operates the Stirling engine 240 to convert the thermal energy into mechanical energy, and the electricity generation section 244 connected to a shaft may convert the mechanical energy into electric energy to produce electric power. Moreover, electric power produced by the electric power production section 230 may be formed to use the electric power through a power system 260. Moreover, the heat exchange section 220 capable of cooing the outer wall of the reactor vessel 210 may be formed in a hemispherical shape as illustrated in the drawing, and the heat exchange section 220 may cool the outer wall of the reactor vessel 210 without a coating member or a heat transfer enhancement member. According to the present embodiment, there is a disadvantage in that the evaporator section 280 is added thereto compared to the embodiment of FIG. 1A, but it has an effect of physically separating a fluid circulating through the first circulation section and the second circulation section, and there is an advantage capable of cooling the reactor vessel 210 using a single-phase fluid. FIG. 2B is a conceptual view illustrating the operation of the reactor cooling and power generation system 200 during a normal operation of the nuclear power plant associated with another embodiment of the present disclosure. Referring to FIG. 2B, it is a conceptual view illustrating system arrangement and supercritical fluid flow during a normal operation of the nuclear power plant. Main feedwater (water) is supplied from the feedwater system 10 to the steam generator 213, and heat received from the core 214 by the reactor coolant circulation is transferred the secondary system through the steam generator 213 to increase the temperature of the main feedwater and produce steam. The steam produced from the steam generator 213 is supplied to the large turbine 15 along the main steam line 14 to operate the large turbine 150 and rotate the large generator (not shown) connected through the shaft to produce electric power. The power produced through the large generator may supply electricity to an on-site or off-site from the power system. On the other hand, the single-phase fluid inside the heat exchange section 220 having a shape capable of cooling the outer wall of the reactor vessel 210 is transferred to the evaporator section 280 by receiving the heat of the outer wall of the reactor vessel 210. The cooling and power generation system 200 may be formed to have a first circulation section in which the single-phase fluid transferred to the evaporator section 280 transfers heat to a fluid to be supplied to the electric power production section 230 including the Stirling engine 240, and then circulates through the heat exchange section 220 and the evaporator section 280. Moreover, the compressor 284 and the blower (not shown) connected to the motor 283 may be formed to efficiently perform the circulation of the single-phase fluid circulating through the first circulation section. On the other hand, the fluid supplied from the small pump 253 to the evaporator section 280 through the valve 254, the check valve 255 and the pipe 256 may be supplied to the electric power production section 230 including the Stirling engine 240 while circulating through the foregoing second circulation section, and the thermal energy of the fluid may be converted into mechanical energy by operating the Stirling engine 240, and the mechanical energy may be converted into electrical energy by the electricity generation section 244 connected to a shaft to produce electric power. Moreover, electric power produced by the electric power production section 230 may be formed to use the electric power through a power system 260. FIG. 2C is a conceptual view illustrating a forced circulation operation of the reactor cooling and power generation system 200 during a nuclear design basis accident associated with an embodiment of the present disclosure. Referring to FIG. 2C, it is illustrated a conceptual view of the operation of the reactor cooling and power generation system 200 when the operation of the small pump 253 and the electric power production section 230 is enabled during a nuclear design basis accident. Specifically, when an accident occurs in a nuclear power plant due to various causes, safety systems such as a passive residual heat removal system, a passive safety injection system and a passive containment cooling system, including the emergency cooling water storage section 20, which are installed in a plurality of trains, may operate automatically. Further, steam generated by the operation of the safety system may be discharged from the steam discharge section 25 of the emergency cooling water storage section 20. The operation of the safety system may remove residual heat generated in the reactor coolant system and the core 214. In addition, safety injection water is supplied to the reactor coolant system to reduce the pressure and temperature of the reactor coolant system, reduce the temperature of the core 214, and suppress a pressure increase inside the reactor containment (not shown) by the operation of the passive containment cooling system to protect the reactor containment. On the other hand, while the isolation valves 12, 13 provided in the main feedwater line 11 and the main steam line 14 are closed, the operation of the large turbine 15 is stopped. However, since the temperature of the reactor vessel 210 at an initial stage of the accident is similar, the heat exchange section 220 and the electric power production section 230 respectively connected to the evaporator section 280 may be operated in a substantially similar state as normal operation. When the temperature of the reactor vessel 210 decreases as residual heat generated in the core 214 decreases and the reactor vessel 210 is cooled by the safety system over time, the electric power production section 230 may be operated similar to normal operation while controlling the power production amount. FIG. 2D is a conceptual view illustrating a natural circulation operation of the reactor cooling and power generation system 200 during a nuclear design basis accident associated with an embodiment of the present disclosure. Referring to FIG. 2D, it is a conceptual view in which the operation of the small pump 253 is disabled due to a natural circulation operation during a design basis accident of the reactor cooling and power generation system 200. A safety system such as a passive residual heat removal system, a passive safety injection system, and a passive containment cooling system including the emergency cooling water storage section 20 installed in a plurality of trains by the related signals may be automatically operated as in the foregoing case of FIG. 2C. Accordingly, the reactor coolant system 211 is cooled, and the residual heat of the core 214 is removed, and safety injection water is supplied to the reactor coolant system 211 to reduce the pressure and temperature of the reactor coolant system 211, reduce the temperature of the core 214, and suppress a pressure increase inside the reactor containment (not shown) to protect the reactor containment. On the other hand, while the isolation valves 12, 13 provided in the main feedwater line 11 and the main steam line 14 are closed, the operation of the large turbine 15 is stopped. In detail, when supply of feedwater from the small pump 253 is stopped for various reasons, the valve 257 and the check valve 258 connected to the condensate storage section 250 may be opened by the related signal or the operator's action, 259 to supply feedwater from the condensate storage section 250 to the evaporator section 280 through the pipe 259, and at this time, the feedwater may be supplied by natural circulation due to gravity. In other words, gravity acts on the condensate in the condensate storage section 250 to supply the condensate by natural circulation. Accordingly, the actuation state of the heat exchange section 220 and the electric power production section 230 may be operated in a state similar to that in the normal operation except for the small pump 253. When the residual heat of the core 214 gradually decreases to reduce a steam production amount over time, the heat exchange section 120 and the electric power production section 230 may be operated similarly to the normal operation while controlling a power production amount of the electric power production section 130. FIG. 2E is a conceptual view illustrating the operation of the reactor cooling and power generation system 200 during a severe accident of the nuclear power plant associated with an embodiment of the present disclosure. Referring to FIG. 2E, it is illustrated a conceptual view of the operation of the reactor cooling and power generation system 200 when the operation of the reactor cooling and power generation system 200 is stopped during a severe nuclear accident. A safety system such as a passive residual heat removal system, a passive safety injection system, and a passive containment cooling system including the emergency cooling water storage section 20 installed in a plurality of trains by the related signals may be automatically operated as in the foregoing case of FIG. 2C. However, when the probability of occurrence is extremely low, but various safety systems and non-safety systems do not operate, it may occur an accident in which the core temperature rises and the fuel melts. For example, when a severe accident such as the occurrence of the core melt 214′ during a nuclear accident occurs during a nuclear accident, the operation of the heat exchange section 220 and the electric power production section 230 respectively connected to the evaporator section 280 may be stopped. As a result, the valve 271 and the check valve 272 connected to the IRWST 270 may be opened by the related signal or the operator's action to supply feedwater from the IRWST 270 through the pipe 273 so as to cool a lower portion of the reactor vessel 210, and the valve 274 provided in the second discharge section 275 is opened to discharge the generated steam. Depending on the characteristics of the power plant, a pump (not shown) may be installed in the pipe 273 connected to the IRWST 270 to forcibly inject feedwater or inject feedwater using gravity. Moreover, even when a severe accident such as damage to the reactor vessel or exposure of the reactor core 214 occurs during a nuclear accident, in addition to the occurrence of the core melt 214′ in the reactor, in case where the operation of the heat exchange section 220 and the electric power production section 230 respectively connected to the evaporator section 280 is stopped, the opening of the valve 274 connected to the first discharge section 275 and the injection of feedwater through the IRWST 270 may be allowed from a preventive point of view. Furthermore, according to still another embodiment electric power production section 230 described below, the same or similar reference numerals are designated to the same or similar configurations to the foregoing example, and the description thereof will be substituted by the earlier description. FIGS. 3A through 3E are conceptual views of a reactor cooling and power generation system associated with still another embodiment of the present disclosure. Referring to FIG. 3A, the shape of the heat exchange section 320a of the reactor cooling and power generation system 300a may be hemispherical. In addition, the heat exchange section 320a having a shape capable of cooling the outer wall of the reactor vessel may further include a coating member 321a for preventing corrosion or increasing heat transfer efficiency. In an embodiment, the surface of the coating member 321a may be reformed in various ways, and may also be processed in an uneven shape (cooling fin) to increase the heat transfer surface area. Moreover, the surface of the coating member 321a may further include a heat transfer member (not shown) that can be chemically treated to increase the surface area so as to improve heat transfer efficiency. In other words, the surface of the coating member 321a and the heat transfer member may be chemically treated to increase the surface area, thereby efficiently performing heat transfer. Referring to FIG. 3B, the shape of the heat exchange section 320b of the reactor cooling and power generation system 300b may be a mixture of a hemispherical shape and a cylindrical shape. Various shapes may be employed to increase a heat transfer area of the heat exchange section 320b. Furthermore, the heat exchange section 320b may also cool the outer wall of the reactor vessel 310 without a coating member or a heat transfer enhancement member. Referring to FIG. 3C, the reactor cooling and power generation system 300c may further include a core catcher 327 inside the heat exchange section 320b having a form capable of cooling the outer wall of the reactor vessel 310, and the core catcher 327 may be formed to receive and cool the melt when the reactor vessel 310 is damaged. In addition, the heat exchange section 320c may further include a coating member 321c for preventing corrosion or increasing heat transfer efficiency. Referring to FIG. 3D, in the reactor cooling and power generation system 300d, a shape of the heat exchange section 320d having a form capable of cooling the outer wall of the reactor vessel may be a shape in which the heat exchange section 320d encloses the entire reactor vessel 310 in a double vessel form. Various shapes may be employed to increase a heat transfer area of the heat exchange section 320d. In addition, the reactor cooling and power generation system 300d may further include an evaporator section 380 connected to the heat exchange section 320d in a similar manner to the heat exchange section 220 of FIG. 2A described above. The evaporator section 380 may be formed to exchange heat with the internal fluid of the heat exchange section 320d and the condensate of the condensate storage section 350. In other words, the reactor cooling and power generation system 300d may be formed to have a dual circulation loop of the first circulation section and the second circulation section. Referring to FIG. 3E, the reactor cooling and power generation system 300e may further include a gas-water separator 390 formed to mount a valve 391 connected to the discharge pipe 322 thereof, and the gas-water separator 390 may be formed to transfer only a gas in a fluid circulating into the heat exchange section 320e to the electric power production section 330 through the discharge pipe 322″. Moreover, a coolant return line 392 and a pump 394 may further be formed to return a liquid separated from the gas-water separator 390 to the condensate storage section 350. A motor 393 that operates the pump 394 may receive electric power by a connected line 338″ branched from a connected line 338. A liquid separated from the gas-water separator 390 may be returned to the condensate storage section 350 through the coolant return line 392, the pump 394, the check valve 395 and the valve 396. Specifically, it may be designed and operated at a pressure higher than the vaporization point of liquid to induce single-phase liquid circulation from the pump 394 to the gas-water separator 390, and the liquid and the gas may be separated by the gas-water separator 390. FIG. 4 is a conceptual view of a reactor cooling and power generation system 400 associated with still another embodiment of the present disclosure. Referring to FIG. 4, the reactor cooling and power generation system 400 may be formed to circulate the internal fluid of the heat exchange section 420 through the electric power production section 430 in a single-phase liquid state. When the fluid circulating through the reactor cooling and power generation system 400 is a single-phase liquid, a pressure of the circulation loop may be rapidly increased when the volume changes with temperature, and therefore, a pressure control section 4100 may be provided to absorb a change in volume of the single-phase liquid, and control the pressure. On the other hand, when the fluid circulating through the reactor cooling and power generation system 400 is a single-phase liquid (liquid-phase fluid), the heat transfer efficiency may increase as compared with a case where a high-pressure gas (gas-phase fluid) circulates. Furthermore, when the pressure control section 4100 is used to pressurize to a predetermined pressure, the requirement of a net positive suction head (NPSH) of the small pump 453 may be relaxed. In addition, when the fluid circulating through the reactor cooling and power generation system 400 is a single-phase liquid, the foregoing condensate storage section and pipes and valves associated with the condensate storage section may be removed to construct the reactor cooling and power generation system 400 in a simplified manner. In other words, the circulation of the fluid in the reactor cooling and power generation system 400 may be simplified, and as the pipe and the circulation loop are simplified, the application of safety grade or seismic design may be facilitated. FIGS. 5A through 5C are conceptual views of a reactor cooling and power generation system associated with yet still another embodiment of the present disclosure. Hereinafter, a reactor cooling and power generation system provided with a heat exchange section 520 inside a reactor vessel to operate even during a normal operation and during an accident of the nuclear reactor so as to generate electric power will be described in more detail with reference to FIGS. 5A through 5C. Referring to FIG. 5A, the heat exchange section 520 of the reactor cooling and power generation system 500a may be provided inside the reactor vessel 510 to receive heat from the reactor coolant system 511 inside the reactor vessel 510. In detail, the heat exchange section 520 may be formed to circulate a fluid capable of receiving heat from the reactor coolant system 511 to perform cooling in the reactor vessel 510. In other words, the heat exchange section 520 may perform cooling on reactor coolant inside the reactor vessel 510 during a normal operation of the nuclear power plant. In case of a nuclear accident, cooling on the reactor coolant and the core melt may be carried out. Referring to the layout of the detailed structures of the heat exchange section 520, the heat exchange section 520 may include an inlet header arranged with inlets into which the fluid is injected, and an outlet header arranged with outlets from which the fluid is discharged, and an inner flow path for exchanging heat with the fluid. Furthermore, a core catcher may be formed as layout of an additional structure of the heat exchange section 520 to receive and cool the melt of the core 514 during a severe accident. The detailed description of the heat exchange section 520 will be described later with reference to FIGS. 6A through 6C and 7A through 7C. In addition, the fluid in the heat exchange section 520 may pass through the discharge pipe 522, the valve 522″ and the discharge pipe 522′, and the discharge pipe 522′ may be branched to the pipe 524 passing through the valve 523. As a result, the fluid in the heat exchange section may be supplied to the electric power production section 530. The fluid supplied to the electric power production section 530 may perform power production in the Stirling engine 540 through heat exchange. Moreover the heat-exchanged and discharged fluid is condensed and transferred to the condensate storage section 550 along the pipe 539. The condensate generated as a condensed fluid collected in the condensate storage section 550 may be circulated through the heat exchange section 520 and the electric power production section 530. Moreover, the condensate storage section 550 may be connected to the heat exchange section 520, the pipes 556, 556′ and the valve 556″ to supply the condensate to the heat exchange section 520. Specifically, the condensate in the condensate storage section 550 may be supplied to the pipes 556, 556′ connected to the heat exchange section 520 through the valve 554 and the check valve 555 by the motor 552 and the small pump 553 connected to the pipe 551. Furthermore, the condensate may be supplied to the pipe 556′ connected to the heat exchange section 520 by gravity through the valve 557 and the check valve 558 connected to the pipe 559 of the condensate storage section 550. The discharge pipe 522′ may further include a first discharge section 526 connected to the valve 525, and the first discharge section 526 may be formed to discharge at least a part of the fluid excessively supplied to the electric power production section 530 or allow it to bypass the electric power production section 530. Moreover, the heat exchange section 520 may be connected to the IRWST 570 to supply refueling water through the pipe 573. Specifically, the IRWST 570 may be connected to the valve 571 and the check valve 572. As a result, a second discharge section 575 connected to the valve 574 may be provided to discharge the refueling water supplied from the IRWST 570 to the pipe 573 and the pipe 556′ through the second discharge section 575 during an accident. Specifically, the second discharge section 575 is configured to cool the inside of the reactor vessel 510 even when cooling and power generation using the reactor heat exchange section 520 and the power production section 530 is not carried out due to a failure thereof or the like during a severe accident on a pipe for discharging a fluid (gas/steam, a mixture of gas/steam and liquid/hot water or liquid/hot water) from the heat exchange section 520 into the reactor containment (not shown), and the like. Referring to FIG. 5B, the reactor cooling and power generation system 500b may be formed to circulate the internal fluid of the heat exchange section 520 through the electric power production section 530 in a single-phase liquid state. As a result, similarly to FIG. 4, a pressure control section 5100 may be provided to control the pressure of the single-phase liquid. In addition, when the fluid circulating through the reactor cooling and power generation system 500b is a single-phase liquid, the foregoing condensate storage section and pipes and valves associated with the condensate storage section may be removed to construct the reactor cooling and power generation system 500b in a simplified manner. Referring to FIG. 5C, the reactor cooling and power generation system 500c may additionally further include a heat exchange section 520′ having a shape capable of cooling the outer wall of the reactor vessel 510. The heat exchange section 520′ may be formed to enclose the reactor vessel 510 and receive heat discharged from the reactor vessel 510 so as to cool the outer wall of the reactor vessel 510. In detail, the shape of the heat exchange section 520′ may be hemispherical. However, the shape of the heat exchange section 520′ is not limited to a cylindrical shape, and at least a part of the shape of the heat exchange section 520′ may include a cylindrical shape, a hemispherical shape, and a double vessel shape or a mixed shape thereof. In addition, the heat exchange section 520c′ may further include a coating member (not shown) for preventing corrosion or increasing heat transfer efficiency. The surface of the coating member may be reformed in various ways, and may also be processed in an uneven shape (cooling fin) to increase the heat transfer surface area. Further, the surface of the coating member may further include a heat transfer member (not shown) that can be chemically treated to increase the surface area so as to improve heat transfer efficiency. Besides, the heat exchange section 520′ may be connected to the IRWST 570′ to supply refueling water through the pipe 573′. Specifically, the heat exchange section 520′ may be connected to the valve 571′ and the check valve 572′. Moreover, when a serious accident occurs, the heat exchange section 520′ may further include a discharge section 575′ connected to a valve 574′ to discharge refueling water supplied through the pipe 573′ from the IRWST 570′ through a discharge section 575′. FIGS. 6A through 6C are views for specifically explaining a heat exchange section 520 of FIGS. 5A through 5C. FIG. 6A is an enlarged view of the conceptual view of a heat exchange section 520. FIG. 6B is a side view of the heat exchange section 520. In addition, FIG. 6C is a top view of the heat exchange section 520. Referring to FIGS. 6A through 6C, the heat exchange section 520 may include an inlet header 5202, an outlet header 5203, an internal flow path 5204, and structures 5201, 5201′ for forming the internal flow path 5204, and may be formed to include a core catcher including a core melt flow path 520c capable of receiving and cooling the core melt during a severe accident. In detail, the heat exchange section 520 arranges the inlets 5202a, 5202b, 5202c, 5202d in the inlet header 5202 to inject a fluid (a fluid during a normal operation, IRWST refueling water during a severe accident) 5204) into the internal flow path 129. In addition, the inner flow path 5204 may be formed in a U-shape so as to surround the structure 5201′ so that fluid at low temperature surrounds the structure 5201′ and receives heat while rotating the structure 121′ to increase the temperature. Further, the fluid having the increased temperature while passing through the internal flow path 5204 may be discharged to the outlets 5203a, 5203b, 5203c, 5203d of the outlet header 5203. In detail, as illustrated in FIG. 6C, the heat exchange section 520 may be formed to allow the fluid to flow into the inlet 5202a and be discharged to the outlet 5203a through the flow path 5204a. In addition, the inlets 5202a through 5202d may be formed to correspond to the flow paths 5204a to 5204d and the outlets 5203a to 5203d, respectively. The core melt generated by the melting of the core during a severe accident may be cooled by the fluid (IRWST refueling water) while spreading radially from a central portion of the heat exchange section 520 to its edge along the core melt flow path 520c. FIGS. 7A through 7C are cross-sectional views taken along lines A-A′, B-B′ and C-C′, respectively, of the heat exchange section 520 in FIG. 6A. Specifically, FIG. 7A is a top cross-sectional view of the heat exchange section 520 cut along line A-A′ in FIG. 6A. Referring to FIG. 7A, a fluid having an increased temperature while passing through the flow paths 5204a to 5204d of the heat exchange section 520 cut along line A-A′ may be formed to be discharged to the outlets 5203a, 5203b, 5203c, 5203d. Furthermore, FIG. 7B is a middle cross-sectional view of the heat exchange section 520 cut along line B-B′ in FIG. 6A. Referring to FIG. 7B, a fluid (a fluid during a normal operation, refueling water during a severe accident) is formed to circulate upward from the bottom to the top while passing through the internal flow path 5204 of the heat exchange section 520 cut along B-B′, and the fluid is formed to receive heat while circulating upward so as to increase the temperature of the fluid. Moreover, FIG. 7C is a bottom cross-sectional view of the heat exchange section 520 cut along line C-C′ in FIG. 6A. Referring to FIG. 7C, the heat exchange section 520 may be formed such that the fluid having a low temperature flows into the inlets 5202a, 5202b, 5202c, 5202d of the heat exchange section 520 cut along the line C-C′, and passes through the flow paths 5204a through 5204d to be discharged to the outlets 5203a, 5203b, 5203c, 5203d at an upper portion of the heat exchange section 120. In addition, since the heat exchange section 520 can be configured in various similar forms, the present disclosure is not limited to a form of this embodiment. FIGS. 8 through 10 are conceptual views illustrating various embodiments of a Stirling engine applied to a reactor cooling and power generation system of the present disclosure. Referring to FIG. 8, the Stirling engine 840 may include a first heat exchange section 831′ capable of receiving heat from a heat exchanger (not shown) provided adjacent to the Stirling engine 840. The first heat exchanging part 831′ may be formed to transfer heat to the high temperature section 841. Meanwhile, a second heat exchange section 832′ for dissipating heat transferred from the Stirling engine 840 may be provided. Specifically, the second heat exchange section 832′ may be connected to a low temperature section 842 to discharge the heat of the working gas transferred from the low temperature section 842 to the outside. The high temperature section 841 and the low temperature section 842 may be connected to each other in a communicable manner by a first heat exchange section 831, a second heat exchange section 832, a regenerative heat exchange section 833, and a connection flow path. For example, one end portion of the connecting flow path is connected to the high temperature section 841, the other end portion of the connecting flow path is connected to the low temperature section 842, and thus the working gas flows from the high temperature section 841 to the low temperature section 842 or from the high temperature section 842 to the high temperature section 841 through the first heat exchange section 831, the second heat exchange section 832, the regenerative heat exchange section 833, and the connecting flow path. Moreover, the regenerative heat exchange section 845 may be further provided between the first heat exchange section 831 and the second heat exchange section 832 which reciprocate between the high temperature section 841 and the low temperature section 842. The regenerative heat exchange section 845 is a component for increasing the performance and efficiency of the Stirling engine 840. A power production section 843 of the Stirling engine 840 may include a reciprocator 843a, a piston 843b and a connecting rod 846 inside a cylinder 843c. The connecting rod 846 may include a plurality of connecting rods 846a, 846b. The power production section 843 produces reciprocating power while working gas between the high temperature section 841 and the low temperature section 842 passes through the processes of heating, expansion, cooling, and compression, and the reciprocating power may be transferred to the power transmission section 843d by the connecting rods 846a, 846b. The power transmission section 843d may be mechanically connected to the rotation sections 843d′, 843d″ to generate rotational power, and the rotational power may be converted into electric energy through the electricity generation section 844. On the other hand, referring to the layout of the power production section 843, the first heat exchange section 831′ and the second heat exchange section 832′, the cylinder 843c, the first heat exchange section 831′ and the second heat exchange section 832′ are arranged side by side. Furthermore, the first heat exchange section 831′ and the second heat exchange section 832′ are continuously connected to each other on the same plane, and the regenerative heat exchange section 845 may be disposed adjacent between the first heat exchange section 831′ and the second heat exchange section 832′. In addition, the Stirling engine 840 may include a battery 847 to provide power for an initial engine operation or store the produced electric power. According to still another embodiment electric power production section 230 described below, the same or similar reference numerals are designated to the same or similar configurations to the foregoing example, and the description thereof will be substituted by the earlier description. Referring to FIG. 9, a first heat exchange section 931′ connected to a high temperature section 941 may be disposed below the Stirling engine 940. Furthermore, the second heat exchange section 932′ of the Stirling engine 940 may be disposed to surround at least a part of a cylinder 943c so as to enclose at least a part of a low temperature section 942. As a result, the processes of heating, expansion, cooling and compression by working gas between the high temperature section 941 and the low temperature section 942 may be carried out by heat exchange through the first heat exchange section 931′ and the second heat exchange section 932′. On the other hand, a regenerative heat exchange section 945 may be disposed between the first heat exchange section 931′ and the second heat exchange section 932′, and may have a shape of surrounding at least a part of the cylinder 943c. Referring to FIG. 10, the first heat exchange section 1031′ of the Stirling engine 1040 may be disposed in parallel with the cylinder 1043c in connection with the high temperature section 1041 at a lower portion of the Stirling engine 1040. Furthermore, the second heat exchange section 1032′ of the Stirling engine 1040 may be disposed in parallel with the cylinder 1043c in connection with the low temperature section 1042. Moreover, the first heat exchange section 1031′ and the second heat exchange section 1032′ may be arranged in parallel to each other, and a bent pipe may be connected between the first heat exchange section 1031′ and the second heat exchange section 1032′, and a regenerative heat exchange section 1045 may be disposed on the bent pipe. Although the reactor cooling and power generation system and the Sterling engine according to various embodiments of the present disclosure have been described above, the present disclosure is not limited to the foregoing reactor cooling and power generation system and the Sterling engine, and may include a nuclear power plant having the same. In detail, the nuclear power plant of the present disclosure may include a reactor vessel, a heat exchange section formed to receive heat generated from a core inside the reactor vessel through a fluid, and an electric power production section including a Stirling engine formed to produce electrical energy using the energy of the fluid whose temperature has increased while receiving the heat of the reactor, and may be formed to circulate the fluid that has received heat from the core through the electric power production section, and formed to operate even during a normal operation and during an accident of the nuclear power plant to produce electric power. It is obvious to those skilled in the art that the present disclosure can be embodied in other specific forms without departing from the concept and essential characteristics thereof. In addition, the detailed description thereof should not be construed as restrictive in all aspects but considered as illustrative. The scope of the invention should be determined by reasonable interpretation of the appended claims and all changes that come within the equivalent scope of the invention are included in the scope of the invention. |
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042960746 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT The present invention provides a method for the treatment of an assembly comprising a cladding material and a core of uranium, thorium or mixtures thereof to separately recover the cladding material and the core. The method is particularly applicable to the treatment of a nuclear fuel element comprising a cladding material containing a metallic fuel such as fissile or fertile uranium, thorium and combinations thereof. The cladding material generally comprises a stainless steel which consists principally of iron alloyed with chromium and containing minor amounts of other metal additives. The present invention also is applicable to zirconium or zirconium alloy cladding materials. The zirconium alloy generally consists principally of zirconium and contains minor amounts of one or more alloying materials such as nickel, chromium, tin or iron. For convenience, the present invention will be described with respect to its particularly preferred application, namely, the treatment of a nuclear fuel assembly. Referring to the sole FIGURE, the fuel assembly first is introduced into a cladding piercing zone 10 where at least a portion of the cladding material is pierced, perforated, scored, sheared or the like to at least partially expose the thorium or uranium core. The precise mechanical means used to accomplish the exposure of the core is not particularly critical. However, it generally is preferred to expose at least a portion of the surface of the core at intervals of about 1/2 to 11/2 inches throughout the length of the fuel assembly. The perforated or scored fuel assembly is next introduced into a hydriding-dehydriding zone 12. It will be appreciated that both the piercing of the cladding and the hydriding-dehydriding could be accomplished in a single zone; however, in accordance with the particularly preferred embodiment set forth herein, each operation is performed in separate zones. In hydriding-dehydriding zone 12, the assembly is reacted with hydrogen at a hydrogen pressure of from about 0.5 to 2 atmospheres (360 to 1400 torr) and preferably at about one atmosphere (760 torr). Lower pressures substantially reduce the hydriding reaction rate and at higher pressures the reaction rate is not significantly increased. The temperature during the hydriding reaction is maintained within a range of from about 400.degree. C. to 650.degree. C. When the fuel is uranium, the temperature is optimally maintained within a range of about 450.degree. C. to 600.degree. C. and for thorium, a temperature range of from 500.degree. C. to 650.degree. C. is preferred. In accordance with the particularly preferred embodiment, when the core of the assembly comprises uranium, thorium or mixtures thereof, the temperature is cycled during the hydriding process such that the core is exposed to both an optimum hydriding temperature and an optimum hydrogen diffusion temperature. The time required to achieve substantially complete reaction will vary, of course, depending upon the size and shape of the metallic thorium or uranium core as well as the amount of surface area of the core exposed in the cladding and piercing operation. Generally, when the assembly is treated in accordance with the preferred conditions set forth herein, it is found that the reaction is substantially complete in a time of from about 15 to 90 minutes. Following the hydriding reaction, the temperature in zone 12 is increased to from about 700.degree. C. to 900.degree. C. to decompose the hydride to elemental metal and release the hydrogen which is withdrawn via a pump 14. When the fuel assembly is one which has been irradiated and contains gaseous fission products, the hydrogen withdrawn preferably is introduced into a volatile removal zone 16 where the gaseous stream is treated, for example by condensation, to remove a major portion of the volatile fission products. It is a particular advantage of the present invention that forming the hydride and then dehydriding the product so formed, releases substantially all of the volatile fission products. In addition, during the hydriding step, the hydride is formed as discrete particles of substantially increased volume. These particles of increased volume tend to split or rupture the cladding material and increase the size of the openings therethrough such that after the subsequent dehydriding, the core material is left in the form of small, friable discrete particles which are readily recoverable from the assembly by simple mechanical means such as sieving or mechanical agitation to separate the particulate core material from the larger substantially intact pieces of cladding material. Advantageously, the hydriding-dehydriding step is repeated at least once and preferably twice to ensure complete release of any volatile fission products present as well as to ensure that substantially all of the elemental core material has been exposed to and reacted with hydrogen at least once. The resulting particulate fuel material is readily processible to produce new fuel assemblies by enrichment, if necessary, and sintering or arc casting to reform pellets of the fuel. Thus, it is seen that the present invention provides a method for treating such assemblies without the necessity of complex and expensive gaseous or liquid phase processing. Further, in accordance with the present invention, any plutonium which may be present is never isolated but remains with the fuel in such a dilute form, however, as to substantially negate the possibility of it being used in the production of a nuclear weapon. The following example is set forth to more clearly illustrate a specific embodiment of the present invention as applied to the decladding of a nuclear fuel element and recovery and separation of the valuable constituents of the core from the undesirable radioactive gaseous fission products. EXAMPLE The purpose of this example is to demonstrate the method of the present invention to (1) declad the fuel, (2) cominute the fuel so it will fall free from the cladding, (3) release the volatile fission product, and (4) restore the fuel to its initial chemical form (i.e., metal). To determine the ability of the present method to release volatile fission products without the necessity of using radioactive materials, it was determined to monitor the radon evolved during the tests. Radon is a decay daughter of thorium, uranium and plutonium that is produced in situ within the fuel, just as xenon and krypton is produced during fission. Calculations indicate that the radon contained in one gram of metallic thorium that had decayed for a year since it was arc cast was sufficient to produce several hundred disintegrations per minute. Therefore, monitoring the radon radioactivity when thorium is pulverized in accordance with the present method would provide an excellent measure of the amount of volatile fission products which would be released during the treatment of the irradiated thorium fuel. A simulated fuel assembly was built which comprised 1/4.times.1/4.times.3-inch square strips of thorium which were rounded and cut into 1/2-inch lengths to simulate fuel pellets. The pellets were loaded into a 4-inch long.times.1/4-inch O.D. piece of stainless steel tubing which had a wall thickness of 0.112 inches. After the pellets were loaded, the tubing was crimped on each end and a 1/8-inch diameter hole was punched in the tubing at 1-inch intervals along one side. The simulated clad fuel assembly was then treated at various hydriding-dehydriding conditions in accordance with the present invention. After treatment, the assembly was removed from the reaction chamber and examined. The conditions and results are set forth in the following table. For the tests in the table, it was found that the thorium hydriding and dehydriding temperatures appear to be most rapid around 600.degree. C. and 900.degree. C., respectively. Based on the differential hydrogen pressure in the system, it appears that maximum hydrogen absorption reaction occurred around 600.degree. C. Hydrogen pressure in the closed system increased as a result of the initial heatup of thorium pellets to 350.degree. C. At 350.degree. C., the pressure leveled off and then decreased slowly with continued heating. The decrease in pressure became more pronounced in the 600.degree. C. range and continued to decrease (indicating continued hydrogen absorption and reaction) until a temperature of about 680.degree. C. was reached. Sharp pressure increases were observed when the temperature was increased to above 700.degree. C. Maximum pressure increase was obtained at about 900.degree. C. Thus, hydriding occurs between 350.degree. C. and 680.degree. C. for thorium and is most rapid around 680.degree. C. while dehydriding occurs above 700.degree. C. and is most rapid around 900.degree. C. Complete pulverization of the thorium metal by repeated hydriding and dehydriding (three cycles) was readily achieved. Comminution of the metal to less than 400-mesh without mechanical treatment was not achieved. However, the dehydrided metal is extremely friable and readily comminuted to a size of less than 400-mesh by ball milling, pressure screening or the like. TABLE __________________________________________________________________________ PROCESSING OF CLAD THORIUM PELLETS Radon Released* Pressure (Torr) Hydriding Dehydriding % of Accum Series Test No. H.sub.2 Total# Temp .degree.C. Time (hr) Temp .degree.C. Time (hr) Cts Total % __________________________________________________________________________ 1 a 380 760 520 0.25 850 0.33 45 1 -- b 380 760 600 1 800 1 20 Nil -- c 380 760 560 0.5 800 0.25 4 Nil 1 d 380 760 700 0.25 1000 0.25 Nil -- 1 e 380 760 -- -- -- -- 80 1 2 f 380 760 500 0.5 700 0.25 30 Nil 2 __________________________________________________________________________ 2 a 760 760 525-600 2 800 0.25 60 1 3 b 550 550 -- -- -- -- 170 2 5 c 550 550 500-660 3 700-800 1 2350 24 29 d 550 550 570 0.5 810 0.25 Nil -- 29 e 550 550 660 0.5 710 0.25 Nil -- 29 f 450 450 400-700 3 950 1.0 1500 15 44 g 500 500 -- -- 800 0.5 Nil -- 44 h 150 150 -- -- -- -- Nil -- 44 i 900 900 400-650 2 -- -- 750 7 51 j 400 400 -- -- -- -- -- -- -- Cooled to Room Temperature, Disassembled, Cladding __________________________________________________________________________ Inspected 3 a 900 900 550 2 -- -- 700 7 58 b 900 900 400-600 2 900 0.3 1600 16 74 c 900 900 500 1.5 800-900 1.0 900 9 83 d 900 900 650 0.25 870 0.25 Nil -- 83 e 1000 1000 500 0.5 800-900 1.00 400 4 87 f 640-1000 640-1000 500 0.25 900 0.25 1200 12 99 g 1000 1000 460-560 0.5 750 0.25 Nil -- 99 h 1000 1000 -- -- -- -- -- -- 99 Cooled to Room Temperature and Disassembled __________________________________________________________________________ *Total counting rate of radon if completely released = 9850. #Balance of gas was argon. Substantial amounts of radon were involved during the hydriding-dehydriding of the fuel. The radon counting rate in the hydrogen increased rapidly above 400.degree. C. to a maximum at a temperature of about 900.degree. C. The radon appeared to evolve during both the hydride and dehydride portion of the cycle. When the foregoing example is repeated, using uranium clad in stainless steel, zirconium or a zirconium alloy, or a mixture of uranium and thorium clad in such alloys, substantially the same results are obtained. Specifically, the uranium, thorium or mixture thereof is reduced to a fine friable particulate form and the cladding material is sufficiently ruptured by the hydride form, so that on subsequent dehydriding, the particles are readily removable from the cladding by mechanical means. It is readily apparent that the present invention provides an economical, safe and easy to operate method for the recovery and separation of uranium, thorium or mixtures thereof from a cladding material. While the foregoing example and description exemplify what are presently considered to be the preferred embodiments of the invention, it will be appreciated that many changes might be made in the embodiments described. The application of the method of the present invention to other elements clad or sheathed in various metals also will be readily apparent. Thus, the foregoing description is to be construed and interpreted as illustrative only and not in a limiting sense; reference being had to the claims for such latter purpose. |
051280680 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The invention is suitable for treating various types of particulate materials, and especially contaminated soil, but it can also be used to treat sludges, sediments, scrap yard dust and the like. These particulate materials can be contaminated with heavy metals, organics and radioactive species either alone or in combination. FIG. 1 illustrates a first embodiment of the invention. Initially, the excavated soil is processed to remove large rocks and debris. This step is not shown in FIG. 1. The soil is then processed in a mechanical size separator 10 such as for instance a rotating drum or vibrating screen device to sort and prewash the feed soil with a contaminant mobilizing solution provided through line 12. Large pieces of soil, for instance larger than 5 mm are washed with the contaminant mobilizing solution, rinsed with water supplied through line 14, checked for residual contaminants, and returned to the site as recovered soil. The contaminant mobilizing solution used to wash the soil will be dependent upon the contamination to be removed. For soluble contaminants, the solution will contain a leaching agent. Many suitable leaching agents are known and common leaching agents suitable for leaching soluble radioactive compounds include for example potassium carbonate, sodium carbonate, acetic acid, sodium hypochloride, and others. Leaching agents for the soluble contaminants typically found in contaminated soils and the like are well known. For dispersible contaminants, the contaminant mobilizing solution contains a suitable surfactant. Again, suitable surfactants for dispersing contaminants such as oil, grease, polychlorinated biphenyls, etc., are also known. The contaminant mobilizing solution may contain various combinations of leaching agents and surfactants, again, depending on the contaminants in the soil to be cleaned. The effluent of soil particles smaller than 5 mm and contaminant mobilizing solution discharged from the mechanical separator 10 through line 16 is then processed in a countercurrent flow size separator such as the mineral jig 18. In the jig 18, additional contaminant mobilizing solution supplied through line 13 flows upwardly countercurrent to the effluent. The fines are carried upwardly with the upward flow of contaminant mobilizing solution to form a slurry which is discharged through a line 20. These fines typically include heavy metal particles. The velocity of the upward flow of contaminant containing solution in the mineral jig 18 is set to separate fines of a desired size, for example fines smaller than 60 microns in diameter. The slurry discharged in the line 20 includes, in addition to the fines, contaminant mobilizing solution which contains leached and dispersed metals and organics. Heretofore, mineral jigs such as that disclosed in U.S. Pat. No. 4,783,253, have only been operated in a cocurrent flow mode. We operate the mineral jig 18 in a countercurrent flow mode. For such countercurrent flow operation, the jig can be operated with a stroke length of 1/2 to 3/4 inch, a pulse frequency of 300 to 400 per min, an upflow rate of contaminant mobilizing solution of 1 to 8 litters per min, an underflow rate of 1 to 3 liters per min, with one layer of balls 3/16 inch in diameter or greater to provide a soil under flow of 80 to 95 percent and soil over the top of 20 to 5 percent. The intermediate sized particles between 5 mm and 60 microns in diameter, which are discharged from the bottom of the mineral jig 18, are abraded in an attrition scrubber 22 which dislodges mineral slime or fines from them. The intermediate sized particles and the dislodged fines discharged from the attrition scrubber 22 through line 24 are rinsed in a second countercurrent flow size separator such as the second mineral jig 26 operated in the manner discussed above in connection with jig 18. The countercurrent flow in the second mineral jig 26 is wash water which flows upwardly at a velocity again selected to separate the dislodged fines, typically of 60 microns in diameter and smaller. The slurry of fines and wash water is discharged through line 28. The remaining intermediate sized particles discharged from the second mineral jig 26 are processed in a density separator such as a cross-current flow jig 30 to extract higher density heavy metal solid waste particles. The mineral jig 30, which is similar to the jigs 18 and 26 is operated in the cross-current flow mode with a stroke length of 1/8 to 3/16 inch, a pulse frequency of 100-400/min, a water upflow rate of 1 to 8 liters/min, one to three layers of balls less than 3/16 inch to provide soil over the top of 80 to 95 percent and a soil underflow of 20 to 5 percent. The cross-current flow carrying the intermediate sized soil particles is discharged through a line 32 into dewatering apparatus such as, for instance, a clarifier 34 or a hydroclone. Sludge from the clarifier 34 is pumped by a pump 36 onto a drying pad 38. The dried particles recovered from the drying pad are checked for cleanliness and returned to the site as additional cleaned soil. Water removed by the clarifier 34 is circulated by a pump 40 through a line 42 as the countercurrent wash water for the second mineral jig 26, and through line 44 as the cross-current flow for the density separator jig 30. The two waste slurry streams in the lines 20 and 28 from the first and second mineral jigs 18 and 26, respectively, are discharged into precipitation equipment 46 to which is added a precipitant to precipitate the dissolved metals. A sulfide or other suitable agent can be used to precipitate the dissolved metals present in a particular contaminated soil. These precipitates and fine soil particles will be highly contaminated with organics and heavy metals. A flocculant, such as for example Nalco 7182, an anionic polymer that does not interfere with trace metal absorption and co-precipitation, supplied by the Nalco Chemical Company, Naperville, Ill., is added to the precipitates and fines conveyed from the precipitation equipment 46 through a line 48 to dewatering apparatus 50 which may include for instance Bardles-Mozley concentrator 52 which separate micron size particles of high specific gravity. Simultaneously, fine particles are washed by the high shear, orbital shaking of the table. Fine soil solution which is washed from the table is passed through high intensity matrix magnetic separators which remove micron sized particles coated with weakly paramagnetic hydroxides containing inorganic contaminants. Solids from the remaining solution are then separated from the stream by either filtration or flocculation settling and pelletizing in apparatus 54. The organically contaminated fractions can be further treated biologically, chemically or thermally and returned to the site. Concentrated solids removed by the Bardles-Mozley concentrator 52 can be disposed of or sold as a concentrate. The filtrate is passed through the line 55 to an activated carbon bed 56 to remove all organics before being sent through line 58 for recycling. The recycled solution is discharged in the one of two contaminant containing solution makeup tanks 60 and 62 which is not currently being used to feed the process. The contaminated activated carbon in the bed 56 can be thermally or chemically treated or buried. The recycled contaminant mobilizing solution is analyzed and an active component such as caustic or emulsifier are made up on a batch basis in the off-line makeup tank 60 or 62. Contaminant mobilizing solution from the active one of the tanks 60 and 62 is pumped by the pump 64 or 66, respectively, through the line 12 to the mechanical size separator 10 and through the line 13 to the first mineral jig 18. FIG. 2 illustrates a modified embodiment of the invention in which the contaminated soil, after large pieces have been removed, is fed to a mechanical size separator in the form of the screw washer/classifier 68 where the soil is washed with the contaminant mobilizing solution supplied through a line 70, and where the larger particles are rinsed with a water based cleaning solution introduced through line 72 and discharged as clean large solids. The intermediate sized particles and fines are passed through a line 74 to a first attrition scrubber 76 where attached fines are dislodged from the intermediate sized particles. The abraded particles are then discharged into a countercurrent flow size separator in the form of a first mineral jig 78. The countercurrent flow in mineral jig 78 is provided by contaminant mobilizing solution supplied through the line 79. A slurry of fines and contaminant mobilizing solution containing dissolved and or dispersed contaminants is discharged from mineral jig 78 through the line 80. The intermediate sized particles are passed through a second attrition scrubber 82 where they are again abraded to dislodge additional attached fines, and a second countercurrent flow size separator in the form of a mineral jig 84 which uses an upward flow of wash water to separate the additional dislodged fines in a waste slurry which is discharged through line 86. The remaining intermediate sized particles are dewatered in a hydroclone 88 and then clarified in a tank 90. Sludge from the tank 90 is deposited through a line 92 on a drying bed 94 by a pump 93 to produce additional cleaned soil to be returned to the site. Water removed by the cyclone 88 is recycled as the wash water through line 96 to the second mineral jig 84. Makeup water is added as required through line 97. The two waste slurry streams in lines 80 and 86 are delivered through line 98 to dewatering apparatus which includes hydroclones 100. The cleaned fines from the hydroclones 100 are discharged through a line 102 into a precipitation reactor 104 to which a flocculant is added. Dewatered fines can be removed from the reactor 104 for disposal, or for further treatment. Overflow solution from the tank 104 and discharged from the cyclone 100 is recycled. Where the contaminants include radioactive compounds or heavy metals, the recycled solution can be passed through an ion exchange bed 106 to remove the soluble metals before being discharged into the contaminant mobilizing solution makeup tanks 108 and 110. Again, while makeup chemicals are being added to one makeup tank 108 or 110, contaminant mobilizing solution is being pumped by a pump 109 or 111 from the other tank to the screw washer/clarifier 68 and the first mineral jig 78. FIG. 3 illustrates yet another embodiment of the invention. This embodiment utilizes a screen/washer mechanical size separator 112 similar to that used in the first embodiment to wash the feed soils with contaminant mobilizing solution supplied through line 113 and to separate and rinse with water provided through line 115 the large particles such as those over 5 mm. The intermediate sized particles and fines are then carried through a line 114 to a first attrition scrubber 116 which dislodges attached fines from the intermediate sized particles. The fines including those dislodged in the attrition scrubber 116 are then separated from the intermediate sized particles in a countercurrent flow size separator such as the first mineral jig 118 where the countercurrent flow is contaminant mobilizing solution provided through the line 120. The waste slurry containing the fines and solubilized and dispersed contaminants is discharged through the line 122. The remaining particles are passed through a second attrition scrubber 124 and then through a line 126 to a second mineral jig 128 for size separation by the countercurrent flow of rinse water. The waste slurry containing the fines is discharged from the second mineral jig 128 through line 130. The intermediate sized particles discharged from the second mineral jig 128 are passed through a classifier or gravity separator such as a cross-current flow jig 132 to remove heavy metal particles for disposal. The remaining intermediate sized particles are dewatered such as in clarifier 134. Again, the sludge from the clarifier 134 is discharged by pump 136 onto a drying pad 138 to produce additional clean soil. Water removed in clarifier 134 is recirculated by the pump 140 through a line 142 to supply the countercurrent flow to the second mineral jig 128 and through a line 144 to the cross-current flow jig 132. As in accordance with the invention, the waste slurry stream in lines 122 and 130 is treated to remove the contaminants and recirculate the contaminant mobilizing solution. The particular treatment of this waste slurry depends on the type of contaminants extracted from the soil. In the embodiment shown in FIG. 3 dissolved metal contaminants are precipitated in reactor 146 and the resulting precipitants and fines are separated by dewatering which includes the addition of a flocculant. The dewatering apparatus 148 may comprise the apparatus used in the embodiments in FIGS. 1 and 2 or other dewatering apparatus. Organic contaminants are removed from the recycled contaminant mobilizing fluid in a carbon bed 150 while the soluble radioactive contaminants which were not removed by precipitation are extracted in an ion exchange bed 152. Again, the recycled contaminant mobilizing solution is returned to the one of two makeup tanks 154 and 156 which is not currently in use, and is pumped by a pump 158 or 160 from the active tank to the screen/washer 112 and the first mineral jig 118. Examples of soil cleanup using the various embodiments of the invention follow. The standards for these examples were the toxic chemical leaching procedures (TCLP) established for the particular site by the Environmental Protection Agency. For the first three examples, the results are illustrated in line graph form to show a continuium of the effect of the settings of the countercurrent flows in the mineral jigs which determines the size of fines removed, and consequently the percentage of the feed soil recovered. EXAMPLE 1 Industrial site soil contaminated with about 11,000 ppm of copper was treated in accordance with the embodiment of the invention set forth in the flow chart of FIG. 1. The contaminant mobilizing solution was a one percent by weight aqueous solution of acetic acid which was used in the initial wash phase in the screen/washer 10 and in the first mineral jig 18. Water recovered from the clarifier 34 was used as the rinse in the second mineral jig 26 and the cross-current density separator 30. The results of the tests are shown in FIG. 4. The untreated soil is represented by the trace 162, the results of soil washed only with water shown by the trace 164 and the results of the use of acetic acid as the contaminant mobilizing solution which dissolves the copper which is then carried off with the waste slurry from the mineral jigs 18 and 26 is shown by the trace 166. While the initial contamination was about 11,000 ppm of copper, it can seen that with the use of the invention, most of the copper was removed. The clean soil limit for this site was 250 ppm. It can be seen that by adjusting the countercurrent flow in the mineral jigs so that 80% of the initial soil was recovered that this clean soil limit was satisfied. Even at 90% recovery, the residual copper contamination was only 50 ppm above the clean soil limit. EXAMPLE 2 Soil contaminated with 69 ppm of radium was treated according to the embodiment of the invention shown in FIG. 2 using a 0.1 molar aqueous solution of potassium carbonate and a 0.1 molar solution of sodium carbonate as the contaminate mobilizing solution. The rinse water was the water recovered by the dewatering hydroclone 88. In FIG. 5, which illustrates the results of this example, the trace 168 represents the untreated soil, trace 170 represents soil washed only with water, and the trace 172 shows the results of the soil treated with the potassium carbonate and sodium carbonate chemical wash and rinsed with water. It can be seen from FIG. 5 that most of the contamination resides in the fine fraction so that even untreated soil from which only about 25% of the smaller particles are removed meets the clean soil limit of 42 ppm of uranium shown by the dotted line. With the invention, over 90% of the soil was recovered within the clean soil limit of 42 ppm of uranium. EXAMPLE 3 Soil contaminated with approximately 295 ppm of polychlorinated biphenyls was treated according to the embodiment of the invention illustrated in FIG. 2. The contaminant mobilizing solution in this example was a one percent by weight solution of NP90 a surfactant produced by Henkel Corporation together with a one percent by weight solution of Adsee 799, a surfactant supplied by Witco Corporation. The results of the test are shown in FIG. 6 where trace 174 is the untreated soil, trace 176 is soil washed only with water, and the cross hatched area 178 shows the results of soil washed with the surfactant solution. As can be seen, only soil treated in accordance with the invention met the clean soil limit of 25 ppm shown by the dashed line, and virtually all of the soil was recovered by this process. EXAMPLE 4 Sewer sediment having the following initial contaminant levels: ______________________________________ Uranium 140 to 200 ppm Mercury 900 to 1000 ppm PCBs 5 to 10 ppm ______________________________________ the remediation requirements were: ______________________________________ Uranium 50 ppm Mercury 12 ppm PCB 2 ppm Pass TCLP ______________________________________ The sewer sediment was treated by attrition scrubbing and initial fines separation using a sodium hypochloride solution (20 g/l), washing with water and density separation using the embodiment of the invention illustrated in FIG. 3. The results of the test are shown in the bar chart of FIG. 7. The uranium target of 50 ppm was easily met using the invention. The chemical limit of 12 ppm of mercury was not met. However, this limit was arbitrarily set on the assumption that the mercury contamination was in the form of elemental mercury. In fact, the mercury was in the form of an intermetalic amalgam of uranium and mercury which is highly insoluble. As a result, the mercury level achieved passed the TCLP. EXAMPLE 5 Oil land farm soil with the following initial contamination levels: ______________________________________ Uranium 120 ppm PCB 7 to 14 ppm Oil/Grease 3 to 6 wt. % ______________________________________ was treated according to the embodiment of the invention shown in FIG. 3. The remediation requirements were as follows: ______________________________________ Uranium 80 ppm PCB 2 ppm Pass TCLP Test ______________________________________ The contaminant mobilizing solution was a surfactant mixture of 0.1 wt. % APG - 325 available from Henkel Corporation and 0.1 wt. % ASO available from Witco Corporation. This surfactant mixture was mixed with a leaching solution containing sodium hypochloride (20 g/1) and sodium carbonate (21 g/l). The results of this example for virtually 100% recovered soil were: ______________________________________ Uranium 60 ppm PCB <2 ppm Passed TCLP ______________________________________ The uranium levels for untreated soil, water washed soil and soil treated in accordance with the invention are shown in FIG. 8. From the above, it can be seen that the invention provides a versatile method and apparatus for treating various types of particulate materials contaminated with various substances. 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 arrangements 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. |
abstract | A system includes a containment vessel configured to prohibit a release of a coolant, and a reactor vessel mounted inside the containment vessel. An outer surface of the reactor vessel is exposed to below atmospheric pressure, wherein substantially all gases are evacuated from within the containment vessel. |
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abstract | An observation instrument for use in a cavity in a hot cell includes a radiation sensor located between a dome and a shield and a travel mechanism for moving the sensor between a retracted position away from the dome and an observation position in which the sensor extends inside the dome and a method of maintaining the cell including extracting the shield, sensor, and travel mechanism from the cavity; where appropriate, replacing the sensor; inserting a replacement dome into the cavity and sliding it into the proximity of the dome that has remained in position in the cavity; inserting the shield, sensor and travel mechanism into the cavity and sliding them into contact with the replacement dome; and moving the dome that has remained in position until it is expelled into the cell, by pressing the replacement dome against it. |
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description | The present invention is directed to the production of radiochemicals. In particular, the invention is directed to the production of Actinium-225 using Radium-226 as a starting material. The invention generally involves irradiating Radium-226, to produce Radium-225, which then undergoes a beta decay to Actinium-225. Actinium-225 can be used to produce its daughter, Bismuth-213. The Actinium-225 product of the present invention may be produced in an amount of about 5 mCi Radium-225 per 100 mCi Actinium-225. It should be noted that the present invention involves the use of Radium-226, which is the naturally occurring isotope of Radium, having a half-life of 1600 years. Thus, when the term Radium-226 is used throughout the specification, it may be considered that Radium having a natural isotopic abundance is being used. Thus, reference to Radium-226 as a starting material is not meant to imply an isotopically pure form of Radium-226. A. Preparation of Radionuclides 1. Actinium-225 The invention generally involves the conversion of Radium-226 to Radium-225, using high-energy photons to drive the conversion. This reaction can be described as a photodisintegration reaction. Radium-225 decays to Actinium-225, which is then separated using a chemical separation process. a. Theory The reaction for the conversion of Radium-226 to Radium-225 is a photodisintegration reaction, where absorption of high-energy electromagnetic radiation in the form of gamma-ray photons causes a Radium-226 nucleus to eject a neutron, resulting in the formation of Radium-225. This reaction will be referred to herein as a xe2x80x9cgamma,nxe2x80x9d or xe2x80x9cxcex3,nxe2x80x9d reaction, where xe2x80x9cnxe2x80x9d refers to the neutron ejected. The high-energy photons are generated by bombarding a converting material with high-energy electrons. The converting material is a material that gives off high-energy photons upon bombardment, and it should be a material that is refractory to the electron bombardment. Examples of such materials include Tungsten, Tantalum, Platinum, and Copper. The high-energy electrons used to bombard the converting material need to be of sufficient energy so that the photons produced are of sufficient energy to drive the photodisintegration reaction. The energy necessary for the photodisintegration is an energy level that is at least equal to the threshold (minimum) energy level of the giant resonance region of the cross-section versus energy curve for the isotopic conversion reaction. (Giant resonances are the energy average of the compound nucleus resonances of the compound system. These resonances have widths on the order of 1 MeV, and can be derived from the Kapur-Peierls theory of the scattering of a single neutron by a potential.) This is the energy necessary to produce the desired reaction between a photon and the Radium-226. The intensity of high-energy photons generated by the converting material is proportional to the power density (PD) of the electron beam in the converting material. Power density is calculated according to the following equation: PD=Exi/V where E is the energy of the electron beam, i is the current of the electron beam, and V is the volume of convertor through which the electron beam passes. While the minimum energy is governed by the threshold energy level of the giant resonance region, the maximum energy is governed by the converting material. That is, the converting material is going limit the energy which can be put into the system. For example, the energy of the high-energy electrons should be balanced against the ability of the converting material to absorb the energy. The energy of the beam should be sufficient to generate photon energy in a range suitable to convert the isotope, yet not be so great that a large percentage of electron beam energy passes through the converting material. Similarly, if the converting material is too thick, photons will be degraded as they pass through the material. Thus, the preferred thickness of the converting material depends on the electron beam energy, the composition of the converting material, and the giant resonance region threshold energy of Radium-226. b. Preparation of a Solid Target In one embodiment of the invention, the Radium-226 is coated onto the converting material. Thus, the as the converting material is bombarded with high-energy electrons, high-energy photons are produced. The high-energy photons then impact the Radium-226 coating on the converting material. In the method of the present invention, the converting material serves the role of converting high-energy electrons into high-energy photons. Thus, any material which has this capacity may be used for this purpose, provided that Radium-226 may be coated onto it. Such materials are described as xe2x80x9cconvertorxe2x80x9d materials in U.S. Pat. No. 5,949,836, to Lidsky et al. The converting material may be any material that exhibits the desired converting properties, is relatively refractory to the process, and may be electroplated. The converting material may have an atomic number higher than about 30. Examples of converting materials include, but are not limited to, Copper, Tungsten, Platinum and Tantalum. The converting material may be a metal plate, which may be milled, lapped, sanded, washed with distilled water, and dried. The converting material generally will have a thickness of from about 0.5 mm to about 1.7 mm, or a thickness of from about 0.8 mm to about 1.2 mm, or about 1 mm. As noted above, the converting material is coated with the radioisotope, e.g., Radium-226, for the reaction. This coating may be performed by electroplating of the radioisotope onto the converting material. The radioisotope coated onto the coating material comprises Radium-226, which forms Radium-226 dioxide upon exposure to air. Electroplating of Radium-226 onto the converting material may be performed using a Platinum electrode, although other types of electrodes may be used. Thus, electroplating may be performed using a Platinum electrode in a Radium-226 solution, which may be prepared by dissolving Radium-226 in a basic alkali metal hydroxide solution. Examples of alkali metal hydroxides include, but are not limited to, sodium hydroxide and potassium hydroxide. In an alternative embodiment, an electroplating metal substrate (converting material), such as a Copper, Tungsten, or Tantalum plate, is placed into a nickel-plating solution and the metal substrate is electroplated with nickel. Nickel-plating may be performed using a Watts nickel bath procedure. Briefly, this technique involves operating at a temperature of about 30-60xc2x0 C., usually with air agitation, and at a pH of about 3.5 to 5.0. Current density is usually from about 2 to 7 A/dm2. The bath composition includes nickel chloride (40-60 g/l), nickel sulfate (240-300 g/l), and boric acid (25-40 g/l). Alternatively, the nickel plating may be performed using the method described by Yoda et al., in U.S. Pat. No. 5,985,124. The resulting nickel-plated substrate is then placed into a Radium-226 dioxide plating solution and electroplated with Radium-226. This procedure is described briefly as follows. Sufficient Radium-226 is dissolved in 8 molar NHO3 to form a 0.1 M Radium-226 solution. Cells for electroplating are constructed according to Krishnaswami and Sarin, (Krishnaswami, S., and M. M. Sarin (1976), Anal. Chim. Acta, 83, 143-156). A teflon stir bar is placed in the electroplating device. Limiting values of the power supply are set to 6 V and 0.8 A. The device is placed on a stir plate in a fume hood. Stirring is started, and the power supply is current limited to 0.8 A. When sufficient plating has occurred, the plating should be terminated by disconnecting the power and adding concentrated ammonia. The electroplated target should be rinsed with distilled water and dried before proceeding. Alternatively, the converting material may be placed into a plating solution containing both nickel and Radium-226. The nickel and Radium-226 are then electroplated onto the converting material. In another embodiment, Radium Bromide or Radium Oxide may be mixed in a varnish and painted onto the converting material plate, using processes developed for production of Radium watch dials. In still other embodiments, the Radium may be plated onto the converting material using a method described by Chan et al. in U.S. Pat. No. 6,103,295, xe2x80x9cMethod of affixing radioisotopes onto a surface of a device.xe2x80x9d Regardless of the process chosen, the Radium-226 may be coated onto the substrate until a concentration of at least about 80 mg/cm2 is achieved. In fact, the concentration of Radium-226 may range from about 80 mg/cm2 to about 160 mg/cm2. However, the concentration may be lower or higher, depending on other factors, including energy of the electron beam. The coating may be performed in a manner that leaves a portion of the converting material exposed for contact with the electron beam. This may be achieved by pouring a spot of molten plastic having a high melting point onto the plate, and allowing it to harden prior to immersing the plate in the bath. The electroplating will then take place around the plastic spot. The plastic spot can be later removed, leaving an uncoated portion. Regardless of how the Radium-226 target is prepared, the Radium-226-coated converting material is ready for irradiation in accordance with the invention. c. Preparation of a Liquid Target In another embodiment, Radium-226 in solution may be converted using the present invention. The Radium-226 solution may comprise Radium-226 chloride, and may be in a concentration of from about 0.5 to about 1.5 molar, or about 0.75 to about 1.25 molar, or about 1 molar. In this embodiment, the solution of Radium-226 may be contained or uncontained. For example, an uncontained solution of Radium-226 may be flowed over a converting material. Samples of the material flowing off the converting material may be regularly sampled, and the solution may be recycled until sufficient product is produced. A target solution of Radium-226 may also be used in a contained form. For example, a solution of Radium-226 may be placed in a quartz vial. The solution may be stirred or unstirred. A beam of electrons is then targeted at a converting material, and the photons produced thereby are targeted at the quartz vial of Radium-226 solution. In this manner, a photodisintegration reaction occurs. There are a number of advantages to the use of a liquid target. In particular, it is advantageous that the product is already in solution. That is, a separate step for isolating solid product from solid reactant is not necessary. In this embodiment, the product may easily be separated by chromatographic separation. Steps for such separation are detailed below. d. Electron Bombardment Irradiating of the target is performed using an electron beam, which may be provided by an electron accelerator, and in particular, a linear accelerator. For a converting material having thickness of about 1 mm, the electron beam that is used should have a current of from about 100 to about 1000 microampere. Alternatively, the current may range from about 250 to about 750 microampere, and may be about 500 microampere. The electron beam may be continuous energy, or may be pulsed. Usually, the energy of the electron beam is about 2 to 3 times the energy level of the peak of the giant resonance region of the targeted isotope. For example, for the (gamma,n) isotopic conversion of Radium-226 to Radium-225, a significant portion of the of high-energy photons will have energy levels falling within the giant resonance region for this reaction, specifically from about 10 MeV to about 25 MeV, or about 15 MeV. Thus, the electron energy that impacts the converting material is from about 20 MeV to about 25 MeV. The high-energy electron bombardment is performed for a period of time sufficient to obtain the desired quantities of product. Generally, that period is from about 10 days to about 30 days, or from about 18 days to about 23 days. In one embodiment, the bombardment period is about 20 days. However, the period is dependent on a number of factors, including the electron beam energy (higher energyxe2x80x94less time; less energyxe2x80x94more time), the converting material (greater production of photonsxe2x80x94less time; fewer photonsxe2x80x94more time), the thickness of the converting material (too thinxe2x80x94electrons pass throughxe2x80x94inefficient conversionxe2x80x94more time; too thickxe2x80x94photons not efficiently producedxe2x80x94more time), and the concentration of the coating material (less material to photodisintegratexe2x80x94less time; more materialxe2x80x94more time). Ideally, the period should not be excessively long. Therefore, the efficiency of the reaction should be maximized to shorten the reaction. As a general rule, which may be applied to the production of other radioisotopes, the reaction should proceed for a period of time that is about 3 times the half-life of the product, or to approximately 80-90% of the maximum production capacity. Generally, one consideration is that the conversion of high-energy electrons into high-energy photons creates a great deal of heat in the converting material. The heat generated can be so great as to limit the rate the reaction can be performed. Thus, one can optionally include a mechanism for cooling the target, i.e., the coated converting material, during the reaction. The cooling mechanism may rely on radiative, conductive, or convective dissipation of the heat, and the mechanism may allow dissipation in, around, or through the target. Thus, for example, the target may be formed with channels therein, to allow passage of a coolant through the target; it may be solid, with coolant surrounding; or the target may be porous, to allow the coolant to flow into the interstices of the target. Suitable fluid coolants include liquids, such as water, or liquid gallium, and gases, such as helium. Liquid targets may be cooled by freezing prior to bombardment, or may be cooled by having a cooling coil submersed in or adjacent to the liquid target. Alternatively, the liquid target may be circulated through a cooling apparatus, such as heat exchanger. In still other embodiments, a liquid target is cascaded over a cooled converting material, which is bombarded by electrons. e. Separation of Products from Reactants Radium-225 decays by emission of a beta particle to Actinium-225. When sufficient Actinium-225 is produced by decay, it may be separated from the other materials by chemical separation techniques. In one embodiment, the irradiated Radium-226 and Radium-225 are dissolved from the target plate by means of an alkali metal hydroxide solution, such as sodium hydroxide solution (5 M), containing equal volumes of 30-H2O2, plus sufficient de-ionized is H2O to cover the target. Following dissolution, the solution containing the dissolved materials is then transferred to a vessel containing aluminum powder and then optionally purged with air. The Actinium-225 may then be chemically isolated and separated from the target. For example, after the final volume is adjusted to specific needs, the Actinium-225 is passed through a fine glass filter. The precipitated Radium-225 is retained in the filter. In some embodiments, all of the Radium and Actinium bound to the converting material is dissolved at once. This leaves a solution of both Actinium and Radium which must be separated. A liquid target would also be in this form, i.e., with a combination of dissolved Actinium and Radium. Briefly, the Actinium/Radium separation process involves dissolving a (dried) sample containing Actinium-225 and Radium-225 in 0.03M HNO3. The dissolved sample is passed over an ion exchange column designed for separating radiochemicals, for example, an LN(copyright) resin column (Eichrom Industries, Inc., Darien, Ill.). Radium-225 and Radium-226 pass through with the effluent, and remaining Radium can be additionally washed from the column with 0.03M HNO3. Bound Actinium-225 is eluted from the column with 0.35M HNO3. Of course, although a column is described here, the method would be applicable to batch, or other methods as well. In another alternative, separation of Actinium-225 from Radium-226 and Radium-225 may be achieved by crystallization of Radium Nitrate, wherein the supernatant contains the soluble Actinium. For example, Actinium possesses the same 2s,1d outer electron structure as Lanthanum and Yttrium. It possesses a slightly larger ionic radius than Lanthanum; otherwise its chemistry is very similar. The basis for Actinium-225 separation from Radium is an anion separation in which the HNO3 concentration of the Radium feed is adjusted to 5 M and Radium is loaded onto an ion resin column. Trivalent Iron, Chromium, and all divalent and monovalent ions pass through. The Actinium-225 follows with a slight delay. The Actinium-225 is collected separately from the contaminants. The Radium-225 and Radium-226 is stripped from the column with 0.35 M HNO3 and is retained for reuse in target fabrication. Isolated Actinium-225 may then be purified by an oxalate precipitation followed by cation exchange. Briefly, the Actinium-225 is precipitated as an oxalate by the addition of an oxalic acid solution. Filtration is performed and the supernatant discarded. Oxalates are then destroyed by boiling concentrated HNO3 and HClO4, taking to fumes of HClO4. The Actinium-225 is then taken up in 2 M HCl and loaded on a cation exchange column. The column is washed with 1 bed volume of HCl. Any remaining divalent ions are eluted with 3 bed volumes of 3 M HNO3. Actinium is eluted with 5 bed volumes of 6 M HNO3. Separation of Radium-225 from Actinium-225 is described in U.S. Pat. No. 5,809,394, to Bray et al. 2. Bismuth-213 a. Theory As Bismuth-213 is considered a xe2x80x9cdaughterxe2x80x9d of Actinium-225, it also may be produced in accordance with the present invention. The radioactive decay chain in which Bismuth-213 is found is well known: Uranium-233 (txc2xd=1.62xc3x97105 yr)xe2x86x92Thorium-229 (txc2xd=7,300 yr)xe2x86x92Radium-225 (txc2xd=14.8 day)xe2x86x92Actinium-225 (txc2xd=10 day)xe2x86x92bismuth-213 (txc2xd=46 min). FIG. 4 shows the complete decay chain of Uranium-233 to Actinium-225 to Bismuth-213. b. Elution, Separation and Purification Bismuth-213 may be produced through the radioactive decay of Actinium, using Actinium as a xe2x80x9ccow.xe2x80x9d Bismuth-213 produced may be separated through the use of an organic anion exchange resin to adsorb Bismuth-213 from other materials present. The ability to extract bismuth as an anion as a function of HCl concentration is well known and is described in Kraus, K. A. and F. Nelson, 1955, Proceedings of the International Conference on the Peaceful Uses of Atomic Energy, Nuclear Chemistry and the Effect of Irradiation, Vol. VII, P/837, xe2x80x9cAdsorption of the elements from hydrochloric acid,xe2x80x9d held in Geneva, August 8-20, 1955. The distribution for the bismuth chloride complex anion in HCl increases with decreasing acid concentration. Other chelator interfering ions of interest, i.e., rare earths, Radium and Actinium, do not extract as chloride anions using anion exchange resin. Therefore, the use of the anion exchange resin allows Bismuth-213 to be effectively removed from these and other ions which do not extract as chloride anions using an anion exchange resin. The separation of Bismuth-213 from other materials is described, for example, in U.S. Pat. No. 5,749,042, to Bray et al.; in xe2x80x9cAn improved Generator for the Production of Bi-213,xe2x80x9d by Wu et al., American Chemical Society Meeting (1996); and in xe2x80x9cGenerator System Development of Ra-223, Bi-212, and Bi-214 Therapeutic Alpha-Emitting Radionuclides,xe2x80x9d by Ramirez et al., American Chemical Society Meeting (1996). B. Use of Radionuclides Produced According to the Invention Actinium-225 produced in accordance with this invention is produced in sufficient production yield and radiochemical and radionuclidic purity that it is especially suited to a number of uses. For example, it is especially suited for medical uses, including, but not limited to, radioimmunotherapy, radiation therapy and for the detection of metastatic disease, such as with an intraoperative probe for detection of occult cancers. Medical applications for the radionuclides of the present invention include their use in radiopharmaceuticals and/or radiochemicals, as those terms are known in the art. Non-medical uses include the use as a standard, or as a tracer. 1. Use Alone (in xe2x80x9cneatxe2x80x9d Form) In medical uses, the radionuclide may be used alone, or it may be linked to another material. Examples of applications in which the radionuclide is used alone include medical imaging, radiation synovectomy, etc. For example, Actinium-225, Bismuth-213, or mixtures thereof can be incorporated into a hydrogel. The alpha-emitting radioactive gel may be infused internally for treatment of sarcomas, carcinomas and prostate disease, or may be used for external treatment of Kaposi""s Sarcoma, or other diseases. Actinium-225, Bismuth-213, or mixtures thereof can also be combined with compounds that are not targeted at specific cells, such as styrenes, or styrene polymers, acrylic polymers, biodegradable, or bioerodable materials such as hydrogels, or other products that can be formed into a colloidal dispersion or particulate form and may then used for radiation synovectomy. By incorporating releasable therapeutic drugs in a radioactive polymer or gel, this invention also aims to provide for the optimization of post-procedure management to improve the efficacy and safety of patient treatment. a. Preparation of a Pharmaceutical Composition Generally, the preparation of the radionuclide pharmaceutical preparation will depend on the route of administration and the condition being treated. However, general guidelines are presented here. These guidelines are equally applicable for radionuclides complexed with targeting molecules, described below. Examples of pharmaceutical compositions include a radionuclide, or chelated radionuclide, or a chelated radionuclide attached to a targeting molecule, in some embodiments by a linker, or any other composition including a radionuclide of the present invention, along with a pharmaceutically acceptable carrier, diluent, excipient, or vehicle. Suitable pharmaceutically acceptable carriers, diluents, excipients, and vehicles include, but are not limited to, neutral buffered saline or saline. Additionally, the pharmaceutical composition may contain other constituents, including for example buffers, carbohydrates such as glucose, sucrose, or dextrose, preservatives, as well as other stabilizers or excipients. Methods for preparing such formulations are well known. A formulation may be in the form of a suspension, injectable solution or other suitable formulation. Physiologically acceptable suspending media, with or without adjuvants, may be used. The formulations of the present invention are in the solid or liquid form containing the active radionuclide, and optionally the chelator/linker/targeting agent. These formulations may be in kit form such that the two components (i.e. chelator, radionuclide, linker, and targeting agent) are mixed at the appropriate time prior to use. Whether premixed or as a kit, the formulations may include a pharmaceutically acceptable carrier. Other examples of kits include kits for incorporating Actinium-225, Bismuth-213, or mixtures thereof into a steroid group, an aryl group, a substituted aryl group, a vinyl group, an isothiocyanate, or an isocyanate group capable of coupling with antibodies. Similarly, Actinium-225, Bismuth-213, or mixtures thereof can be incorporated into an aromatic amine, an aromatic isocyanate, an aromatic carboxylic acid, an aromatic isothiocyanate, benzoic acid, a substituted benzoic acid group, or a vinylestradial group. Any person could make use of such a kit, including a researcher, a pharmacist, a doctor, or even the end user, the patient. For injectable compositions, the present invention may be either in suspension or solution form. In solution form the complex (or when desired the separate components) is dissolved in a physiologically acceptable carrier. Such carriers generally comprise a suitable solvent, preservatives such as benzyl alcohol, if needed, and buffers. Useful solvents include, for example, water, aqueous alcohols, glycols, and phosphonate or carbonate esters. Such aqueous solutions generally contain no more than 50 percent of the organic solvent by volume. Injectable suspensions are compositions of the present invention including a liquid suspending medium, with or without adjuvants, as a carrier. The suspending medium may be, for example, aqueous polyvinylpyrrolidone, inert oils such as vegetable oils or highly refined mineral oils, or aqueous carboxymethylcellulose. Suitable physiologically acceptable adjuvants, if necessary to keep the complex in suspension, may be chosen from among thickeners such as carboxymethylcellulose, polyvinylpyrrolidone, gelatin, or alginates. Many surfactants are also useful as suspending agents, for example, lecithin, alkylphenol, polyethylene oxide adducts, napthalenesulfonate, alkylbenzenesulfonates, and the polyoxyethylene sorbitan esters. Many substances which effect the hydrophobicity, density, and surface tension of the liquid suspension medium may be used in injectable suspensions in individual cases. For example, silicone antifoams, sorbitol, and sugars are all useful suspending agents. The radionuclide may be formulated into vehicles for topical administration, and such vehicles also include solutions, but may additionally include gels, lotions, creams, or salves. Where necessary, the radionuclide may be formulated into an oral dosage form, the types of which are too numerous to list. Essentially, there is no limit to the method of administration, as long as the radionuclide can be effectively delivered to the site of interest. Actinium-225, Bismuth-213, or mixtures thereof can be incorporated into a hydrogel. An alpha-emitting radioactive gel may be infused internally for treatment of sarcomas, carcinomas and prostate disease, or may be used for external treatment of Kaposi""s Sarcoma, or other diseases. Actinium-225, Bismuth-213, or mixtures thereof can also be combined with nonspecific compounds, such as styrenes, or styrene polymers, acrylic polymers, biodegradable, or bioerodable materials such as hydrogels, or other products that can be formed into a colloidal dispersion or particulate form and may then used for radiation synovectomy. By incorporating releasable therapeutic drugs in a radioactive polymer or gel, this invention also aims to provide for the optimization of post procedure management to improve the efficacy and safety of patient treatment. b. Administration An xe2x80x9ceffective amountxe2x80x9d of the formulation is used for therapy. The dose will vary depending on the disease being treated. Although in vitro diagnostics can be performed with the formulations of this invention, in vivo diagnostics are also contemplated using formulations of this invention. Although appropriate dosages may be determined by experimental trials, about 5xc3x971010 to 5xc3x971011 conjugate complexes/70 kg of adult weight may be administered assuming an approximate 1:1 ratio of targeting agent to the alpha-emitter. Nevertheless, the amount and frequency of administration will depend, of course, on many factors such as the condition of the patient, the nature and severity of the disease, as well as the condition being treated. In addition, it may be desirable to first mask pre-deliver the targeting agent, without radionuclide, in order to minimize non-specific binding, and damage to normal healthy tissues. 2. Use Linked to a Targeting Agent Generally, it may be desirable to attach the radionuclide to a different material in order to specifically target a part of a person""s or animal""s body. For example, in order to target the radionuclide to a cancer, the radionuclide may be linked to a material that specifically interacts with that cancer, and not with other parts of the body. Examples of applications in which the radionuclide may be attached to another material include treatment and diagnosis of all types of cancer, and many other diseases. For synthesis of labeled organic molecules, the Actinium-225 can be passed through a cation-exchange column to remove salts and trace metals prior to labeling. It is preferable for labeling of organic compounds, such as proteins, monoclonal antibodies, and natural products, that the radionuclide solutions be as chemically pure as possible. The targeting agent may be used solely to carry to the radionuclide to the site of interest, or may have pharmacological activity of its own. For example, Actinium-225 and Bismuth-213 produced in accordance with the invention may be used in the treatment of AML Leukemia. In this embodiment, Actinium-225 or Bismuth-213 are attached to an anti-angiogenesis agent for adjuvant therapy. Such agents include, but are not limited to, endostatin, angiostatin and combrestatin. Other specific examples of targeting agents, with and without their own pharmacological activity, are described below. a. Targeting Agent The radionuclides of the present invention may be carried to their destination by attaching them to a targeting agent. Targeting agents include those agents that have a specific affinity, for example, to a molecule, or to a subcellular structure such as a receptor. These targeting agents carry the radionuclide to the specific destination. Alternatively, the targeting agent could be administered first, followed by the radionuclide, thereby catching and holding the radionuclide. The targeting agent usually holds the radionuclide in place until the radionuclide decays. Thus, the interaction of the targeting agent with the target usually lasts longer than the half-life of the radionuclide. There are a number of examples of agents that may be used as targeting agents. Useful targeting molecules include, but are not limited to, proteins and enzymes generally, including monoclonal antibodies, prostate secretory proteins, as well as statins, taxol, tamoxifen, taxene, and estrogen receptor modifiers. The possibilities are limitless, and for the sake of brevity, details are provided for only a few. Antibodies that may be linked to radionuclides of the present invention include monoclonal and polyclonal antibodies. Monoclonal antibodies are immunoglobulins of well-defined chemical structure, in contrast to polyclonal antibodies, which are heterogeneous mixtures of immunoglobulins. A characteristic feature of monoclonal antibodies is reproducibility of function and specificity, and such antibodies can be and have been developed for a wide variety of target antigens, including tumor cells. Chimeric monoclonal antibodies and fragments have been prepared by recombinant techniques (Morrison, S. L., Hospital Practice (Office Edition), 65-80 (1989)). Methods for obtaining monoclonal antibodies or fragments have been extensively discussed and are well-known in the art. 20 Such methods are detailed in Monoclonal Antibodies (R. H. Kennett, T. J. McKearn and K. B. Bechtol eds. (1980); see also Koprowski et al. (U.S. Pat. No. 4,196,265). The selection of a monoclonal antibody for the practice of this invention will depend upon the end use for which the radionuclide conjugated to the antibody will be employed. Such selection is within the skill of the art. Specific examples include antibodies that are directed against a cancer. Antibodies raised against a known marker for a cancer may be used to target that cancer. Prostate specific antigen is one example of an antigen that may be targeted with antibodies raised to the antigen. In this manner the radionuclide is directed specifically to the targeted cancer, and the radionuclide is held at the site without non-specific distribution around the body. Other antigens that are known to be expressed by specific cancer cells may be targeted in this manner. This embodiment may also be used to target foreign invaders, such as fungi, bacteria, or even viruses. Antibodies specific to these pathogens are well known in the art. By linking the radionuclide to such an antibody, the foreign pathogen can be killed by the radionuclide attached to the antibody that it binds. Methods of producing antibodies are well known in the art. Such methods include, for example, harvesting antibodies from an individual afflicted with cancer, or infected with a foreign pathogen. After being isolated and purified, the antibody can be linked to the radionuclide and placed back into the host. Alternatively, antibodies can be raised against antigens in vitro, followed by isolation and purification, linking to a radionuclide, and introduction into a patient in need of treatment. Antibodies that have been xe2x80x9chumanizedxe2x80x9d may also be used as targeting molecules with radionuclides, in accordance with the present invention. Such antibodies are generally from an animal origin, but have been modified by replacing part of their structure with the equivalent structure from human antibodies. Antigen specificity is maintained, while immunogenicity to the antibody itself is decreased. Another use of the radionuclides of the present invention relies on a target already present in a body, i.e., receptors. As is well known in the art, animals have many different kinds of receptors, for which natural and synthetic ligands are known. The examples are too numerous to list, but include examples such as steroid receptors and opioid receptors. Both natural and synthetic ligands are known for receptors, and by linking a radionuclide to these ligands, the receptors may be specifically targeted. This is especially important for conditions in which these receptors need to be targeted in a disease state. In one embodiment, Bismuth-213, Actinium-225, or mixtures thereof can be attached to a PSP94 prostate secretory protein and its immunogenic peptides and targeted at prostate cancer. As another example, receptors for regulatory peptides have been identified in a number of different cancer cell types. Examples of such peptides include, but are not limited to, somatostatin, vasoactive intestinal peptide, and cholecystokinin. By linking a radionuclide to a regulatory peptide, the cancer cell may be preferentially targeted. Alternatively, radionuclides of the present invention may be conjugated to compounds recognized as growth factors. Like the other targeting molecules discussed above, the growth factor is chosen because it is capable of specifically binding to a defined population of cancer cells. Many growth factors known to one of ordinary skill in the art may be utilized within the present invention. Representative examples include platelet derived growth factors, transforming growth factor-beta, interleukins (ie., IL-1, IL-2, IL-3, IL-4, IL-5, IL-6, IL-7, IL-8, or IL-9), granulocyte-macrophage colony stimulating factor (GMCSF), erythropoietin, tumor necrosis factor, endothelial cell growth factor, platelet basic proteins, capillary endothelial cell growth factor, cartilage-derived growth factor, chondrosarcoma-derived growth factor, retina-derived growth factor, hepatoma derived growth factor, bombesin, and parathyroid hormone. Other growth factors include epidermal growth factor, transforming growth factor-alpha, fibroblast growth factors, insulin-like growth factor I and II, and nerve growth factor. Growth factors are generally selected for their capacity to specifically bind to a defined population of cancer cells which include, for example, preneoplastic cells, premetastatic cells, and tumor cells (both benign and malignant). As will be understood by one of ordinary skill in the art, a defined population of cancer cells may generally be differentiated from normal cells based upon the greater number of growth factor receptors on the cell surface. Alternatively, a radionuclide may be linked to a ligand for hormone receptors, to target cancer cells that express such hormone receptors. Ligands that are particularly suited for linking include hormones such as estrogens, or estrogen derivatives, androgens, and steroids. Cholesterol and diethylstilbestrol may be used in a similar manner. Other ligands that may be linked include drugs which are known to target such receptors. Tamoxifen and taxene are specific examples of ligands that may be used. Specific ligands of interest that are not believed to fall within the above-identified categories include taxol and thalidomide. b. Preparationxe2x80x94Attachment to a Targeting Agent Linking the radionuclide to the molecules of interest is fairly easily accomplished using techniques known in the art. Examples of such techniques are discussed in U.S. Pat. No. 5,364,613, to Sieving et al., and U.S. Pat. No. 5,958,374, to Meares et al. Because the radionuclide is generally in its molecular state, i.e., it is not covalently bonded to another molecule, it may be necessary to join the radionuclide in some other manner. Because the radionuclide is usually a charged metal, chelating is a good choice. Thus, the radionuclide may be chelated in a larger molecule. The chelator may be covalently bonded to another functional moiety, such as a targeting agent. Thus, for example, a growth factor may be covalently bonded to a chelator, which is used to chelate the Actinium-225 or Bismuth-213. The radionuclide is then carried with the growth factor to its specific site in the body. The targeting agent may be joined to the chelator in any manner, including through the use of a linker. Generally, a linker will be covalently bonded to the chelator on one xe2x80x9cendxe2x80x9d and the other xe2x80x9cendxe2x80x9d will have a moiety for reacting, and covalently bonding, with a targeting agent. The chelator and linker may therefore be viewed as one molecule, having a chelating moiety on one end, and a reactive moiety on the other. Thus, to summarize, the radionuclide-containing composition may include 1) a chelator, 2) a linker, and/or 3) a targeting agent. In some embodiments, the chelator will act serve the role of targeting agent, and a separate targeting agent and linker will be unnecessary. Alternatively, a chelator may be covalently bonded directly to a targeting molecule, eliminating the need for a separate linker. The terms xe2x80x9cchelator,xe2x80x9d xe2x80x9clinker,xe2x80x9d and xe2x80x9ctargeting agentxe2x80x9d are conceptual terms meant to simplify the understanding of the complex, and should not in any way be considered limiting. Thus, combinations in which the radionuclide is chelated within the targeting agent, combinations including multiple chelators, linkers, or targeting agents, or combinations lacking any chelator, linker, or targeting agent, are contemplated. All that is necessary is that a radionuclide of the present invention be included. The radionuclide may be attached to a targeting molecule by two general procedures. In the first, a chelator is attached to a targeting agent, generally by a linker. The resulting conjugate then chelates the radionuclide. Alternatively, a linker may be bonded to a chelator, which is then pre-chelated by combining it with the radionuclide. The radionuclide/chelator/linker is then bonded to the target molecule. A variety of diverse organic macrocyclic complexing agents may be used to sequester the alpha-emitting radionuclide including, but not limited to, the following groups: (1) spherands, (2) cryptaspherands, (3) cryptands, (4) hemispherands, (5) corrands (modified crown ethers), and (6) podands (acyclic hosts) (see Cram, Science 240:760-67 (1988). In general, these macrocyclic ring compounds are large, somewhat spherical organic compounds which resemble cage structures, and have the ability to hold a heavy radionuclide as a ligand holds a metal ion. The chelator should be selected such that it has both a high affinity and specificity for the alpha-emitting radionuclide as well as a low intrinsic mammalian toxicity. High specificity avoids displacement by other divalent cations (Mg+2 and Ca+2) that are prevalent in physiological fluids. Additionally, the compound should either contain a functional group, or have chemistry which is compatible with the introduction of an appropriate functional group, to allow attachment to the linker. The affinity of the chelator for the alpha-emitting radionuclide is defined by the system energetics as described by Cram (supra). More specifically, as inferred by X-ray crystallographic data of complexed and non-complexed crown ethers, it is believed that the solution conformations of non-complexed ethers lack well-defined cavities with the associated convergently aligned binding sites. During the process of complexation, the crown ether undergoes desolvation and reordering of structure, a process which requires energy. If the chelator presents a rigid prestructured and desolvated cavity to the ion (as is the case for spherands), the energy normally consumed by desolvation and reorganization is reflected in a larger binding constant for the ion. Based on this fundamental principle of reorganization, Cram lists the affinity of hosts for their most complimentary guests as: spherands greater than cryptaspherands greater than cryptands greater than hemispherands greater than corrands greater than podands. The difference in binding affinity between spherands and podands is dramatic, for example, the binding constant of a lithium chelating spherand was found to be 1012 higher than its corresponding open-chain podand (see Cram, supra). Thus, although many different chelators may be utilized within the context of the present invention, spherands which are designed and synthesized specifically to sequester Actinium-225 or Bismuth-213 are particularly preferred. Particularly preferred chelators include 18-crownxe2x88x926 or 21-crown-7 ethers, including for example modified crown ethers such as dicyclohexano-21-crown-7 (Case and McDowell, Radioact. Radiochem. 1:58 (1990); McDowell et al., Solvent Extr. Ion Exch. 7.377 (1989); for other crown ethers or macrocyclic polyethers, see Pedersen, Science 241:536-540 (1988); U.S. Pat. No. 4,943,375, Eia et al.; Heterocycles 32(4):711-722 (1991); Wai and Du, Anal. Chem. 62(21):2412-14 (1990); Tang and Wai, Analyst (London) 114(4):451-453 (1989)). Briefly, Ac2+ is bound by the etherate oxygen network comprising the interior cavity of the spherical crown-ether molecule. This binding is believed to be pH dependent: Ac2+ complexes with a combination of a proton and smaller Group IA ions for the binding site within the crown cavity. These crown ethers may additionally be modified with polarizable functional groups (similar to changes made with closo- and nido-carboamyl species used in boron-neutron capture therapy), resulting in compounds with greater solubility in aqueous media (see generally, Mizusawa et al., Inorg. Chem. 24:1911 (1985)). Such changes improve retention of biological specificity after conjugation, and improve the conjugate loading capability of the biological agent. These modifications may be accomplished in tandem with the synthesis of the above-noted crown ethers under appropriate conditions for mild conjugation to the biological delivery system. Additional crown ethers suitable for use within the present invention may be synthesized, or purchased from various sources including, among others, Aldrich Chemical Co. (Milwaukee, Wis.), Fluka Chemical Corp. (Ronkonkoma, N.Y.), and Nisso Research Chemicals, (Iwai Co. Ltd., Tokyo, Japan). Chelation of the alpha-emitting radionuclide may be achieved by mixing the chelator with a salt of the alpha-emitting radionuclide which has been dissolved in solvent. The particular solvent chosen depends of course on the solubility of the chelator and alpha-emitting radionuclide. For example, Cram and co-workers prepared the sodium complex of a spherand simply by adding excess salt dissolved in acetonitrile to a methylene chloride solution of the spherand (see Cram and Lein, J. Am. Chem. Soc. 107:3657-3668 (1985)). The ability of the crown ether to sequester or complex with the alpha-emitting radionuclide may be readily determined (see Cox et al., xe2x80x9cRates and Equilibria of Alkaline-Earth-Metal Complexes with Diaza Crown Ethers in Methanol,xe2x80x9d Inorg. Chem., 27:4018-4021 (1988); see also Mohite and Khopkar, xe2x80x9cSeparation of Barium From Alkaline Earths and Associated Elements by Extraction with Dibenzo-18-crown-6 From a Picrate Medium,xe2x80x9d Analytica Chimica Acta, 206:363-367 (1988)). Briefly, separation of the complexed and free radionuclide can be accomplished by partitioning between an organic solvent (such as chloroform) and water. The complexed radionuclide will partition into the organic phase, whereas the free radionuclide will reside exclusively in the aqueous phase. Alternatively, a variety of chromatographic techniques such as High Performance Liquid Chromatography (HPLC) or Reverse-Phase High Performance Liquid Chromatography (RP-HPLC) may be utilized to separate chelated radionuclide from the free cation. Once isolated, verification of the molecular architecture may be accomplished. Briefly, the mode of cation binding can take two forms: (1) through external association (ie., anion/cation pairing without bond formation), or (2) via coordination of the cation to the crown-ether oxygen network. Specificity and strong binding, which are preferred for the present applications, are dependent on the latter type of association. Single crystal X-ray diffraction techniques may be used to unambiguously assign the type of interaction for the solid materials, and 17O, 13C and 1H-NMR may be used to determine the structures of target materials in solution. Other chelators capable of chelating radionuclides include polyaza- and polyoxamacrocycles. Examples of polyazamacrocyclic moieties include, but are not limited to, those derived from compounds such at 1,4,7,10-tetraazacyclododecane-N,Nxe2x80x2,Nxe2x80x3,Nxe2x80x2xe2x80x3-tetraacetic acid (herein abbreviated as DOTA); 1,4,7,10-tetraazacyclotridecane-N,Nxe2x80x2,Nxe2x80x3,Nxe2x80x2xe2x80x3-tetraacetic acid (herein abbreviated as TRITA); 1,4,8,11-tetraazacyclotetradecane-N,Nxe2x80x2,Nxe2x80x3,Nxe2x80x2xe2x80x3-tetraacetic acid (herein abbreviated as TETA); and 1,5,9,13-tetraazacyclohexadecane-N,Nxe2x80x2,Nxe2x80x3,Nxe2x80x2xe2x80x3-tetraacetic acid (abbreviated herein abbreviated as HETA). Other chelators include linear or branched chelating moieties including, but are not limited to, those derived from compounds such as ethylenediaminetetraacetic acid (herein abbreviated as EDTA) and diethylenetriaminepentaacetic acid (herein abbreviated as DTPA). In other embodiments, a chelator may have a pharmaceutical application simply by its chelation of the radionuclide. For example, the chelated radionuclide my result in greater specific uptake by certain parts of the body than would be observed for the radionuclide delivered alone. Generally, however, the chelated radionuclide will be linked to a targeting agent. Linking the chelated radionuclide to the targeting agents is generally a matter of simple chemistry between reactive groups. The linker provides a covalent bridge between the chelator and the targeting agent. Ideally, the linker does not interfere with the ability of the chelator to sequester the radionuclide, or with the ability of the targeting agent to properly interact with its specific target. These goals are achieved in a variety of different ways. When the chelating moiety is macrocyclic, the linker may be attached to any annular atom. For example, when the chelating moiety is a polyazamacrocycle, the linker may be attached to an annular carbon atom or an annular nitrogen atom. When the linking moiety is attached to an annular nitrogen atom, the compound may be referred to as an N-substituted polyazamacrocycle. Chelating moieties having carboxylic acid groups, such as DOTA, TRITA, HETA, HEXA, EDTA, and DTPA, may be derivatized to convert one or more carboxylic acid groups to amide groups, and thereby provide a point of attachment to the chelator. The other end of the linker, i.e., the end for attachment to the targeting agent, includes a functional group that will facilitate that attachment. Functional groups capable of covalently binding to targeting molecules include, but are not limited to, those functional groups which can be activated by known methods, so as to be capable of covalently binding to targeting molecule(s). For example, the formation of active esters (xe2x80x94C(xe2x95x90O)OR, wherein R is, for example, succinimidyl) from carboxylic acids, the formation of acid halides (xe2x80x94C(xe2x95x90O)X, wherein X is typically Cl or Br) from carboxylic acids. The functional group(s) present on the linker which are capable of covalently binding to targeting agent may be chosen according to the targeting agent to which the chelating agent will ultimately be bound. Reactive pairs of functional groups permit conjugation of the chelating moiety with the targeting molecule, via the linker moiety, wherein one member of the pair is present on the chelating agent and the other member of the pair is present on the targeting molecule. For example, when the targeting molecule is a protein possessing a free amino (xe2x80x94NH2) group, a functional group such as isothiocyanate (xe2x80x94NCS) present on the linker permits reaction to form a joining linkage (in this case, a thiourea linkage), thereby forming a chelating agent-linker-targeting molecule complex. Other examples of appropriate reactive pairs of functional groups include, for example, xe2x80x94NH2 with xe2x80x94C(xe2x95x90O)OR (active ester) or with xe2x80x94C(xe2x95x90O)OC(xe2x95x90O)R (anhydride) or with xe2x80x94C(xe2x95x90O)X (acid halide) to yield an amide linkage; xe2x80x94NH2 with xe2x80x94NCO (isocyanate) to yield a urea linkage. Other reactive pairs involving xe2x80x94NH2 include xe2x80x94NH2 and xe2x80x94S(xe2x95x90O)2X (sulfonyl halide); xe2x80x94NH2 and xe2x80x94C(=NR)OR (imidate ester); and xe2x80x94NH2 and xe2x80x94OC(xe2x95x90O)X (haloformate). Examples of reactive pairs of functional groups include xe2x80x94SH and xe2x80x94C(xe2x95x90O)CH2X (haloacetyl) to yield a xe2x80x94SCH2 C(xe2x95x90O)xe2x80x94 linkage; xe2x80x94SH and -alkyl-X (alkyl halide) or xe2x80x94SH and xe2x80x94S(xe2x95x90O)O-alkyl (alkyl sulfonate) to yield a thioether; and xe2x80x94SH and xe2x80x94SH (sulfhydryl) to yield a xe2x80x94SSxe2x80x94 (disulfide) linkage. The purpose of the xe2x80x9clinkerxe2x80x9d is to attach the chelator to the targeting agent. If, however, the chelator includes a reactive functionality to which the targeting agent can attach, then a separate xe2x80x9clinkerxe2x80x9d molecule may be unnecessary. For example, if the chelator includes an isothiocyanate (xe2x80x94NCS), and the targeting agent includes an amino (xe2x80x94NH2), then the chelator can be attached directly to the targeting agent. Any such combination may be used, and the need for a separate linker molecule eliminated. However, the close proximity of the chelator to the targeting agent should not compromise the ability of either moiety to perform its role. For example, the chelator should still be able to effectively sequester the radionuclide, and the targeting agent should be able to interact with its biological target. If these purposes might be compromised, a longer linker molecule may be used. For example, in an embodiment of the present invention in which the targeting agent is a polymer of amino acids (e.g., peptide, polypeptide, protein, etc.), the alpha-emitting radionuclide is positioned within a chelator, which is in turn coupled by a linker to the amino (xe2x80x9cNxe2x80x9d) or carboxy (xe2x80x9cCxe2x80x9d) terminus of the targeting agent. The linker may act to place an inert xe2x80x9cspacerxe2x80x9d between the biologically active targeting agent and the alpha-emitting radionuclide containing complex. This space minimizes steric interactions that may interfere with the targeting agent""s affinity toward its target. The optimum length of the spacer arm is primarily dependent on the affinity of the targeting agent for its target. The higher this affinity, the smaller the relative importance of stearic repulsion between the chelator and the target receptors. A virtually limitless number of linkers may be selected which are suitable for use within the present invention, and this list includes disulfides, dicarboxylic acids, polycarbon chains, and modified polycarbon chains. Linkers may include hydrocarbon chains which range in length from 4 to 18 carbon atoms. Linkers may have six or more methylene units, such as hexamethylene diamine. The linker may be attached to any of a number of extraanular functionalities on the chelator, including carboxy and amino functionalities. Within one aspect of the invention, if the extraanular functionalization is a carboxy group, then a first synthetic step may involve reaction of the chelator with hexamethylene diamine. Subsequent reaction with the C-terminus of the targeting agent would complete synthesis of the conjugate. Alternatively, as noted above, the linker may be coupled to other aspects of the growth factor such as the N-terminus. Within this embodiment, after reaction with hexamethylene diamine the chelator may be reacted with succinic anhydride. Subsequent coupling of the linker to the targeting agent may then be accomplished through the N-terminus of the targeting agent. Alternatively, within another aspect of the present invention, the chelator may contain an amino functionality. In these cases, a dicarboxylic acid linker (for example, octanedioic acid) may be utilized to couple the chelator to the N-terminus of the targeting agent. On the other hand, if the chelator is reacted with ethylene diamine after condensation with the dicarboxylic acid, linkage to the targeting agent may be accomplished through the C-terminus. Specific examples of useful compounds include CHX DTPA-A and CHX DTPA-B. Methods for making these compounds are described in U.S. Pat. Nos. 5,286,850, 5,124,471, and 5,434,287. As used herein, DTPA CHX-A and DTPA CHX-B are used synonymously with CHX DTPA-A and CHX DTPA-B. Additional methods for attaching radionuclides to targeting molecules are found in WO 93/09816. Other methods are described in U.S. Pat. Nos. 4,923,985, 5,286,850, 5,124,471, 5,428,154 and 5,434,287 to Gansow et al. c. Preparation of a Pharmaceutical Composition The preparation of the radionuclide pharmaceutical preparation described above for xe2x80x9cneatxe2x80x9d compositions applies equally to compositions in which the radionuclide is used with a targeting agent. That information will not be repeated here. d. Administration An xe2x80x9ceffective amountxe2x80x9d of the formulation is used for therapy. The dose will vary depending on the disease being treated. Although in vitro diagnostics can be performed with the formulations of this invention, in vivo diagnostics are also contemplated using formulations of this invention. Although appropriate dosages may be determined by experimental trials, about 5xc3x97101 to 5xc3x971011 conjugate complexes/70 kg of adult weight may be administered assuming an approximate 1:1 ratio of targeting agent to the alpha-emitter. Nevertheless, the amount and frequency of administration will depend, of course, on many factors such as the condition of the patient, the nature and severity of the disease, as well as the condition being treated. In addition, it may be desirable to first mask pre-deliver the targeting agent, without radionuclide, in order to minimize non-specific binding, and damage to normal healthy tissues. 3. Use Linked to a Non-Targeting Agent In addition to linking the radionuclide to an agent which serves to target a specific part of the body, the radionuclide may be linked to another cell toxin, to increase the cell killing efficacy. For example, the radionuclide may be linked to an antineoplastic agent, increasing its efficacy. Antineoplastic agents work by the general mechanism that they are toxic to cells. However, these drugs are taken up to a greater extent by the more rapidly growing cancer cells. The antineoplastic effect can be made even more pronounced by linking the antineoplastic agent to a radionuclide. Such antineoplastic agents include, but are not limited to, vicristine, vinblastine, methotrexate, cisplatin, fluorouracil, oxyuridine, and adriamycin. 4. Other Routes of Delivery In addition to the routes of delivery described above, the compositions of the present invention may also be delivered from devices and/or implants. For example, the present compositions may be released from a battery-driven pump at a desired rate, for delivery to a site of interest. Alternatively, the present compositions may be formulated as extended-, prolonged-, or delayed-release formulations in polymeric vehicles. Such formulations may be prepared as pellets or implants, which are placed into a targeted site for delivery. Alternatively, such polymeric compositions of the present invention could be coated onto devices such as stents or catheters for delivery to sites of interest. Such embodiments are particularly advantageous when the disease or lesion to be treated involves unchecked vascular proliferation, such as in restenosis. Methods for making such polymeric formulations, and for making implants, and devices for drug delivery, are well known in the art, and are not restated here for purposes of brevity. The following examples are presented as an illustration of one embodiment of the present invention. These examples should not be construed as limiting the claimed invention in any way. A milled Tungsten plate having the dimensions of 3 mm (width)xc3x973 mm (height)xc3x971 mm (thickness) is obtained. The plate is well sanded, washed with distilled water, and dried thoroughly. A Nickel-plating solution is prepared by mixing nickel chloride (40-60 g/l), nickel sulfate (240-300 g/l), and boric acid (25-40 g/l). The pH is adjusted to approximately 3.5 to 5.0. The Tungsten plate, prepared as described above, is then placed into the Nickel-plating solution in an electroplating apparatus with a Platinum electrode and Nickel is electroplated onto the Tungsten plate. Operating conditions are: temperature of 30-60xc2x0 C., and current density of 2-7 A/dm2. Agitation is performed with air. The resulting nickel-plated substrate is then placed into a Radium-226 dioxide plating solution and electroplated with Radium-226. Briefly, sufficient Radium-226 is dissolved in 8 molar NHO3 to form a 0.1 M Radium-226 solution. Cells for electroplating are constructed according to Krishnaswami and Sarin, (Krishnaswami, S., and M. M. Sarin (1976), Anal. Chim. Acta, 83, 143-156). A teflon stir bar is placed in the electroplating device. Limiting values of the power supply are set to 6 V and 0.8 A. The device is placed on a stir plate in a fume hood. Stirring is started, and the power supply is current limited to 0.8 A. When sufficient plating has occurred, the plating should be terminated by disconnecting the power and adding concentrated ammonia. The electroplated target should be rinsed with distilled water and dried before proceeding. The Tungsten plate should be coated to a concentration of about 120 mg Radium-226/cm2. The target, as prepared above, is ready for bombardment with a high-energy electron beam. The target is placed in the path of an electron beam in a linear accelerator operating at 10 kW, and bombarded with high-energy electrons. The current of the electron beam is set for about 500 microampere. The energy of the electron beam impacting the target should be about 25 MeV. The target is bombarded for approximately 20 days, at a distance of 50 cm from the beam source. The theoretical production yield calculation results are given in FIG. 1, where the production activities of Radium-225 and Actinium-225 are given as a function of irradiation time for a 1.0 gram Radium-226 target and a 25MeV electron beam. The values shown in FIG. 1 were obtained using the results shown in Table I, FIG. 2, and FIG. 3. Table I and FIG. 2 present the gamma flux/spectrum produced by both 20 MeV and 25 MeV electrons. FIG. 3 gives the curve for the Radium-226 (gamma, n) cross-section as a function of energy. Higher specific activities can be achieved by moving the target closer to the converter, and higher total activities can be produced by using a thick wedge of material. Generally, electron disintegration cross sections are about 100 times smaller than photodisintegration cross sections. Since electrons can be converted into photons with greater than 50% efficiency at energies of 20 MeV or higher, it is desirable to work with the bremsstrahlung radiation. The bremsstrahlung dose rate in the forward direction is a function of electron energy when an optimum target is used. It should be noted that production rates in an electron accelerator do not increase much above 25 MeV as the xe2x80x9cgiant resonancexe2x80x9d peak for the target is near 15 MeV (See FIG. 3). The materials, including Radium-226, Radium-225, and Actinium-225, on the target, are dissolved from the Tungsten plate by use of a solution containing equal parts 5 M NaOH and 30% H2O2. After the materials are dissolved from the plate, the solution is neutralized by addition of sufficient HCl to bring the pH to about 7. The entire solution is dried, and re-dissolved in a solution of 0.03 M HNO3. The dissolved sample is passed over an LN(copyright) resin column (Eichrom Industries, Inc., Darien, Ill.) Radium-225 and Radium-226 will pass through with the effluent, and remaining Radium is washed from the column with 0.03M HNO3. Bound Actinium-225 is eluted from the column with 0.35M HNO3. The Actinium-225 in 0.35 M HNO3 is passed over a cation-exchange column to separate any unwanted salts, and to purify the radionuclide prior to complexation. a. Preparation of BOC-p-nitro Phenylalanine Transcyclohexyldiamine Monoamide Dissolve the BOC acid, N-hydroxysuccinamide, and EDC (48 mmol) in ethyl acetate (400 mL). The mixture is stirred for 12 hours. The reaction solution is filtered, and the filtrate is washed sequentially with saturated salt solution, lM HCl, 5% NaHCO3, and saturated salt solution (200 mL each). The organic layer is separated and dried over MgSO4. After filtering, the solution is rotary evaporated to a solid. The solid is taken up in DMF (200 mL) and added dropwise to trans-1,2-diaminocyclohexane over a period of 18 hours. The precipitated diamide is filtered off, and the solution is rotary evaporated to a thick oil. The residue is taken up in chloroform and washed, as above, to remove any of the starting materials. The chloroform solution is dried as before, filtered, and concentrated to a gel-like consistency. This material is poured onto a Buchner funnel and triturated with petroleum ether to leave the product as a light tan solid. b. Preparation of p-Nitrobenzyl-xe2x80x9cCHXxe2x80x9d Diethylenetriamine The BOC group is cleaved by stirring the amide (4.6 g) overnight in dioxane (300 mL) saturated with HCl. Addition of diethyl ether (200 mL), followed by cooling to 4xc2x0 C., adds significant precipitate. The dihydrochloride is collected on a Buchner funnel under argon and vacuum dried. The amide dihydrochloride is suspended in THF (50 mL) in a three neck round bottom flask held in an ice bath. The flask is in fitted with a condenser, thermometer, and a septum. Diborane/THF (6 equivalents) are injected into the flask, and the temperature is raised to 50xc2x0 C. and maintained there until the reduction is complete. The progress of the reaction is monitored by HPLC using a ten minute gradient of 100% 0.1M HOAc in water to 100% 0.1M HOAc in methanol. The column is a Waters DeltaPak C18. After the reaction is finished, the solution is cooled to room temperature, and methanol (50 mL) is added to decompose any excess hydride. The solution is taken to dryness on the rotary evaporator, and the residue is taken up in 100% ethanol (100 mL). This solution is taken to dryness using a high vacuum rotary evaporator. Dioxane (150 mL), previously saturated with HCl, is added to the solid and the suspension as refluxed for four hours. The final suspension is left at 4xc2x0 C. for 18 hours. The product is collected on a Buchner funnel under argon and then vacuum dried. c. Preparation of p-Nitrobenzyl CHX DTPA The triamine (1.0 g, 2.49 mmol) is dissolved in DMF (25 mL) 5 with sodium carbonate (1.992 g), and tert-butyl bromoacetate (2.915 g, 14.95 mmol) is added. The solution is heated to about 80xc2x0 C. overnight under argon after which the reaction mixture is poured into H2O (100 mL) and extracted with CH2Cl2 (100 mL). The organic layer is washed with water (3xc3x97100 mL), separated, dried over MgSO4, filtered, and rotary evaporated to an oil. The oil is further concentrated to a thick oil by high vacuum rotary evaporation. The oil is treated with TFA (25 mL) overnight. The excess reagent is removed by rotary evaporation. Preparative HPLC is performed to separate and collect the two major peaks. After completion of the pre-HPLC, the HPLC buffer is removed by ion-exchange chromatography (AG50 Wxc3x978 200/400 mesh H+ form). The two fractions are labeled as CHX-A or CHX-B. d. Preparation of p-Aminobenzyl CHX DTPA-A, -B Atmospheric hydrogenation of each fraction is performed using 100 mg of each nitro compound with 10% Pd/C (100 mg) at pH 8.5. The reaction is allowed to proceed until the H2 uptake halts. The reaction mixture is filtered on a fine frit with Celite 577. The filtrate is lyophilized to leave an off-white residue. e. Preparation of p-Isothiocyanatobenzyl CHX DTPA-A, -B Each fraction is dissolved in H2O (5 mL) and treated with thiophosgene (20 uL) in CHCl3 (10 mL) with maximum stirring under argon for two hours. The organic layer is removed by room temperature rotary evaporation, and the aqueous layer is lyophilized to leave an off-white solid. f. Final Complexation The reactive CHX DTPA-A (-B could be used as well) is to dissolved in phosphate buffered saline. Equal molar ratios of Actinium-225 are dissolved into the buffer solution. Monoclonal antibody raised against prostate serum antigen is then added in an equal molar ratio. The mixture is mixed for 4 hours at 4xc2x0 C., followed by anion exchange to remove any unbound Actinium-225. About 5xc3x971010 radionuclide complexes is dissolved in a one-milliliter volume of sterile saline solution. The solution is mixed into one liter of sterile lactated Ringers solution, which is then administered intravenously over one-half hour. a. Extraction Actinium-225 from Examples 4 and 5 above is placed in a 20-ml bottle and dried. This Actinium-225 is referred to as the xe2x80x9ccow.xe2x80x9d A 3M anion exchange disc is pretreated with 0.5M HCl by placing the acid in a syringe, locking or attaching the disc to the syringe, and by pushing down on the syringe plunger, forcing the acid through the membrane. The pre-wash acid is discarded. A volume of 10 ml of 0.5M HCl is drawn into a pipettor and ejected into into the xe2x80x9ccowxe2x80x9d storage bottle, allowing the Actinium-225 to dissolve in the solution. A pre-treated 3M filter is attached to the syringe outlet with an appropriate plastic micropipette tip attached to the outlet side of the 3M filter. Through the plastic tip, the dissolved xe2x80x9ccowxe2x80x9d containing the Actinium-225 and its daughters (including Bismuth-213) is pulled into the syringe up through the 3M anion exchange filter and up into the syringe barrel. The plastic tip is removed, as is the Bismuth-213-loaded 3M anion exchange disc. The Actinium-225-0.5M HCl solution is ejected from the syringe into the original bottle, to be reused. b. Washing The Bismuth-213 product has now absorbed onto the 3M anion exchange disc, as has minor traces of Actinium-225 and HCl (which adhere to the interstitial surfaces of the resin). A new syringe is attached to the Bismuth-213-loaded anion exchange disc and a 0.005M HCl wash solution is pulled up through the disc. The disc is then removed and the acid wash, containing traces of interstitial xe2x80x9ccowxe2x80x9d solution, is expelled into a waste bottle. The xe2x80x9cwastexe2x80x9d HCl is discarded. c. Bismuth-213 Elution A solution of 0.05M NaOAc (pH 5.5) is drawn into a new syringe. The washed Bismuth-213-loaded 3M disc is attached to the syringe, and the solution of 0.05M NaOAc (pH 5.5) is ejected through the loaded disk and into a collection bottle. The reactive CHX DTPA-A (prepared as described in Example 5, above) is dissolved in acetate buffer, pH 6.0. An equal molar ratio of Bismuth-213 is dissolved into the buffer solution. Monoclonal antibody raised against prostate serum antigen is then added in an equal molar ratio. The mixture is mixed for 4 hours at 4xc2x0 C., followed by cation exchange to remove any unbound Bismuth-213. About 5xc3x971010 radionuclide complexes is dissolved in a one-milliliter volume of sterile saline solution. The solution is mixed into one liter of sterile lactated Ringers solution, which is then administered intravenously over one-half hour. In summary, this invention is a reliable method for obtaining greater than 1000-millicurie quantities of Actinium-225/Bismuth-213 in less than 5-xcexcCi Radium-225/100 xcexcCi Actinium-225 radionuclide purity via bombardment of Radium-226. The Actinium-225/Bismuth-213 has physical properties that are useful for diagnostic and therapeutic radiopharmaceuticals, particularly when used for radioimmunotherapy. The entire contents of all documents cited in this specification is a part of the present disclosure, and all documents cited herein are hereby incorporated by reference. The foregoing detailed description has been given for illustration purposes only. A wide range of changes and modifications can be made to the preferred embodiment described above. It should therefore be understood that the following claims, including all equivalents, define the scope of the invention. |
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claims | 1. A core shroud comprising:a number of planar members;a number of unitary corners; anda number of subassemblies each comprising a combination of said planar members and said unitary corners,wherein each of said unitary corners comprises a unitary extrusion. 2. The core shroud of claim 1 wherein each of said unitary corners is substantially identical. 3. The core shroud of claim 1 wherein each unitary extrusion including includes a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. 4. The core shroud of claim 1 wherein at least one of said subassemblies comprises a plurality of said unitary corners; and wherein the first planar portion of one of said unitary corners is joined to the second planar portion of another one of said unitary corners, in order that said unitary corners are disposed side-by-side in an alternating opposing relationship. 5. The core shroud of claim 1 wherein said planar members include first planar members and second planar members; wherein each of said first planar members and said second planar members includes a first edge, a second edge disposed opposite and distal from the first edge, and a width measured by the distance between the first and second edges; and wherein the width of each of said second planar members is greater than the width of each of said first planar members. 6. The core shroud of claim 5 wherein at least one of said subassemblies comprises one of said unitary corners and one of said first planar members; and wherein the first planar portion of said unitary corner is joined to a corresponding one of the first and second edges of said first planar member. 7. The core shroud of claim 5 wherein at least one of said subassemblies comprises one of said unitary corners and one of said second planar members; and wherein the first planar portion of said unitary corner is joined to a corresponding one of the first and second edges of said second planar member. 8. The core shroud of claim 5 wherein a plurality of said subassemblies are combined to form a quarter perimeter segment of said core shroud; andwherein each quarter perimeter segment includes eleven of said unitary corners, two of said first planar members, and one of said second planar members. 9. The core shroud of claim 1, further comprising a number of flow deflectors; wherein each of said unitary corners includes a curved interior junction and a curved exterior junction; wherein each of said flow deflectors includes a curved portion and a number of substantially flat portions disposed opposite the curved portion; and wherein said curved portion is structured to mate with a corresponding one of the curved interior junction and the curved exterior junction. 10. A nuclear reactor comprising:a pressure vessel;an annular core barrel seated within and supported by the pressure vessel; anda core shroud supported within the core barrel, the core shroud comprising:a number of planar members,a number of unitary corners, anda number of subassemblies each comprising a combination of said planar members and said unitary corners,wherein each of said unitary corners comprises a unitary extrusion. 11. The nuclear reactor of claim 10 wherein each of said unitary corners of said core shroud is substantially identical. 12. The nuclear reactor of claim 10 wherein said planar members of said core shroud include first planar members and second planar members; wherein each of said first planar members and said second planar members includes a first edge, a second edge disposed opposite and distal from the first edge, and a width measured by the distance between the first and second edges; and wherein the width of each of said second planar members is greater than the width of each of said first planar members. 13. The nuclear reactor of claim 12 wherein a plurality of said subassemblies are combined to form a quarter perimeter segment of said core shroud; and wherein each quarter perimeter segment includes eleven of said unitary corners, two of said first planar members, and one of said second planar members. 14. The nuclear reactor of claim 10 wherein said core shroud further comprises a number of flow deflectors; wherein each of said unitary corners of said core shroud includes a curved interior junction and a curved exterior junction; wherein each of said flow deflectors includes a curved portion and a number of substantially flat portions disposed opposite the curved portion; and wherein said curved portion is structured to mate with a corresponding one of the curved interior junction and the curved exterior junction. 15. The nuclear reactor of claim 14, further comprising a number of grids disposed within said core shroud; and wherein said substantially flat portions of said flow deflectors are structured to engage and support a portion of a corresponding one of said grids. |
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abstract | A container body of a concrete cask includes a storage portion that stores a canister. A refilling system includes a reduction cylinder having an outside diameter smaller than the diameter of the storage portion and an inside diameter larger than the diameter of the canister, a first lift mechanism configured to raise and lower the reduction cylinder between a down position in the storage portion and an up position wherein it is drawn out of the storage portion, and a second lift mechanism having a holding portion for holding one end portion of the canister and configured to raise and lower the canister with respect to the container body. In refilling operations, the canister is loaded into and unloaded from the storage portion of the container body by the second lift mechanism with the reduction cylinder set in the storage portion. |
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claims | 1. An x-ray generation apparatus comprising an x-ray reflecting mirror formed on an inner surface of a concave aspheric surface, and an x-ray generation part for receiving at least one incident energy beam, wherein said x-ray generation part is arranged near a focal point including a focal point of a paraboloid, and said x-ray reflecting mirror has at least one aperture formed in a position except for a part of the concave aspheric surface crossing an axis including the focal point of the concave aspheric surface, and an incident energy beam irradiates said x-ray generation part through the aperture. 2. An x-ray generation apparatus as set forth in claim 1 , wherein claim 1 the energy beam is a laser ray. 3. An x-ray generation apparatus as set forth in claim 1 , wherein claim 1 said x-ray generation part has a target surface which forms an angle of not more than 45xc2x0 degrees with respect to the axis. 4. An x-ray generation apparatus as set forth in claim 3 , wherein claim 3 the incident energy beam is guided with respect to the target surface through the aperture in a direction outside a range of 30 degrees with respect to a normal to the target surface. 5. An x-ray generation apparatus as set forth in claim 1 , wherein claim 1 the aspheric surface of said x-ray reflecting mirror has a structure in which at least two kinds of a plurality of layers are layered in a plurality of times, and said x-ray reflecting mirror uses layers containing at least one of Rh, Ru and Mo and layers containing at least one of Si, B, Be, and Be compounds as constituent materials. 6. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the constituent materials of said x-ray reflecting mirror are materials for soft x-rays with a wavelength range from 69.5 xc3x85 to 140 xc3x85. 7. An x-ray generation apparatus as set forth in claim 1 , wherein claim 1 said x-ray generation part uses, as a target material for generating x-rays, a cryotarget in which a rare gas is kept at a low temperature. 8. An x-ray generation apparatus as set forth in claim 7 , wherein claim 7 the rare gas is in a liquid state. 9. An x-ray generation apparatus as set forth in claim 7 , wherein claim 7 the rare gas is in a solid state. 10. An x-ray generation apparatus as set forth in claim 7 , wherein claim 7 the rare gas is in a low temperature gas state having vapor density close to liquid density. 11. An x-ray generation apparatus as set forth in claim 1 , wherein claim 1 said x-ray reflecting mirror uses as an x-ray radiation part an x-ray optical system comprising reflecting curvature mirrors having 0.1 or higher steradian condensing solid angle around said x-ray generation part. 12. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the aspheric surface of said x-ray reflecting mirror is a multilayer film comprising one layer of Mo and another layer of one component selected from the group consisting of Si, B, Be, and Be compounds, and a thickness of the Mo layer falls within a range from 30% to 60% a repeated cycle length of the multilayer film. 13. An x-ray generation apparatus as set forth in claim 12 , wherein claim 12 one of the B, Be, and Be compounds in another layer is selected from the group consisting of BBe, B 2 Be, and B 6 Be. 14. An x-ray generation apparatus as set forth in claim 12 , wherein claim 12 a thickness of Si, B, Be, and Be compounds is 30% to 80% the cycle of the structure repeated in a plurality of times. 15. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 2 to 20 at. % of C, B, N, or O are added to at least one layer of the plurality of layers. 16. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the Be compound contains at least one of Ca, Co, Fe, Mo, Nb, Ti, V and W. 17. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the layers containing at least one of Rh and Ru are alloy layers of either Mo alloyed with Rh or Mo allayed with Ru each having the composition ratio of Rh or Ru to Mo within a range from 30% to 70% and the thickness of each alloy layer within a range from 30% to 70% a cycle of the repeated layers. 18. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the layers containing at least one of Rh and Ru are alloy layers of Rh and Ru having the composition ratio of Rh to Ru within a range from 30% to 70%, and a thickness of each allay layer within a range from 10% to 60% a cycle of the repeated layers. 19. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the aspheric surface of said x-ray reflecting minor is a multilayer film comprising one layer of Mo and another layer of one of B, Be and Be compounds, and a thickness of the Mo layer falls within a range from 30% to 60% a repeated cycle length of the multilayer film. 20. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the layers containing at least one of Rh and Ru are layers of Ru, and a thickness of the Ru layers falls within a range from 30% to 70% to a cycle. 21. An x-ray generation apparatus as set forth in claim 5 , wherein claim 5 the layers containing at least one of Rh and Ru are layers of Rh, and a thickness of the Rh layers falls within a range from 30% to 70% to a cycle. 22. An x-ray generation apparatus as set forth in claim 1 , wherein claim 1 the aspheric surface is elliptical and has an x-ray radiation part for taking out an x-ray from said x-ray generation part with a target surface to the outside, said x-ray radiation part having 0.1 steradian condensing solid angle around said x-ray generation part, and including an x-ray optical system in which a partial surface of an ellipsoid of revolution obtained by rotating an ellipse about a major axis thereof as a rotation axis passing through said x-ray generation part by a rotation angle of less than 180 degrees is defined as said x-ray reflecting mirror, said x-ray generation part is located near a focal point including one of focal points of the ellipsoid of revolution, a normal of the target surface includes the rotation axis and is included on a plane located at a position corresponding to a half of the rotation angle, said x-ray reflecting mirror has a multilayered structure in which at least two kinds of a plurality of layers are stacked in a plurality of times, and constituent materials of said x-ray reflecting mirror include a layer containing at least one of Rh, Ru, and Mo and a layer containing at least one of Si, B, Be, and Be compounds. 23. An x-ray generation apparatus as set forth in claim 1 , wherein claim 1 the aspheric surface comprises a paraboloid, and said apparatus further comprises an x-ray radiation part for taking out an x-ray from said x-ray generation part with a target surface to the outside, said x-ray radiation part having 0.1 steradian condensing solid angle around said x-ray generation part and including an x-ray optical system in which a partial surface of a paraboloid of revolution obtained by rotating the paraboloid of revolution about a major axis thereof as a rotation axis passing through said x-ray generation part by a rotation angle of less than 180 degrees is defined as said x-ray reflecting mirror, wherein said x-ray generation part is located near a focal point including one of focal points of the paraboloid of revolution, a normal of the target surface includes the rotation axis and is included on a plane located at a position corresponding to a half of the rotation angle, said x-ray reflecting mirror has a multilayered structure in which at least two kinds of a plurality of layers are stacked in a plurality of times, and constituent materials of said x-ray reflecting mirror include a layer containing at least one of Rh, Ru, and Mo and a layer containing at least one of Si, B, Be, and Be compounds. |
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051270285 | summary | BACKGROUND OF THE INVENTION Efficient x-ray diffractors having a large collection solid angle and high spectral resolving power have surfaces and planes that are cylindrically curved to different radii as in the Johansson geometry, or are doubly curved as in the spherically bent plate or in diffractors with spherical planes and toroidal surface. However, the ideal geometries involving doubly curved diffracting planes and surfaces that are curved differently than the planes cannot always be used for diffractors fabricated from bulk single crystals for several reasons. First, natural crystal are not available with sufficiently large d spacings to diffract long wave length x rays. Second, some natural crystals with desirable d spacings and high diffraction efficiency cannot be formed to the desired configuration because of difficulty in grinding or polishing the surfaces that are not parallel to cleavage planes or because of the tendency to fracture when elastic bending is used or because of the undesirable distortion of the crystal lattice if plastic bending is used. The latter problems are enhanced when using thick pieces of single crystals to form large area diffractors because of the greater stresses produced in bending. On the other hand, layered structures obtained by multiple deposition processes have not yet been made with sufficiently small d spacings to diffract x-rays of short wave length at large Bragg angles and have not been made sufficiently thick to be configurable to the desired geometry. The purpose of the present invention is to provide a diffractor geometry with a stepped surface that will increase the possible ways for fabricating large area, high efficiency diffractors and will increase the variety and types of diffracting materials that can be used in the fabrication of these diffractors. SUMMARY OF THE INVENTION This invention is a diffractor with surface steps configured to approximate the case wherein the diffractor has doubly curved diffracting planes and a smooth doubly curved surface on the side on which the radiation impinges. Such doubly curved diffractors have been described by Wittry and Sun in the Journal of Applied Physics, Feb. 15, 1990, pages 1633-1638, and by Wittry in U.S. Pat. Nos. 4,599,741 (Jul. 8, 1986), 4,807,268 (Feb. 21, 1989), and 4,882,780 (Nov. 21, 1989). The stepped diffractor is configured so that the surface of the steps have the same shape as the diffracting planes of the continuous case. Thus, the diffracting material has diffracting planes parallel to the surface of the steps. This makes it possible to fabricate doubly curved diffractors of high radiation collection efficiency using diffracting materials that cannot normally be employed in the fabrication of high efficiency diffractors having a continuous and smooth surface. For the diffractor with doubly curved surface steps, two factors govern the width of the steps for a useful diffractor, namely, a) the effect of the width of the steps on satisfying Bragg's law within a certain range of the desired value of the diffracting angle .theta..sub.B, and b) the effect of the finite width of the steps on the focussing properties of the diffractor. As will be shown in the section on embodiments of the invention, relatively simple equations can be derived that show these relationships and indicate how the step width should vary over the surface of the diffractor in order to satisfy the constraints imposed by both the rocking curve width of the diffracting material and the desired focussing accuracy. Although stepped diffractors have been previously proposed as approximations to the Johannson geometry by Okano in U.S. Pat. No. 3,469,098 (Sep. 23, 1969) and by Wittry in U.S. Pat. No. 3,927,319 (Dec. 16, 1975), neither of these inventions took into consideration the true effects resulting from the three dimensional nature of the problem of diffracting as much radiation as possible emanating from a point source. With the results of calculations of the type made by Wittry and Sun in their 1990 Journal of Applied Physics paper it is possible to see the advantages of diffractors with doubly curved planes and doubly curved surfaces. The unique problems that occur in the fabrication of these latter types of diffractors, which were discovered only by considerable experimentation, has led to the present invention in which the steps on the diffractor have a doubly curved surface. |
description | The present invention concerns a zirconium-based alloy, suitable for use in a corrosive environment where it is subjected to increased radiation and comprising 0.5-1.6 percentage by weight Nb and 0.3-0.6 percentage by weight Fe. The invention also concerns a component in a nuclear energy plant, which comprises an alloy of the mentioned kind. According to the prior art it is known to provide, in a nuclear energy plant, a component which comprises a zirconium-based alloy of the above-mentioned kind. Such an alloy has the advantage of fulfilling the requirements which are demanded on mechanical as well as corrosion properties of a material which in a corrosive environment is subjected to an increased radiation, in particular neutron radiation of the fast neutron kind. Thanks to its relatively high Fe-content it is possible through a suitable heat treatment, comprising annealing and quenching, to obtain secondary phase particles consisting of Zr, Fe and Nb in a matrix of a-phase of the zirconium-based alloy. By a suitable choice of the heat treatment variables time and temperature it is furthermore, with given contents of the included alloying materials Nb and Fe, possible to control the size of and the distribution of the secondary phase particles. The secondary phase particles may have a positive effect on the corrosion resistance of the alloy. It is therefore important to optimize the distribution of and the size of the existing secondary phase particles. It is thereby highly important to find a suitable composition of the alloying elements included in the alloy. The document U.S. Pat. No. 5,560,790 describes a zirconium-based alloy which comprises 0.5-1.5 percentage by weight Nb, 0.9-1.5 percentage by weight Sn and 0.3-0.6 percentage by weight Fe. Furthermore, this alloy comprises 0.005-0.2 percentage by weight Cr, 0.005-0.04 percentage by weight C, 0.05-0.15 percentage by weight O, 0.005-0.15 percentage by weight Si and the rest Zr. Thereby a microstructure is achieved in the material which includes particles of the kind Zr(Nb,Fe)2, Zr(Nb,Cr,Fe) and (Zr,Nb)3Fe. These secondary phase particles give the material good corrosion properties and good mechanical properties. Thanks to the high Fe-content, precipations of β-Nb-phase are avoided, which would have a negative influence on the resistance of the material against local corrosion attacks. Sn is said to have a high solubility in the α-phase and will therefore, when it is present to the given amount, be dissolved in the α-phase and contribute to improved corrosion properties and mechanical properties of the same. It is pointed out that a too low content of Sn (below 0.9 percentage by weight) in the material influences the tensile strength of the material both in the long and in the short term. Furthermore, such a low Sn-content suppresses to a smaller extent a negative effect of a possible nitrogen incorporation on the corrosion resistance of the material. A Sn content above 1.5 percentage by weight influences the susceptibility of the material to working and in particular to cold working. It is mentioned that Si and C contribute to a reduction of the size of the particles and to bring about a structural homogeneity in the material. Oxygen is said to contribute to a finer structure of the material and is also used as a means for reinforcing the material through the solid solution, a so-called “solid solution strengthener”. Nb is said to contribute to the strength properties of Zr and increases the corrosion resistance of the alloy by forming secondary phase particles together with Zr and Fe. It is furthermore pointed out that with a Nb-content below 0.5 percentage by weight of the material, a Fe-content below 0.3 percentage by weight and a Cr-content below 0.005 percentage by weight, the total portion of secondary phase particles of the above-mentioned kind in the α-zirconium matrix of the end product is considerably lower than 60 percentage by volume of the total amount of iron-containing secondary phase particles, which results in that the corrosion resistance of the material is negatively influenced. With a Nb-content above 1.5 percentage by weight, a large number of large particles of β-Nb phase are formed in the material, which also reduces the corrosion resistance of the same. It is also mentioned that a Cr-content above 0.2 percentage by weight may result in the formation of binary intermetallic compounds of Zr—Cr, which has an opposite i.e. negative, influence on the workability and the tensile strength of the material. A purpose with the present invention is to provide a zirconium-based alloy with such a composition that the distribution of and the size of secondary phase particles in the alloy, the kind of secondary phase particles and the content of different alloying elements in the α-phase of the alloy are such that the alloy is optimized with respect to physical and mechanical properties as well as corrosion properties. In particular, these properties should be optimized with respect to an application where the alloy is subjected to an increased radiation of the fast-neutron kind in a corrosive environment, such as in the reactor core of a nuclear energy plant. In particular it is aimed at improved corrosion properties of the alloy with respect to the corrosion properties of the above-mentioned alloys according to the prior art. This purpose is achieved by means of an alloy of the kind initially defined, which alloy is characterised in that it comprises 0.5-0.85 percentage by weight Sn. This choice of Sn-content stands in opposition to that which, according to the prior art, is a preferred interval for the Sn-content. The applicant has however found that improved corrosion properties, in particular in the environment which is the case in the area of the reactor core of a nuclear energy plant, may be achieved in the zirconium-based alloy by a careful choice of the Sn-content within the defined interval. According to a preferred embodiment of the alloy, the content of Sn in the alloy is larger than or equal to 0.65 percentage by weight. A preferred interval for the Sn-content should thus be 0.65-0.85 percentage by weight with the purpose of achieving as good corrosion properties in the alloy as possible under the otherwise given conditions. According to a further preferred embodiment, the alloy comprises up to 0.2 percentage by weight Ni. Thereby secondary phase particles containing Zr, Ni and Fe may be obtained in the alloy. Such secondary phase particles contribute to improved corrosion properties of the alloy and have good stability under neutron radiation. According to a further preferred embodiment, the alloy comprises up to 0.6 percentage by weight Cr, which is more than the maximum 0.2 percentage by weight which has previously been recommended with respect to the formation of binary intermetallic compounds of Cr and Zr. With the remaining composition which the alloy according to the invention has, a content of up to 0.6 percentage by weight Cr may however be permitted in order to improve the corrosion properties of the alloy, without the alloy thereby obtaining considerably worse mechanical properties, such as a deteriorated tensile strength. Unlike the prior art, the present invention thus suggests a zirconium-based alloy with a Cr-content above 0.2 percentage by weight, up to 0.6 percentage by weight. According to a further preferred embodiment, the total content of Nb and Sn is larger than or equal to 1.15 percentage by weight. Such a total content of Nb and Sn contributes to improved mechanical properties of the alloy. Which requirements on mechanical properties and corrosion properties that finally are demanded on the alloy depend on in which application the alloy finally is to be used. According to a preferred embodiment of the invention, the alloy constitutes at least a part of a component in a nuclear energy plant. The component is preferably arranged in the area of the reactor core and constitutes, according to a further preferred embodiment, a part of a fuel assembly. In such an application high requirements will at least be demanded on the corrosion properties of the alloy. Depending on to which extent the component has a supporting function, specific requirements will also be demanded on the mechanical properties of the alloy. An alloy of the kind which is suggested by the invention is in particular suitable to constitute at least a part of a cladding tube, a spacer or a box. A further purpose of the invention is to provide a component in a nuclear energy plant, which component in particular has satisfactory corrosion properties with respect to the specific conditions which may be assumed to be the case in the nuclear energy plant, in particular in the area of the core of the same, where the component is subjected to an increased radiation of the fast neutron kind, in a corrosive environment, e.g. surrounded by a corrosive medium, such as water. This purpose is achieved by means of a component of the initially defined kind, which comprises an alloy according to the invention. According to a preferred embodiment, the component constitutes a part of a fuel assembly, i.e. it is arranged in the area of the reactor core. Thereby specific requirements are demanded on its corrosion properties in the environment of increased radiation and corrosive media which it is subjected to. The choice of a zirconium-base alloy with a suitable composition is consequently highly important. According to a further preferred embodiment, the component defines a cladding tube. Thereby also specific mechanical properties of the component are required, which are fulfilled by the alloy according to the invention. According to a further preferred embodiment, at least a part of the inner circumference of the cladding tube comprises a layer of a material which is more ductile than the alloy according to the invention. The cladding tube is thereby made less sensitive to the direct contact with the fuel within these. The risk for crack formation of the cladding tube in areas where it comes into direct contact with and possibly is subjected to wear caused by the fuel is reduced, under the condition that the layer of the more ductile material is arranged in these areas, which preferably is the case. Said layer comprises here a zirconium-based alloy with a total content of alloying materials which does not exceed 0.5 percentage by weight. Further advantages with and features of the alloy according to the invention and the component, respectively, will be clear from the following, detailed description. A component arranged in a nuclear energy plant, more precisely in the area of the reactor core, is subjected to increased radiation of the fast neutron kind in a corrosive environment. The reactor may be a pressure water or a boiling water reactor. The component constitutes a part of the fuel assembly. In this example the component is a cladding tube arranged to contain the reactor fuel. The component comprises a zirconium-based alloy which has the following composition: 0.5-0.85 percentage by weight Sn, 0.3-0.6 percentage by weight Fe, 0-0.6 percentage by weight Cr, 0-0.2 percentage by weight Ni, 0.65-1.6 percentage by weight Nb and the rest zirconium. The content of Ni is preferably within the interval 0.05-0.2 percentage by weight. According to an alternative embodiment the alloy comprises 0.65-0.85 percentage by weight Sn and 0.5-1.6 percentage by weight Nb, with the remaining elements within the previously mentioned intervals. The cladding tube may be formed from a solid bar, in the centre of which a hole has been drilled. Furthermore, the component has, in addition to prior annealings in connection with the working of the same, finally been annealed in the β-phase area of the alloy and then been quenched by a β-quenching in the α-phase area of the alloy. By the annealing in the β-phase area, course structures and other effects of the prior heat treatment history are removed from the alloy. Furthermore, the orientated texture which has been obtained during prior working of the work piece of the tube is removed, whereby different tendencies to growth in different directions of the component, when it is exposed to neutron radiation in the core, are avoided. The cooling to the α-phase area is so fast that an entity of short α-phase laminae is formed in the prior α-phase grains. Short α-laminae improve the mechanical strength of the alloy. Furthermore at the quenching from the β-phase area to the α-phase area secondary phase particles of intermetallic compounds, such as Zr(Nb,Fe)2, Zr(Fe,Cr,Nb) and (Zr,Nb)3Fe, are precipitated, which favours good anticorrosive and mechanical properties of the finished alloy and thereby of the component. The quenching speed should thereby be adjusted such that an optimal secondary phase particle distribution and secondary phase average particle size are obtained. The alloy is preferably cooled with a cooling speed below 100° C./second, preferably below 50° C./second and most preferred in order of magnitude 5-20° C./second. When the component, such as here, is a cladding tube, preferably a layer with a lower total content of alloying elements than the remaining alloy is applied on the inner circumference of the cladding tube. The total content of alloying materials in this layer is preferably below 0.5 percentage by weight, wherein the remaining part constitutes Zr. This layer makes the cladding tube more resistant to mechanical influence from the reactor fuel which is arranged in the tube and which physically may rest against and cause tensions in the walls of the cladding tube. Preferably the alloy according to the invention comprises no essential amount of other materials than those which have been mentioned above. It should however be noted that small amounts of impurities may exist in the alloy. Typical impurities which may exist in zirconium-based alloys are specified in the table below. Furthermore, small amounts of Si and O may exist in the alloy. Typical contents of these materials are also given below: TABLEElementAlBCCaCdClCoCuHHfMax. ppm750.5270300.520205025100ElementMgMnMoNNaPbSiTiUMax. ppm2050508020130120503.5 Si and O may exist in contents where Si is 50-120 ppm and O is 500-1600 ppm. It should be realised that a number of alternative embodiments of the alloy and the component according to the invention will be obvious to a person skilled in the art but still be within the scope of the invention, such as it is defined in the annexed claims. |
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044773773 | abstract | A process of recovering cesium ions from mixtures of ions containing them and other ions, e.g., a solution of nuclear waste materials, which comprises establishing a separate source phase containing such a mixture of ions, establishing a separate recipient phase, establishing a liquid membrane phase in interfacial contact with said source and recipient phases, said membrane phase containing a ligand, preferably a selected calixarene as depicted in the drawing, maintaining said interfacial contact for a period of time long enough to transport by said ligand a substantial portion of the cesium ion from the source phase to the recipient phase, and recovering the cesium ion from the recipient phase. The separation of the source and recipient phases may be by the membrane phase only, e.g., where these aqueous phases are emulsified as dispersed phases in a continuous membrane phase, or may include a physical barrier as well, e.g., an open-top outer container with an inner open-ended container of smaller cross-section mounted in the outer container with its open bottom end spaced from and above the closed bottom of the outer container so that the membrane phase may fill the outer container to a level above the bottom of the inner container and have floating on its upper surface a source phase and a recipient phase separated by the wall of the inner container as a physical barrier. A preferred solvent for the ligand is a mixture of methylene chloride and carbon tetrachloride. |
description | This application is a continuation of and claims priority under 35 U.S.C. §120 to application Ser. No. 10/678,170 “METHOD AND APPARATUS FOR FACILITATING RECOVERY OF NUCLEAR FUEL FROM FUEL POOL,” now U.S. Pat. No. 7,636,652, filed Oct. 6, 2003, the entirety of which is incorporated by reference. The information setting forth the placement of fuel bundles, each of which has various attributes, in a nuclear reactor core is referred to as the loading map. In conventional core design, creating the loading map is an experienced based, trial and error, iterative process. The core designer generally receives plant specific critical to quality factors such as plant cycle energy requirements, thermal and operational limits, shut down margins, etc. The core designer will also have information on the layout of the reactor core; namely, an indication of the how the nuclear fuel bundles are positioned within the core. Some of the critical to quality factors may even concern the layout. For example, the core designer may receive input requiring the positioning of certain fuel bundles within the layout. Given this information, the core designer then makes a guess, based on experience and various rules of thumb he may have developed over time, on the initial positioning of fuel bundles in the reactor core. Specifically, the core designer guesses how many fresh fuel bundles to place in the core, and what types of fresh fuel bundles to use. A fresh fuel bundle is a fuel bundle that has not been exposed. Fuel bundles of the same type have substantially the same attributes. The attributes include but are not limited to: uranium loading, average enrichment, gadolinia loading, number of axial zones, product line, and thermal-mechanical characteristics of the fuel bundles. Different types of fresh fuel bundles have one or more different attributes. In deciding how many fresh fuel bundles to use, the core designer is also deciding how many of the fuel bundles currently in the core to reuse. Reusing the fuel bundles currently present in the core can mean leaving a fuel bundle in its existing location, or moving the fuel bundle to a different location in the core. As part of the core design, the core designer also determines other operational parameters of the reactor core such as control blade positions, core flow, etc. Having specified these operational control parameters, a Nuclear Regulatory Commission (NRC) licensed simulation program is then run on the initial core design. Based on the results of the simulation, the core designer utilizes experience and rules of thumb to fix perceived problems in the design and, in general, improve the design; particularly with respect to the critical to quality factors. These changes may include changing the loading map. The process repeats until the core designer is satisfied with the design. The present invention provides a method and apparatus for using nuclear fuel discarded to one or more fuel pools in a loading map for a new cycle of a nuclear reactor. In one exemplary embodiment, a graphical user interface under the control of a computer processor provides a user with the capability to selectively populate a loading map with fuel bundles residing in at least one fuel pool. For example, the computer processor may include a memory storing at least one fuel pool database. The fuel pool database includes a list of at least a portion of the fuel bundles residing in the fuel pool, and the user may select which of these fuel bundles to use in creating the loading map. In an exemplary embodiment, the fuel pool database indicates one or more attributes for the listed fuel bundles, and the graphical user interface that includes one or more fuel pool database management tools for aiding in the selection process. For example the tools may provide for filtering and/or sorting the fuel pool database. In a further exemplary embodiment, the a graphical user interface is controlled to further allow the user to selectively populate the loading map with different types of fresh fuel bundles. For example, the computer processor may include a memory storing at least one fresh bundle type database. The fresh bundle type database includes a list of fresh bundle types, and the user may select which of these fuel bundles to use in creating the loading map. In an exemplary embodiment, the fresh bundle type database indicates one or more attributes for the listed fuel bundles types, and the graphical user interface that includes one or more fresh bundle type database management tools for aiding in the selection process. For example the tools may provide for filtering and/or sorting the fresh bundle type database. A reactor may then be operated using a loading map that contains fuel bundles recovered from one or more fuel pools. FIG. 1 illustrates an embodiment of an architecture according to the present invention. As shown, a server 10 includes a graphical user interface 12 connected to a processor 14. The processor 14 is connected to a memory 16. The server 10 is directly accessible by a user input device 18 (e.g., a display, keyboard and mouse). The server 10 is also accessible by computers 22 and 26 over an intranet 20 and the Internet 24, respectively. The operation of the architecture shown in FIG. 1 will be discussed in detail below. Creating a Template A user via input 18, computer 26 or computer 22 accesses the server 10 over the graphical user interface 12, and runs a loading map editor program stored in memory 16 according to an exemplary embodiment of the present invention. The loading map editor provides for creating and editing a graphical representation of a nuclear reactor core referred to as a template. However, another form of conveying this information, such as a text file, may also be thought of as the template. FIG. 2 illustrates a quarter-core screen shot of a partially completed template designed according to the methodologies of the present invention using the loading map editor of the present invention. When the loading map editor is initially run, the user has the option via a file menu 30 to access a previously created template or to begin a new template. Assuming the user begins a new template, the loading map editor request the user to identify the nuclear reactor for which the template is being created. The loading map editor then retrieves the geometry of the identified nuclear reactor from a relational database containing nuclear reactor plant characteristics stored in the memory 18. The loading map editor then displays a blank colorless fuel bundle field 36 of the appropriate size based on the retrieved plant characteristics with the rows and columns numbered (such as with the fuel bundle position Row 6, Column 3 in FIG. 2). Within the fuel bundle field 36, the user may then, for example, using a mouse associated with the input 18, computer 26 and computer 22 click on the fuel bundle positions 38 in the array of possible fuel bundle positions to identify the type (fresh, reinsert, or locked) and grouping of the actual fuel bundle in that position. In the context of a template, a bundle group consists of 1, 2, 4, or 8 bundles and an associated symmetry pairing of bundles within the group which may be performed either mirror or rotationally symmetric. As shown on the right side of FIG. 2, the loading map editor provides several tools for performing this assignment task. Specifically, the tools include the headings Load Type 40, Bundle Grouping 50 and Numbering Mode 60. Under the Load Type 40 tool heading, the loading map editor includes a Fresh radio button 42, a Reinsert radio button 44 and a Locked radio button 46. The Fresh, Reinsert and Locked radio buttons 42, 44 and 46 correspond to fresh, reinsert and locked fuel bundle categories. The user, for example, clicks on the desired radio button to choose the desired category and then clicks on the fuel bundle position 38 in the fuel bundle field 36 to assign that category to the fuel bundle position 38. The fresh fuel bundle category indicates to insert fuel bundles that have not been exposed. The loading map editor then displays “F” and a number “N” at the bottom of the fuel bundle position 38. The “F” indicates the fresh fuel bundle category, and the number “N” indicates the Nth fresh bundle type 38. As will be appreciated, the loading map editor maintains a count of the number of fuel bundle types assigned to the core. Multiple bundle positions can be assigned the same bundle type by specifying the same “F” and “N” value for each position. The locked fuel bundle category indicates that a fuel bundle currently occupying an associated fuel bundle position in an actual nuclear reactor core is to remain in that position in creating a new nuclear reactor core loading map. The loading map editor displays “L” and a number “N” in the fuel bundle position 38 when the locked fuel bundle category is assigned. The “L” indicates the locked fuel bundle category, and the number “N” indicates the Nth locked bundle group. The reinsert fuel bundle category indicates to insert a fuel bundle that has been exposed. The loading map editor displays only a number “N” in the fuel bundle position 38 when the reinsert fuel bundle category is assigned. The number indicates a priority of the fuel bundle position 38. The number and the priority indicated by the number will be described in detail below with respect to the Numbering Mode 60 heading. In an exemplary embodiment, the loading map editor displays the fuel bundle positions 38 in a color associated with the assigned category. For example, fresh are displayed in blue, locked are displayed in yellow, and reinserted are displayed in violet. Under the Bundle Grouping 50 heading, the loading map editor includes a “1” radio button, a “2” radio button, a “4” radio button, and an “8” radio button. When the “1” radio button is selected by the user, for example, by clicking on the “1” radio button, the category assigned by the user to a fuel bundle position 38 is associated only with the fuel bundle position 38 chosen. Selecting the “2” radio button and assigning a category to a fuel bundle position 38 causes the category to be assigned to the selected fuel bundle position as well as the fuel bundle position 180 degrees symmetric to the selected fuel bundle position. Selecting the “4” radio button causes the loading map editor to request the user to chose between rotational and mirror symmetry. Rotational symmetry is an image property indicating there is a center point around which the object is turned a certain number of degrees and the object still looks the same (i.e., it matches itself a number of times while it is being rotated). Mirror symmetry (or line symmetry) indicates a correspondence in size, shape, and relative position of parts on opposite sides of a dividing line. If the user assigns a category to a fuel bundle position when rotational symmetry is chosen, this causes the category to be assigned to the selected fuel bundle position as well as the fuel bundle position 38 in each of the other quadrants rotationally symmetric to the selected fuel bundle position. If the user assigns a category to a fuel bundle position when mirror symmetry is chosen, this causes the category to be assigned to the selected fuel bundle position as well as the fuel bundle position in each of the other quadrants symmetric to the selected fuel bundle position. Selecting the “8” radio button causes the loading map editor to consider the total fuel bundle field 36 as octant symmetric—eight symmetric pie pieces. Assigning a category to a fuel bundle position when the “8” radio button is selected causes the category to be assigned to the selected fuel bundle position 38 as well as the fuel bundle positions 38 in each of the other eight pie pieces symmetric to the selected fuel bundle position 38. Under the Numbering Mode 60 heading, the loading map editor includes an Automatic radio button 62 and a Manual radio button 64. Choosing between an automatic numbering mode by selecting the Automatic radio button 62 and a manual numbering mode by selecting the Manual radio button 64 is only permitted when the Reinsert radio button 44 or Fresh radio button 42 has been selected. The numbering mode in general is inapplicable when the Locked radio button 46 is selected. When the Automatic radio button 62 is selected, the loading map editor, which maintains a count of the number of fuel bundle positions 38 assigned the reinsert fuel bundle category, assigns the count plus one to the next fuel bundle position 38 assigned the reinsert fuel bundle category. The assigned number is displayed at the bottom of the fuel bundle position 38. Likewise, the loading map editor maintains a count of the fresh bundle types. When a fuel bundle position 38 is assigned the fresh bundle category the count plus one, referred to above as N, is assigned to that position. “F” and the value of N are displayed at the bottom of the fresh fuel bundle position. When the Manual radio button 64 is selected, the loading map editor maintains the count of the number of fuel bundle positions 38 assigned the reinsert fuel bundle category, but does not assign numbers to the fuel bundle positions 38. Instead, the user may position a cursor in the fuel bundle position 38 and enter the number manually. As alluded to above, the assigned numbers represent assigned priorities. The priorities indicate an order for loading exposed fuel bundles based on an attribute of the exposed fuel bundles. The attributes include, but are not limited to, K infinity (which is a well-known measure of the energy content of the fuel bundle, exposure of the bundle (which is accumulated mega-watt days per metric ton of uranium in the bundle), residence time of the bundle (which is how long the bundle has been resident in the nuclear reactor core), etc. In one exemplary embodiment, the shade of the color associated with the reinserted fuel bundle positions varies (lighter or darker) in association with the assigned priority. The loading map editor according to the present invention also provides several viewing options via a view menu 34 and a zoom slide button 70. Adjusting the zoom slide button 70 by clicking and dragging the zoom slide button 70 to the left and the right decreases and increases the size of the displayed fuel bundle field 36. Under the view menu 34, the user has the option to view a single quadrant of the template, or a full core view of the template. Additionally, the user can control whether certain template attributes are displayed. Specifically, the view menu 34 includes the options of displaying the following in the loading template: control blades, bundle coordinates, core coordinates, etc. Having created the loading template, the user may save the template, or even a partially created template, to the memory 18 by selecting either the “Save” or “Save As” option in the file menu 30. As discussed above, instead of creating a new template, a previously created template may be viewed and, optionally, edited. Using the file menu 30, the user selects an “open” option. The loading map editor then displays the accessible templates stored in the memory 18 or a directory of memory 18. The user then selects an accessible template, for example, by clicking on one of the accessible templates. The loading map editor will then display the chosen template. The user may then edit the chosen template. For example, after selecting a fuel bundle position 38 the user may select under the edit menu to “clear” the category assigned to the fuel bundle position 38. Besides the category assigned to this fuel bundle position 38, the loading map editor also clears the category assigned to associated fuel bundle positions 38. Associated fuel bundle positions 38 are those fuel bundle positions 38 that were assigned the fuel bundle category along with the fuel bundle position 38 selected for clearing because of the bundle grouping chosen when the category was assigned to the fuel bundle position 38 chosen for clearing. When fuel bundle positions 38 assigned the fresh or reinserted category are cleared, the loading map editor adjusts the numbering associated with that category. In the case of the fresh bundle category, this is a conditional action based on whether other bundle positions have been assigned the same fresh bundle type. Specifically, the loading map editor performs a cascade operation such that fuel bundle positions assigned the same category and having higher numbers are renumbered in sequence beginning from the lowest number of a deleted fuel bundle position. For example, if reinsert bundle positions numbered 44, 43 and 42 were cleared, then reinsert bundle position having number 45 would be renumbered 42, reinsert bundle position having number 46 would be renumbered 43, etc. The loading map editor also changes the total count of fuel bundle positions assigned the category being cleared. When unassigned bundle positions are created through editing, the user may then newly assign categories to the unassigned bundle positions in the same manner and using the same tools to create a template as described above. In so doing, the user may decide to manually assign, for example, an existing priority to a newly assigned reinsert fuel bundle position. In this instance, the reinsert fuel bundle position already having this number and each reinsert fuel bundle position having a higher number are incremented by one. As a further alternative, the user may want to adapt an existing template for one reactor to another reactor of the same size and physical bundle configuration. To do this, the user may use the “save as” feature in the file menu 30 to create a duplicate of the loading template. Subsequent changes to the bundle field will then apply to the copied template. In addition to creating a template from ‘scratch’ or editing an existing template, the user may have the loading map editor derive a template from a previously loaded core. In the loading map editor, using the file menu 30, the user selects an “auto-generate template” option. The loading map editor then displays a list of the accessible fuel cycles stored in the memory 18. Each fuel cycle corresponds to an actual loading map for a fuel cycle of a nuclear reactor. As will be appreciated, the memory 18 may store loading maps for cycles of different nuclear reactors. Accordingly, the list of cycles displayed by the loading map editor identifies both the nuclear reactor and the cycle. From the list the user selects the cycle (hereinafter “the selected cycle”) that the template will be derived from. The loading map editor then accesses the loading map for the selected cycle. The user is then presented with a dialog box for entering input parameters of the derivation process. The input parameters include: a primary attribute (e.g., exposure, K infinity, etc.) for deriving the template, a tolerance level (discussed in detail below), group list members (8, 4, or 2 bundle groupings), bundle symmetry for groups of 4, and a maximum number of assignments to each group list member. For example the user may enter K infinity as the primary attribute, and a tolerance level of 0.2 (which, as described in detail below, is used for forming bundle groups). The user may further enter that groups of 8 and 4 are permitted, the groups of 4 should have mirror symmetry and that a maximum of 14 groups of 4 are permitted. In an exemplary embodiment, the loading map editor provides the user with a drop-down menu. The user selects list members desired for the template from the options given in the drop-down menu. These options include: groups of 8, 4 and 2; groups of 8 and 4; groups of 8 (which forces groups of 4 on the minor axis of the reactor core template); and groups of 4 and 2. In selecting the maximum number of assignments for each group, the user enters this data in the order of the smallest to the largest group size. However, the maximum number of assignments for the largest groups is not entered by the user, as this value is automatically determined based on the maximum number of assignments for the smaller groups. Once the user enters the input parameters, the loading map editor will begin generating a template. First the loading map editor asks the user if locked bundle positions are permitted, if so, then the loading map editor requests the user to identify the cycle previous to the selected cycle in the same manner that the selected cycle was identified. The loading map editor then compares the loading map for the selected cycle with the loading map for the previous cycle of the identified nuclear reactor. Specifically, for each bundle position in the reactor, the loading map editor determines if loading maps for the selected and previous cycles have a bundle with the same serial number in the same bundle position. If so, the bundle position is assigned the locked fuel bundle category in the loading template. After the locked fuel bundle positions are identified, the loading map editor identifies the fresh fuel bundle positions. Specifically, for each bundle position not already identified as a locked bundle position, the loading map editor determines from the characteristics of the selected loading map if the fuel bundle in that bundle position is a fresh fuel bundle. For each identified fresh fuel bundle, the loading map editor also determines the type of fresh fuel bundle from the characteristics of the selected loading map. The loading map editor then assigns the fresh fuel category to the associated fuel bundle position in the template and assigns a type count number N to the fuel bundle position. For each type of fresh fuel bundle located in the selected loading map, the loading map editor assigns a count value to that type. This count value is then assigned to the bundle position along with the fresh fuel bundle category assignment so that fresh fuel bundle positions that should have the same type of fresh fuel bundle are identified by the same value ‘N’ in the loading template. Next, the loading map editor determines whether the identified fresh bundle category positions form any bundle groups. As discussed above, the user identifies the bundle group members permitted in the template. The bundle group members form a group members list. For each bundle position assigned the fresh fuel bundle category, the loading map editor first determines if the bundle position (hereinafter the “current bundle position”) has already been assigned to a group. If so, then the loading map editor proceeds to the next bundle position. If not, then the loading map editor selects the largest group from the group member list and identifies each of the bundle positions that form such a group with the current bundle position. If each of the bundles positions forming the group has been assigned the fresh bundle category and are of the same type as the current bundle position, then the loading map editor records the group of bundle positions as a group. If each of the bundles positions forming the group has not been assigned the fresh bundle category or one of the bundles is not the same type as the current bundle position, then the loading map editor performs the above-described process for the next largest bundle group in the group member list. This process keeps repeating until a group is formed or there are no more groups in the group member list to test. If the members of the group member list have been tested, and no group has been formed, then the current bundle position is recorded as not belonging to a group. Next, the loading map editor identifies the reinserted fuel bundle positions. The bundle positions of the template not assigned to the locked or fresh fuel bundle categories are assigned the reinserted fuel bundle category. Then, the loading map editor determines whether the reinserted bundle category positions form any bundle groups. For each bundle position assigned the reinserted fuel bundle category, the loading map editor first determines if the bundle position (hereinafter the “current bundle position”) has already been assigned to a group. If so, then the loading map editor proceeds to the next bundle position. If not, then the loading map editor selects the largest group from the group member list and identifies each of the bundle positions that form the group with the current bundle position. If each of the bundles positions forming the group has not been assigned the reinserted bundle category, then the loading map editor determines if the next largest group in the group member list includes all reinserted fuel bundle positions. If no group from the group member list results in a group of reinserted fuel bundles, then the loading map editor records the current fuel bundle position as not belonging to a group. Once a group has been formed, the loading map editor calculates the average attribute value for the group. As discussed above, the user identified a primary attribute to use in deriving the template. Here, the loading map editor uses that attribute value for each fuel bundle in the selected loading map forming the associated group in the template to calculate the average attribute value. The loading map editor then determines if the attribute value for each fuel bundle in the group is with the tolerance level from the average attribute. Again, here, the tolerance level was a user input design parameter as discussed above. If the attribute value for each fuel bundle in the group is within the tolerance level of the average attribute value, then the loading map editor records the associated fuel bundle positions in the template as belonging to a group. Otherwise, the loading map editor performs the above-described process for the next largest bundle group in the group member list. This process keeps repeating until a group is formed or there are no more groups in the group member list to test. If the members of the group member list have been tested, and no group has been formed, then the current bundle position is recorded as not belonging to a group. The loading map editor then determines if the user specified maximum for a group in the group member list has been violated. If so the editor performs a group recombination and ranking process. For example, if the number of groups of 2 exceeds the user specified maximum the editor does the following: For each group of 2, the loading map editor determines if another group of 2 forms a group of 4 meeting the symmetry requirements entered by the user. The loading map editor then determines the average attribute value and standard deviation for each newly formed potential group of 4 and ranks the potential groups of 4 based on minimum standard deviation. Next, the highest ranked groups (i.e., those with the lowest standard deviation) are assigned to the groups of 4 until the groups of 2 list does not exceed the maximum number allowed based on the user input. Those potential groups of 4 not assigned remain as groups of two. Next, the same process is performed to combine groups of 4 into groups of 8 assuming the user input parameters permit groups of 8 and the user specified maximum for groups of 4 has been violated. As a final step, the reinserted fuel bundles are assigned a priority number that, as described above, appears in the template. The fuel bundles positions are ranked based on (1) the attribute value for the fuel bundle in the associated position in the loading map if the fuel bundle position does not form part of a group; or (2) by the average attribute value of the group if fuel bundle position does form part of a group. A priority number is then assigned by this ranking with the fuel bundles having the same average attribute assigned the same priority number. This completes the template derivation process, the resulting template is then displayed in the loading map editor allowing the user to save the resulting template for future use. Using the present invention as described above, a core designer may capture his experience and rules of thumb associated with the initial design of a loading map. Furthermore, this knowledge may then be used by others to improve or adapt templates to existing core designs. Creating Loading Map The loading map editor according to the present invention includes additional functionality that allows the user to generate a loading map from the loading template. In addition, the loading map editor provides increased flexibility in creating the loading map by allowing the user the option of reloading fuel bundles currently residing in one or more fuel pools. After accessing, creating and/or editing a reactor core template using the loading map editor as discussed above, the user may then create a loading map using the template. From the file menu 30, the user chooses a “load” option. The loading map editor then displays a loading screen that includes a template access window, template information window, reload window and a load fresh window. The template access window provides a user with a drop down menu for selecting a loading template stored in the memory 18. The template information window displays summary information for the selected loading template. The summary information includes, but is not limited to, the number of fresh bundle types, the number of reinserted fuel bundle positions and the number of locked bundle positions in the loading template. The summary information may also indicate the number of fresh bundle types and number of reinserted bundles currently added in creating the loading map. FIG. 3 illustrates an exemplary embodiment of a reload window displayed by the loading map editor. The window is divided into two parts: a filtered fuel pool table 100 and a reloading pool 200. The filtered fuel pool table 100 lists (1) the exposed fuel bundles currently in the nuclear reactor under consideration, except for those fuel bundles in locked fuel bundle positions 38, and (2) the fuel bundles in one or more fuel pools for this and other nuclear reactors. As is well-known, exposed fuel bundles removed from a nuclear reactor are stored in what is known as a fuel pool. Fuel bundles from two or more nuclear reactor cores located at a same site may be stored in the same fuel pool. As shown in FIG. 3, the filtered fuel pool table 100 lists each exposed fuel bundle by its serial number and bundle name. Each fuel bundle is assigned a unique serial number, used to assure traceability of the bundle from a quality assurance perspective. The bundle name is a character string identifier used to identify the fuel bundle product line as well as nuclear characteristics, such as uranium and gadolinia loading. The filtered fuel pool table 100 also lists one or more attributes of each exposed fuel bundle listed. These attributes may include K infinity, exposure, and the last fuel cycle number for which the bundle was resident in the core. Additional attributes for an exposed fuel bundle may include: 1) bundle product line, 2) initial uranium loading, 3) initial gadolinium loading, 4) number of axial zones, 5) historical fuel cycle numbers previous to the most recent for which the bundle was resident in the core, 6) the corresponding reactor in which the fuel bundle was resident for each of the historical fuel cycles, 7) accumulated residence time, and 8) fuel bundle pedigree, a parameter that reflects the usability of the bundle for continued reactor operation. The fuel bundle pedigree is determined from a number of factors the foremost being an inspection of the fuel, either visually or by some other non-destructive test procedure, which is designed to detect a current failed fuel bundle or the vulnerability of the bundle to future failure. Failure mechanisms include such items as corrosion, debris impact, and mechanical bowing of the fuel bundle. Another factor affecting pedigree is possible reconstitution of a fuel bundle, which is a repair process involving the replacement of damaged fuel rods with replacement rods that may be a uranium containing fuel rod or alternatively, a non-uranium containing rod (e.g. stainless steel), henceforth referred to as a ‘phantom’ rod. A pedigree attribute might be ‘RU’ and ‘RP’ for reconstituted with uranium and phantom rods, respectively, and ‘DC’, ‘DD’ and ‘DB’ for damaged by corrosion, debris, and bow, respectively. A ‘blank’ pedigree attribute would designate a bundle that was undamaged and useable. All attributes with the exception of bundle pedigree are populated within the database via a direct relationship with the historical fuel cycles. The fuel pedigree attribute for non ‘blank’ designations are entered into the database via a separate process that is tied to fuel inspection and reconstitution services. In this process, the fuel bundles in a fuel pool are inspected and the pedigrees of the fuel bundles ascertained from the inspection. Then, a bundle status program is accessed. The bundle status program provides a GUI menu for ‘Fuel Inspection’, which is accessed by the user. The user clicks on the pulldown menu ‘Add’ from the ‘Fuel Inspection’ menu, and is presented with a pop-up for typing in the bundle serial number and the pedigree designation, such as ‘DD’ corresponding to a debris damaged bundle. The pedigree data entered in this manner is associated with the fuel pool database. The user may also click a ‘Census’ option from the ‘Fuel Inspection’ menu. Selecting this option will perform a query of the fuel pool database and present the user with a list of bundle serial numbers and corresponding attribute data, as described previously, for those bundles containing a non-null pedigree designation. The user may elect to change existing pedigree information by selecting the bundle entry, right-clicking a ‘Modify’ option, which activates the pedigree attribute field, and entering the modified pedigree information. For example, a bundle that was previously damaged may have been reconstituted. Alternatively, the user may right-click a ‘Delete’ option, which has the effect of reverting the bundle pedigree status back to null. The reloading fuel pool table 200 provides the same information for each fuel bundle as provided by the filtered fuel pool table 100. Additionally, the reloading fuel pool table 200 indicates the priority number 202 for each fuel bundle group as set forth in the loading template. As discussed above with respect to the loading template, reinserted fuel bundles may be assigned as a group of 1, 2, 4 or 8 bundles. Accordingly, FIG. 3 shows that the highest priority reinserted fuel bundle position(s) are a group of four fuel bundles, and the next highest priority reinserted fuel bundle(s) are a group of eight fuel bundles. The reloading fuel pool table 200 is populated by moving fuel bundles from the filtered fuel pool table 100 into the reloading fuel pool table 200. As further shown in FIG. 3, the reload window further includes a set of tools 120 for aiding the user in selecting and moving fuel bundles from the filtered fuel pool table 100 to the reload fuel pool table 200. The set of tools 120 include, but are not limited to, a filter tool 130, a move right tool 160, a move left tool 170 and a delete tool 180. A user selects the filter tool 130 by, for example, clicking on the filter tool 130. This opens a filter window as shown in FIG. 4. As shown, the filter window lists the same attributes listed in the filtered fuel pool table 100, and allows the user to indicate to filter based on the attribute by clicking in the selection box 132 associated with the attribute. When an attribute has been selected, a check is displayed in the associated selection box 132. The user may also unselect an attribute by again clicking in the associated selection box. In this case, the check mark will be removed. For each attribute, the filter window may display one or more filter characteristics associated with the attribute. For example, for the filter characteristics of the K infinity attribute, the user may select a filter operator 134 of greater than, less than, or equal to and enter in a filter amount 136 associated with the filter operator 134. As shown in FIG. 4, a user has selected to filter based on K infinity, chosen the greater than filter operator, and entered the filter amount of 1.2. As a result, the loading map editor will filter the fuel bundles in the filtered fuel pool table 100 to display only those fuel bundles having a K infinity greater than 1.2. As another example, the exposure attribute also has an associated filter operator and filter amount. As will be appreciated, the filter characteristics of an attribute will depend on the attribute. Also, as will be appreciated, other methodologies for indicating the filter characteristics may be possible. For example, for the cycle attribute, the filter window provides a drop down menu for selecting the cycle number. FIG. 4 shows cycles 2 and 4 selected from the drop down menu for the cycle attribute. As a result, the loading map editor filters the filtered fuel pool table 100 to display only those fuel bundles whose most recent residence was in cycle 2 or cycle 4. Similarly, the user may elect to filter bundles based on their pedigree, product line, etc. Once the attributes for filtering on have been selected and the filter characteristics have been entered, the user causes the loading map editor to filter the filtered fuel pool table based on this information by clicking on the OK selection box. Alternatively, the user may cancel the filter operation by clicking on the CANCEL selection box. The filtered fuel pool table 100 also provides a filtering mechanism for filtering the fuel bundles listed therein. A user may sort the filtered fuel pool table 100 in ascending or descending order of an attribute by clicking on the attribute heading in the filtered fuel pool table 100. Once the user clicks on the attribute, the loading map editor displays a popup menu with the options “Sort Ascending” and “Sort Descending”. The filtered fuel pool table 100 is then filtered in ascending or descending order of the attribute based on the option clicked on by the user. To move fuel bundles from the filtered fuel pool table 100 to the reload fuel pool table 200, the user selects the fuel bundles for transfer by clicking and dragging to highlight one or more of the fuel bundles in the filtered fuel pool table 100. Then the user clicks on the move right tool 160. This causes the selected fuel bundles to populate the highest priority unpopulated fuel bundle positions in the reload fuel pool table 200. Alternatively, a user clicks and drags the highlighted fuel bundles into one of the priority sections of the reloading fuel pool table 200. Fuel bundles may also be moved from the reload fuel pool table 200 back into the filtered fuel pool table 100 by selecting fuel bundles in the reload fuel pool table 200 and clicking on the move left tool 170. Alternatively, the selected fuel bundles may be clicked and dragged back to the filtered fuel pool table 100. The delete tool 180 provides the user with the function of deleting fuel bundles from either the filtered or reload fuel pool tables 100 and 200. The user may select one or more fuel bundles in one of the tables, and click the delete tool to delete the selected fuel bundles from the table. Next, the loading of fresh bundles into the template will be described. FIG. 5 illustrates an exemplary embodiment of a load fresh window displayed by the loading map editor. The window is divided into two parts: a fresh bundle types table 300 and a fresh bundle pool table 400. The fresh bundle types table 300 lists the available fresh fuel bundle types. As shown in FIG. 5, the fresh bundle types table 300 lists each fresh fuel bundle type by its bundle name. The bundle name is a character string identifier used to identify the fuel bundle product line as well as nuclear characteristics, such as uranium and gadolinia loading. The fresh fuel bundle types table 300 also lists one or more attributes of each fresh fuel bundle type listed. These attributes may include K infinity, fuel bundle product line, average uranium-235 enrichment, percent (as a function of total fuel weight) of gadolinia burnable poison contained in the fuel bundle, number of gadolinia-containing fuel rods, and number of axial zones, where an axial zone is defined by a cross-sectional slice of the bundle that is homogeneous along the axial direction. Other attributes of the fresh bundle may include parameters for predicted thermal behavior, such as R-factors and local peaking, calculated for various bundle exposure values. R-factors are used as inputs to the critical power ratio (CPR) and are determined from a weighted axial integration of fuel rod powers. Local peaking is a measure of the fuel rod peak pellet and clad temperature. The fresh bundle pool table 400 provides the same information for each fuel bundle as provided by the fresh bundle types table 300. Additionally, the fresh bundle pool table 400 indicates the type number 402 for each type of fresh bundle in the loading template and then number of fresh fuel bundles of that type in the loading template. FIG. 5 shows that the first type of fresh fuel bundle position(s) are a group of four fuel bundles, and the next type of fresh fuel bundle(s) are a group of eight fuel bundles. The fresh bundle pool table 400 is populated by moving fuel bundles from the fresh bundle types table 300 into the fresh bundle pool table 400. As further shown in FIG. 5, the load fresh window includes the same filter tool 130, move right tool 160 and delete tool 180 for aiding the user in selecting and moving fuel bundles from the fresh bundle types table 300 to the fresh bundle pool table 400 as already described above. As will be appreciated, because the attributes for the fresh fuel bundles are different than the reinserted fuel bundles the filtering characteristics may also differ accordingly. The loading map editor also provides, as shown in FIG. 5, for filtering the fresh bundle types table 300 in ascending or descending order of an attribute in the same manner that the filtered fuel pool table 100 may be sorted. The selection and moving process for fresh fuel bundles does differ from the process for moving burnt fuel because the destination of the fuel must be chosen in the grouped fresh fuel bundle pool table 400 located on the right side of the fresh bundle types table 300. Namely, after a user selects the fresh bundle type from the fresh bundle types table 300, the user then selects one or more fuel bundle positions in the fresh fuel bundle pool table 400. By selecting the move right tool 160, the selected fuel bundle positions in the fresh fuel bundle pool table 400 are populated with the selected fresh bundle type. Alternatively, the user may click and drag the bundle type into the fresh fuel bundle pool table 400. Unlike with the filtered fuel pool table 100, the fresh fuel types are not removed from the fresh bundle types table 300 but are, instead, copied as fuel bundles into the fresh bundle pool table 400. Once the reinserted and fresh fuel bundle positions 38 are filled using the tools described in detail above, the user may click on a “populate” button displayed in the loading screen to have the loading map displayed. The user may then save the created loading map by using the “Save” or “Save As” options in the file menu 30. Having created the loading map, the user may then perform simulations on reactor core performance, etc. using the loading map created according to the methodologies of the present invention. By allowing the user to draw on the resources of the fuel pool(s), the present invention provides for greater flexibility in the creation of the loading map and may also reduce the overall cost in loading a nuclear reactor core. The invention 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 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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abstract | A device for removing shielding balls from calandria of a heavy water reactor is provided. The device includes a head for moving the shielding balls positioned inside of an end shield of the calandria to an outside of the end shield; and a mover for moving the head to the end shield of the calandria. The head includes a head body, an opening former installed on the head body and configured to form an opening in the end shield, and a gate installed on the head body and configured to control an amount of the shielding balls discharged to the outside through the opening. |
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041586000 | abstract | Apparatus for handling control rod drives (CRD's) attached by detachable fixing means to housings mounted in a reactor pressure vessel and each coupled to one of control rods inserted in the reactor pressure vessel. The apparatus for handling the CRD's comprise cylindrical housing means, uncoupling means mounted in the housing means for uncoupling each of the control rods from the respective CRD, means mounted on the housing means for effecting attaching and detaching of the fixing means, means for supporting the housing means, and means for moving the support means longitudinally of the CRD. |
claims | 1. A method comprising:providing a target mask for use in integrated circuit photolithography, wherein the target mask is nonpixelated, quantized tone, and in a spatial domain;performing a frequency domain transformation on the target mask to obtain a first mask, wherein the first mask is nonpixelated, continuous tone, and in a frequency domain;computing a first cost function for a first mask to obtain a first value;altering the first mask in a frequency domain to obtain a second mask, wherein the second mask is nonpixelated, continuous tone, and in the frequency domain;computing the first cost function for the second mask to obtain a second value; andrepeating the altering the first mask and the computing the first cost function for the second mask until the second value is less than the first value,wherein the performing a frequency domain transformation, computing a first cost function, and altering the first mask are performed using at least one electronic processor. 2. The method of claim 1 wherein the frequency domain transformation uses a Fourier basis. 3. The method of claim 1 wherein the frequency domain transformation uses a wavelet basis. 4. The method of claim 1 wherein the first mask comprises frequencies up to a multiple of a cutoff frequency. 5. The method of claim 1 comprising:performing an inverse frequency domain transformation on the second mask to transform the second mask to a first finalized mask, wherein the finalized mask is nonpixelated, continuous tone, and in the spatial domain; andquantizing the first finalized mask to obtain a second finalized mask, wherein the second finalized mask is quantized tone and in the spatial domain. 6. The method of claim 5 wherein the inverse frequency domain transformation uses a Fourier basis. 7. The method of claim 5 wherein the inverse frequency domain transformation uses a wavelet basis. 8. The method of claim 5 comprising:using the second finalized mask, generating subresolution assist features to obtain a third finalized mask. 9. The method of claim 8 wherein a simplified resist model is used for the first cost function. 10. The method of claim 8 wherein the subresolution assist features are generated according to preset rules. 11. The method of claim 1 wherein the first cost function comprises a term representative of a measure of variation with exposure. 12. The method of claim 1 wherein the first cost function comprises a term representative of a measure of variation with depth of focus. 13. The method of claim 1 wherein the first cost function comprises a term for weighting different portions of a printed image. 14. The method of claim 1 wherein the first cost function comprises terms representative of a measure of variation with exposure, representative of a measure of variation with depth of focus, and for weighting different portions of a printed image. 15. The method of claim 1 wherein the computing the first cost function for the second mask to obtain a second value comprises:clamping mask tone to minimum and maximum allowed mask values. 16. The method of claim 1 comprising:partitioning the target mask into a plurality of regions, each region comprising geometric shapes. 17. The method of claim 1 wherein the computing the first cost function for the second mask to obtain a second value comprises:adjusting high frequency coefficients in the first cost function while holding low frequency coefficients constant. 18. The method of claim 1 comprising:generating a second cost function by applying a quantization function to the second mask, wherein the quantization function regularizes a mask value of the second mask;computing the second cost function for the second mask to obtain a third value;altering the second mask in the frequency domain to obtain a third mask, wherein the third mask is nonpixelated, continuous tone, and in the frequency domain;computing the second cost function for the second and third masks to obtain a fourth value; andrepeating the altering the second mask and the computing the second cost function for the second and third masks until the fourth value is less than the third value. 19. The method of claim 18 comprising:performing an inverse frequency domain transformation on the third mask to transform the second mask to a first finalized mask, wherein the finalized mask is nonpixelated, continuous tone, and in the spatial domain; andquantizing the first finalized mask to obtain a second finalized mask, wherein the second finalized mask is quantized tone and in the spatial domain. 20. The method of claim 18 wherein the second cost function is different from the first cost function. |
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claims | 1. A method of determining the acceptability of the placement of a new spent nuclear fuel assembly within a spent fuel storage container wherein there are X number of spent fuel assemblies surrounding the placement of the new spent fuel assembly and X+1 is a number of the new spent fuel assembly plus the X number of spent fuel assemblies surrounding the placement of the new spent fuel assembly, comprising the steps of:generating a series of curves made up of a number of curves of enrichment vs. current burnup, each curve representing a different number of the X+1 spent fuel assemblies that have an adverse axial burnup distribution commonly used in the industry;verifying the actual axial burnup distribution of each of the X+I fuel assemblies;noting how many of the X+I spent fuel assemblies have the adverse axial burnup distribution;identifying which of the number of curves applies to the new spent fuel assembly from the number of spent fuel assemblies noted as having the adverse axial burnup distribution;finding a point on the graph on which the curve that applies to the new spent fuel assembly is plotted that corresponds to a current burnup and initial enrichment for the new spent fuel assembly; anddetermining whether the point on the graph is above the applicable curve, wherein if the point on the graph is above the applicable curve the placement of the new spent fuel assembly is acceptable. 2. The method of claim 1 wherein X is equal to three. 3. The method of claim 2 wherein the number of curves in the series of curves amounts to five curves. 4. The method of claim 1 wherein the number of curves in the series of curves equals X+2. 5. The method of claim 4 wherein the series of X+2 curves comprises five spaced curves one above another with the upper most curve representing all of the fuel assemblies having the adverse axial burnup distribution and the lower most curve representing none of the X+1 spent fuel assemblies having the adverse axial burnup distribution with each of the curves in between the upper most curve and the lower most curve representing a different number of the X+1 fuel assemblies having the adverse axial burnup, in descending order of the number of the X+1 fuel assemblies having the adverse axial burnup. |
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summary | ||
059189118 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Although the preferred embodiments illustrated in the drawings are described below in connection with replacement of a welded nozzle in the pressurizer vessel and reactor coolant piping of a nuclear power facility, the invention is not limited to that and encompasses installation of nozzles in other vessels and piping and the repair of existing nozzles. The invention further encompasses the initial installation of nozzles in new as well as in existing vessels and piping. Further, the invention inherently encompasses replacement and repair of nozzles that were previously replaced or repaired. As specifically indicated or suggested herein, or as will be apparent to those of skill in the art, an attachment or sealing system of an embodiment or embodiments described herein, or a feature or features of an embodiment or embodiments described herein may be applicable to, or incorporatable in, other embodiments. The embodiments of full replacement nozzles and assemblies depicted in FIGS. 3-7 are identical to those in FIGS. 6, 7, 9, 10 and 12, respectively, of the prior application. In those embodiments, the original welded nozzle 26 (FIG. 2) has been completely removed, leaving the J-groove weld. 34 and the cladding 29 surrounding the bore 30 for the nozzle substantially in tact. The old nozzle may be machined or drilled out; therefore destroying the nozzle. In the embodiments depicted in FIGS. 3-7 herein, and in the prior application, the replacement nozzle may be clamped, bolted, flanged, or interference fitted in the bore or hole of the vessel. Full nozzle replacement assembly 61 depicted in FIG. 3 (FIG. 6 in the prior application), employs a compressive loading system for mechanically attaching a nozzle body 62 to vessel 20. The compressive loading of the nozzle body 62 is accomplished by bolting a cylindrical sleeve 68 to the bore against the nozzle body 62, which loads the nozzle body in compression within the bore. Here, the nozzle assembly 61 includes the nozzle body 62 and the cylindrical sleeve 68. Loading the nozzle body in compression makes it less susceptible to PWSCC. If a crack developed in the nozzle body due to PWSCC (or any other mechanism), since the nozzle body is loaded in compression, the crack will not propagate, or at least is less likely to propagate. This principle is applied in other mechanical nozzle replacements discussed below. The nozzle body 62 has a tapered portion 63 which may be a full or partial length taper ending in a smaller diameter end 64 projecting from the interior entrance of the bore 60 in the vessel 20 and a larger diameter end 65 within bore 60. The nozzle body 62 also includes a tubular portion 66 projecting from the exterior of the vessel 20. The diameter of the tubular portion 66 is smaller than the larger diameter end 65 of the tapered portion 63, and a flange 67 is formed where the diameter of the bore changes from the smaller to the larger diameter. The nozzle assembly 61 also includes an externally threaded cylindrical sleeve 68, and the bore 60 includes a threaded cylindrical portion 69 and a tapered portion 70. The sleeve 68 has wrenching flats 59 on the exterior circumference thereof for tightening the sleeve in the bore. Tightening the sleeve 68 in the threaded bore portion 69 against the flange 65 of the nozzle body forces the tapered portion 63 of the nozzle body into a compressive mechanical engagement with the tapered portion 70 of the bore 60 to mechanically attach the nozzle body 62 to the vessel 20. A mechanical seal is obtained between the contacting surfaces of the tapered nozzle portion 63 and the tapered bore portion 70 by engagement of the two surfaces, which are polished as described above and in the prior application. Sleeve 68 is tightened sufficiently to ensure that the seal is obtained. Instead of polished surfaces, sealing material as described in the prior application and herein may be positioned within bore portion 70 between the bore wall and the exterior of the nozzle body 63 to create the seal. If necessary, a spring washer as described in the prior application or a Belville washer may be employed between nozzle sleeve 68 and the flange 65 of the nozzle body 62 to live load the seal, i.e., maintain a resilient compressive loading on the seal which self compensates over time due to seal shrinkage or other factors. The seals in most of the embodiments described herein may similarly be live-loaded by a spring or washer, and specific reference thereto is not made in every such embodiment The sealing material may comprise gasket material or packing material, for example, Grafoil seal material, Grafoil seal rings, packing glands etc., as is known in the art. Typically, rings and packing materials may be employed where the seal is radially restrained, e.g., the sealing material 56 in FIG. 8 between the bore shoulder 156 and the nozzle flange 159; and gasket material may be employed elsewhere, e.g., the sealing material 56 in FIG. 8 between the end 162 of the nozzle stub 152 and the end 73e of the nozzle 71e. The packing material and packing gland may be of Grafoil, e.g., a Grafoil seal ring as disclosed, for example, in U.S. Pat. No. 4,826,217 (cited above). FIG. 4 (FIG. 7 in the prior application) illustrates a replacement nozzle 71 bolted to the vessel. Full nozzle 71 is similar to fill nozzle replacement assembly 61, but is one-piece, and includes a tapered nozzle body 72 having a smaller diameter end 73 at the interior entrance of the tapered bore 60 in the vessel 20, a tubular end 75 projecting from the exterior of the vessel 20 and a larger diameter, tubular threaded portion 76. The diameter of the nozzle body 72 increases from the interior end 73 to the threaded portion 76. (Reference numeral 60 indicates a bore with a full or partial taper in it and numeral 30 indicates a cylindrical bore of a single diameter or a plurality of diameters.) The diameter of bore 60 similarly increases from the interior end of the bore to a cylindrical threaded portion 77 at the exterior end of the bore. Wrenching flats 59 are provided on nozzle body 72 adjacent the threaded portion 76 for tightening the nozzle into the threaded portion 77 of the bore 60. The nozzle 71 is structurally attached to the vessel 20 by tightening the nozzle into the bore, which forces the tapered portion of the nozzle body 72 into compressive mechanical engagement with the tapered bore 60 to mechanically bolt the nozzle body 72 to the vessel 20. A mechanical seal is obtained between the contacting surfaces of the tapered nozzle portion 72 and the tapered bore 60 by engagement of the two surfaces, as described above in connection with FIG. 3 and in the prior application. Referring to FIG. 5 (FIG. 10 in the prior application), full nozzle replacement assembly 71c is bolted to vessel 20 by an exterior flange 86c and bolts 88c. Both the nozzle body 72c and the bore 30 are cylindrical, and the mechanical attachment of the nozzle assembly 71c to the vessel is achieved by bolting the flange 86c directly against the exterior wall of the vessel. The flange 86c is a separate piece from the nozzle body 72c and may be attached to the nozzle body in any suitable manner, e.g., by a weld 87 (which should be stressed relieved prior to installation). However, the flange 86c and nozzle body 72c may be formed as one piece, as shown in FIG. 11 of the prior application. The flange 86c is contoured to follow the contour of the exterior wall of the vessel 20 against which it bears. Spacers 91 are provided between the heads of bolts 88c and the flange 86c. A thin corrosion resistant sleeve 89, e.g., made of Alloy 690, is shrink fitted or rolled into bore 30 so as to mechanically seal the sleeve 89 to the bore 30. Sealing material 56 between flange 86c and the exterior vessel wall provides the pressure retaining mechanical seal. Full nozzle replacement assembly 71b depicted in FIG. 6 (FIG. 9 in the prior application) is compressively mechanically attached to vessel 20 by an exterior flange 86 bolted to the vessel 20 by a plurality of bolts 88 and threaded holes 90 in the exterior wall of the vessel Exterior flange 86 is a separate piece from nozzle 71b (or in an alternate embodiment may be one-piece with or without welding to the nozzle), and is engaged with nozzle assembly 71b as follows. Nozzle assembly 71b includes a tubular end 75b of reduced diameter projecting from the vessel 20 which forms a circular shoulder or flange 92. Flange 86 includes a circular recess 94 with a central hole 96 therein. The shoulder 92 on the nozzle assembly 71b is received in the recess 94 in the flange 86 with the tubular portion 75b of the nozzle body 72b passing through the central hole 96 in the flange's recess 94. Bore 60 and nozzle body 72b are tapered generally as described for nozzle 71 in FIG. 4, and tightening bolts 88 causes the flange 86 to draw the nozzle body 72b into bore 60 into compressive frictional engagement therewith. If necessary, spring washers (not shown) may be provided between the heads of bolts 88 and flange 86. A mechanical seal is obtained between the exterior of the nozzle body 72b and the walls of bore 60, as described above in connection with FIGS. 3 and 4. Full nozzle replacement 110 shown in FIG. 7 (FIG. 12 in the prior application) is frictionally attached to the vessel 20 and mechanically sealed by an interference fit of the nozzle 110 in the bore 30. Nozzle 110 is tubular and bore 30 is cylindrical. At equal temperatures of the nozzle 110 and the vessel portion 20 surrounding the bore 30, the diameter of the nozzle is slightly larger than the diameter of the vessel bore. The nozzle 110 is inserted into the bore by creating a substantial temperature gradient between the two so that the diameter of the nozzle is reduced or the diameter of the bore is increased, or both. The temperature of the vessel 20 surrounding the bore 30 is increased to expand the diameter of the bore, or the nozzle 110 is cooled to reduce its diameter, or both. After the nozzle 110 has been inserted into the bore 30, the temperature gradient is reduced so that the nozzle 110 frictionally engages the wall of the bore 30 in an interference fit to both mechanically attach the nozzle and mechanically seal its exterior with the wall of the bore at the operating temperatures of interest. The exterior of the nozzle and the bore are polished to assist in creating a seal therebetween. If desired, a mechanical seal or seals, in addition to the mechanical seal obtained from the interference fit and polished surfaces, may be provided as discussed herein and in the prior application. FIGS. 8-20 illustrate embodiments of full and partial nozzle replacement assemblies which are improvements over the embodiments illustrated in the prior application (including FIGS. 3-7 of this application which are identical to FIGS. 6, 7, 9, 10 and 12, respectively, of the prior application), and which introduce features not disclosed in the prior application as well as applying features disclosed in the prior application for improved full nozzle replacement and partial nozzle replacement, which may also be applied to nozzle repair. As indicated above, individual features, or combinations of features, from the embodiments of FIGS. 8-20 may be applied to other embodiments, or combined in various combinations. FIGS. 8, 9, 10, 11 and 12 show variations of the full nozzle replacement 71 of FIG. 4 for partial nozzle replacement assemblies 150, 170, 180, 190 and 190a that include the part (or stub) 152 of an existing nozzle adjacent the interior of the vessel 20 and a partial replacement nozzle 71e, 71f, 71g, 71h and 71i, respectively. For FIG. 10A, the partial nozzle replacement assembly 185 is a variation of the full nozzle replacement of FIG. 7, which includes an existing nozzle stub 152 and a partial replacement nozzle 110a. In the embodiments of FIGS. 8, 9, 10, 10A, 11 and 12, a portion of the existing nozzle is removed leaving the nozzle stub 152 extending into or flush to the bore from the interior of the vessel. Both the nozzle stub 152 and its J-groove weld 34 are left in tact and substantially undisturbed. Removal may be accomplished as described above for removal of the entire nozzle, and in other ways which are known to those of skill in this art. The vessel bore may be altered only to the extent of threading it adjacent the exterior of the vessel, and/or it may be enlarged or tapered as described below. Referring to FIG. 8, a portion 77e of the bore 30 adjacent the exterior of the vessel is enlarged in diameter and threaded. A shoulder 156 is formed in the bore 30 at the interface of the larger diameter portion 77e and a smaller diameter portion 30e. The replacement nozzle 71e likewise includes a smaller diameter portion 158 sized to be received in the smaller diameter bore portion 30e and a threaded larger diameter portion 76e to be threadedly received in the larger diameter bore portion 77e, and a flange 159 at the interface of the smaller and larger diameter portions. Sealing material 56 (e.g., packing material) is positioned in the bore between the bore shoulder 156 and the nozzle flange 159. The replacement nozzle 71e is tightened to the bore (using the wrenching flats 59) to compressively mechanically and structurally attach the replacement nozzle 71e to the vessel 20. Also, tightening the replacement nozzle 71e in the bore compresses the sealing material 56 between the bore shoulder 159 and the nozzle flange 157 to mechanically seal the replacement nozzle 71e to the vessel. Another mechanical seal may be provided in the bore between the interior end 73e of the replacement nozzle 71e and the end 162 of the nozzle stub 152 by sealing material 56 (e.g., gasket or packing material) which also is compressed by tightening the nozzle to the bore. Where a mechanical seal is provided between the annular edges of adjacent nozzle sections, such as the end 162 of nozzle stub 152 and the interior end 73e of replacement nozzle 71e, the edge is chambered at 163 for the new and existing nozzle (not shown) to cause the seal to compress radially into the nozzle and prevent the packing material from extruding through the gapped region between the two nozzles. Achieving the seal between nozzle stub 152 and replacement nozzle 71e requires some axial loading on the existing nozzle stub 152. Alternatively, a gap may be left between the replacement nozzle 71 and the existing nozzle stub 152 so that the existing nozzle stub 152 is not subjected to any axial or radial loading which might otherwise stress the J-groove weld 34. The length of tubular section 158 is selected so that some thread is available in the vessel bore after initial installation to further tighten the nozzle into the bore during service to compensate for shrinkage of the sealing material 56. The bore and nozzle or external compression sleeve are similarly threaded in FIGS. 3, 9, 10, 11, 12-24. The mechanical seals formed at bore shoulder 156 and the nozzle flange 159, and at the end 162 of nozzle stub 152 and the interior end 73e of replacement nozzle 71e are adjustably-loaded, i.e., the degree of compression of the respective sealing material 56 may be adjusted during service, for example, to compensate for seal shrinkage. In most the embodiments disclosed herein, the mechanical seals are similarly adjustably-loaded, and specific reference thereto will not be made in each embodiment. The partial nozzle replacement assembly 170 depicted in FIG. 9 differs from the partial nozzle replacement assembly 150 in FIG. 8 in that the replacement nozzle 7 If of FIG. 9 seals with the existing nozzle stub 152 around the circumference thereof without axially loading the nozzle stub 152. In this embodiment, the bore 30 from a point overlapping the end 162 of the nozzle stub 152 to the exterior of the vessel is enlarged in diameter to define the enlarged diameter portion 77f and a shoulder 156 in the portion of the enlarged bore overlapping the end 162 of the nozzle stud 152. The enlarged diameter bore portion 77f is threaded adjacent the exterior of the vessel. The replacement nozzle 71f is sized to be received in the larger diameter bore portion 77f and includes a threaded portion 76f threadedly received in the threaded portion of larger diameter bore portion 77f The ID of the replacement nozzle 71f at the interior end 174 thereof is enlarged to receive therein the end 162 of the nozzle stub 152. Sealing material 56 (e.g., packing or gland material) is positioned between the bore shoulder 156 and the end 174 of the replacement nozzle 71f surrounding the end of the nozzle stub 152. The replacement nozzle 71f is tightened to the bore to compressively, mechanically and structurally attach the replacement nozzle 71f to the vessel 20. Also, tightening the replacement nozzle 71f in the bore compresses the sealing material 56 against the shoulder 156 and against the outer circumference of the end 162 of the nozzle stub 152 to mechanically seal the replacement nozzle 71f to the vessel. Sealing material 56 (e.g. gasket material) may also be compressed between the end 162 of the existing nozzle stub 152 and shoulder 175 where the nozzle ID changes diameter however, this would axially load the nozzle stub. The partial nozzle replacement assembly 180 depicted in FIG. 10 differs from the partial nozzle replacement assembly 150 in FIG. 8 in that the replacement nozzle 71g seals axially only with the end 162 of existing nozzle stub 152. The bore 30 is threaded adjacent the exterior of the vessel, but unlike the nozzle replacement assembly 150, the bore 30 extending from the end of the nozzle stub 152 to the exterior of the vessel is not enlarged in diameter, and no shoulder is formed in the bore sufficient to insert a seal. The replacement nozzle 71g is sized to be received in the bore 30 and includes a threaded portion 76g threadedly received in the threaded portion of the bore. Tightening the replacement nozzle 71g to the bore mechanically bolts and structurally attaches the replacement nozzle 71g to the vessel 20. A mechanical seal is provided in the bore between the interior end of the replacement nozzle 71f and the end 162 of the nozzle stub 152 by sealing material 56 (e.g. gasket material) which is compressed therebetween when the nozzle is tightened to the bore. The nozzle replacement 170 (FIG. 9) without the sealing material 56 between the end 162 of nozzle stub 152 and the shoulder 175 of the nozzle 71f is presently preferred over this embodiment and the embodiment of FIG. 8 because those embodiments axially load the existing nozzle stub 152. The partial nozzle replacement assembly 185 depicted in FIG. 10A applies a sleeve 110a in the bore 30, as described for sleeve 110 of FIG. 7 to mechanically attach and seal the sleeve 110a in the bore 30. Additionally, sealing material 56 is compressed between the interior end of the sleeve 100a and the end of 162 of the nozzle stub 152 when the sleeve 110a is attached by hydraulically or other means forcing the nozzle into the vessel. FIG. 11 shows another variation of the full replacement nozzle 71 of FIG. 4 for a partial nozzle replacement assembly 190 that includes an existing nozzle stub 152 adjacent the interior of the vessel 20 and a replacement nozzle 71h. In this embodiment, the bore 60 from which the existing nozzle has been removed is enlarged in three sections. The bore is tapered in a first section 74h immediately adjacent the nozzle stub 152, increasing in diameter as the bore progresses towards the exterior of the vessel. In a second section 182 adjacent the tapered first section 74h, the bore is cylindrical having the largest diameter of the tapered section 74h. In a third section 77h, the bore has a diameter further enlarged from that of bore section 182 and is threaded. A shoulder 156 is formed in the bore 60 at the interface of the two larger diameter sections 182, 77h. The replacement nozzle 71h likewise includes a tapered portion 72h sized to be received in the tapered section 74h of the bore, a smaller cylindrical portion 184 sized to be received in the cylindrical section 182 of the bore, and a threaded larger diameter portion 76h sized to be threadedly received in the threaded larger diameter section 77h. The replacement nozzle 71h is structurally attached to the vessel 20 by tightening the nozzle into the bore, which forces the tapered portion 72h of the nozzle into mechanical engagement with the tapered section 74h of the bore to mechanically bolt the replacement nozzle 71h to the vessel 20. A mechanical seal is obtained between the contacting surfaces of the tapered nozzle portion 72h and the tapered bore section 74h by engagement of the two surfaces, as described above and in the prior application. A gap 186 is left between the end 162 of the nozzle stub 152 and the interior end of the replacement nozzle 71h. However, another mechanical seal may be provided in gap 186 as shown in FIG. 12 (described below). The partial nozzle replacement assembly 190a shown in FIG. 12 is identical to partial nozzle replacement assembly 190 except that no gap is left between the tapered nozzle portion 72h and the end 162 of the nozzle stub 152, and sealing material 56 (e.g., gasket or packing material) is positioned thereat and compressed when the replacement nozzle 71i is tightened to the bore to form another mechanical seal within the bore. In the embodiments of FIGS. 10, 11, and 12, a shoulder similar to shoulder 156 in FIG. 8 may be provided in the gap region 188 in the bore, and a flange may be provided on the nozzle similar to flange 159 in FIG. 8, and another mechanical seal may be provided in this gap, as shown in FIG. 8. (This additional mechanical seal may also be provided in the embodiments of FIGS. 13, 14, 15 and 16.) FIGS. 13-16 respectively illustrate full nozzle replacement assemblies 61a, 61b, 232 and 242 in which the existing nozzle has been completely removed. The nozzle replacement assemblies 61a, 61b and 232 in the embodiments of FIGS. 1315, respectively, employ full nozzles which extend to the interior of the vessel while the replacement nozzle assembly 242 in the embodiment of FIG. 16 employs a partial nozzle that terminates within the bore of the vessel sealed with respect to a shrink fit sleeve 244 that extends to the interior of the vessel and packing material 56. The full nozzle replacement assemblies 61a, 61b, 232 and 242 all utilize a nozzle assembly of at least two parts excluding the seals themselves which includes a drive or compression sleeve threaded to the bore of the vessel like the nozzle assembly 61 shown in FIG. 3. Referring to FIGS. 13-16, the procedure for the installing full nozzle replacement assemblies 61a, 61b, 232 and 242 removes the existing nozzle entirely, and the vessel bore is altered to provide tapered and/or larger diameter sections, and/or threads in the bore adjacent the exterior of the vessel. Full nozzle replacement assembly 61a depicted in FIG. 13, like the nozzle assembly 61 of FIG. 3, employs a compressive loading system for mechanically attaching the nozzle to vessel 20, and includes a nozzle body 62a having a tapered portion 63a having a smaller diameter end 64a projecting from the interior entrance of the bore 60 in the vessel 20 and a larger diameter end 65a within bore 60. The nozzle body 62a also includes a tubular portion 66a projecting from the exterior of the vessel 20. The diameter of the tubular portion 66a is smaller than the larger diameter end 65a of the tapered portion 63a, and a flange 67a is formed where the diameter of the nozzle changes from the smaller to larger diameter. The nozzle replacement assembly 61a also includes an externally threaded cylindrical drive sleeve 68a, and the bore 60, includes a threaded cylindrical portion 69a and a tapered portion 70a. Tightening the sleeve 68a (using wrenching flats thereof, not shown) into the threaded bore portion 69a against the flange 67a of the nozzle body forces the tapered portion 63a of the nozzle body into compressive mechanical engagement with the tapered portion 70a of the bore 60 to mechanically attach the nozzle body 62a to the vessel 20. A mechanical seal is obtained between the contacting surfaces of the tapered nozzle portion 63a and the tapered bore portion 70a by engagement of the two surfaces, as described above. Sleeve 68a is tightened sufficiently to ensure that the seal is obtained. The full nozzle replacement assembly 61a described so far with respect to FIG. 13 is basically the same as for the full nozzle assembly 61 of FIG. 3, except for the lengths of the tapers in the nozzle body 62a and the bore 60a the length of the taper in the bore and nozzle may vary from embodiment to embodiment. For example, the taper may be less than 1/2" in length. In addition, nozzle replacement assembly 61a provides packing material for a secondary seal at the interface of the tapered and cylindrical sections of the bore (i.e., at the flange 67a of the nozzle), and includes an anti-rotation device 206 described in detail below. Still referring to FIG. 13, a thrust bearing 202 is positioned between the flange 67a of the nozzle body 62a and the interior end of the nozzle drive sleeve 68a, which facilitates tightening the sleeve 68a in the bore 60a. Sealing material 56 (e.g., packing or gland material) is also positioned between the thrust bearing 202 and the interior end of the drive sleeve 68a. The anti-rotation device 206 is implemented in the full nozzle replacement assembly 61a of FIG. 13 as an optional feature by an axial slot defined by adjacent slot portions 207a, 207b in the bore 60 and in the tapered portion 63a of the nozzle body 62a, respectively, and key stock 208 projecting into engagement with both the axial slot portions 207a, and 207b in the bore and the nozzle. The full nozzle replacement assembly 61b shown in FIG. 14, like the full nozzle replacement assembly 61a of FIG. 13, is compressively mechanically attached to the vessel and utilizes a tapered bore/tapered nozzle, but does so without actually tapering the vessel bore. Full nozzle replacement assembly 61b similar to nozzle assembly 61a of FIG. 13, includes a nozzle body 62b and a threaded drive sleeve 68b. The nozzle body 62b includes a tapered portion 63b, a cylindrical portion 222 and a flange 224 at the interface of the cylindrical and tapered portions of the nozzle body. The bore 30 for full nozzle replacement assembly 220 does not include a tapered portion, and instead includes a smaller diameter cylindrical portion 226 and a larger diameter cylindrical portion 69b, which form a shoulder 67b at the interface thereof. Instead of a tapered bore section, a sleeve 204 having a cylindrical OD and a tapered ID is positioned in the smaller diameter bore section 226 sized to engage the tapered nozzle portion 63b when the drive sleeve 68b is tightened to the bore. The sleeve 204 has a flanged interior end 223 which engages the flange 67b in the bore to initially hold and position the sleeve 204 in the bore. Sealing material 56 (e.g., packing or gland material) is positioned between the nozzle flange 224 and the interior end of the drive sleeve 68b which is compressed when the drive sleeve is tightened to the bore. Tightening the drive sleeve 68b to the bore compressively loads the nozzle body 62b to the vessel. Full nozzle replacement assembly 61b provides a seal at the interface of the different diameter sections of the bore 226, in addition to the mechanical seal between the tapered surfaces of the nozzle and the sleeve 204. Alternatively, as shown in FIG. 16, the flange 224 on the nozzle body 62b may be omitted, the tubular section 222 of the nozzle body adjacent the exterior end of the drive sleeve 68b may be threaded along with drive sleeve 68b, to mechanically attach the nozzle body to the vessel and compress the sealing material 56 between the end of the drive sleeve and the flanged portion 223 of the sleeve 204, and lock nuts may be tightened against the drive sleeve to prevent the drive sleeve from loosening in the event the vessel or nozzle experiences vibration due to fluid-induced vibration or some other mechanism. As discussed above full nozzle replacement assembly 61b provides a seal at the interface of the different diameter sections of the bore, in addition to the mechanical seal between the tapered surfaces of the nozzle and the sleeve 204. This arrangement allows separate loadings for the two sealing surfaces, and can be applied to other embodiments as well. Referring to FIG. 15, full nozzle replacement assembly 232 employs a cylindrical bore 30 threaded adjacent the exterior of the vessel 20 and a full replacement nozzle that is mechanically attached to the vessel 20 by a drive or attaching sleeve 68b threaded to the vessel. Full nozzle replacement assembly 232 includes a nozzle body 233, and the drive sleeve 68b and a seal alignment spacer sleeve 235. The nozzle body 233 has tubular section 234 ending in a flanged end 236 projecting into the interior of the vessel 20. Spacer sleeve 235 and the end 236 of the nozzle body in the interior of the vessel are flanged at the angle of the ID of the vessel. Drive sleeve 68b, is threaded adjacent its exterior end and is sized to be threadedly received in the threaded portion of the bore 30 structurally attaching the sleeve and nozzle assemble to the vessel. The spacer sleeve 235 is positioned in the bore between the drive sleeve 68b and the flanged end 236 of the nozzle body 233. Sealing 56a (e.g., a first packing seal ring or gland is positioned between sleeves 68b and 235 and sealing material 56b (e.g., a second packing seal ring or gland) is positioned between sleeve 235 and the flanged end 236 of the nozzle body 233. The nozzle body 233 is positioned in the bore 60a, followed by sealing material 56b, the inner sleeve 235 and sealing material 56a. The drive sleeve 68b is threaded to the bore 30, and a nut 238 is tightened to a threaded section 229 of the tubular section 234 of the nozzle body 233 to mechanically compress sealing materials 56a and 56. Referring to FIG. 16, full nozzle replacement assembly 242 (referred to as a full nozzle since all of the existing nozzle is removed) employs a bore 30 similar to the one shown in FIG. 14, which includes a cylindrical section 226 of smaller diameter opening to the vessel interior and a cylindrical section 60b of larger diameter opening to the exterior of the vessel which is threaded adjacent the opening to the exterior of the vessel. The cylindrical bore sections 222 and 60b of different diameter form a shoulder 67b within the bore at the interface of the two sections. The full nozzle replacement assembly 242 includes a nozzle body 243 having a tubular section 222 and a drive sleeve 68b. The tubular section 222 terminates within the smaller diameter section 226 of the bore and projections therefrom in an exteriorly threaded section 229. The drive sleeve 68b is exteriorly threaded adjacent the interior end of the sleeve and sized to be threadedly received in bore section 60b. The drive sleeve 68b is interiorly threaded adjacent the end projecting from the vessel. The full nozzle replacement assembly 242 also includes a sleeve 244 shrink fitted to the cylindrical nozzle section 222 and extending to the shoulder 67b in the bore as described above in connection with nozzle 110 in FIG. 7, with the end 246 of the nozzle body 243 received in the end of the sleeve 244. Sealing material 56 (e.g., a packing seal ring or gland) is positioned in the bore between the shoulder 67b and the flange 204 of the nozzle body. After the sleeve 244 has been shrink fitted to the bore 30, the sealing material 56 is positioned at flange surface 67b and the drive sleeve 68b is tightened to the bore to compress the sealing material 56. The threaded section 229 of the nozzle body 243 is threaded to the drive sleeve 68b to structurally attach the nozzle body 243 to drive sleeve 68b which is structurally attached to the vessel with threads 69b. The locking nuts 228 prevent the nozzle from loosening. In the embodiments depicted in FIGS. 3-4, 6-16 and 21, mechanical seals are provided solely within the bore. In FIGS. 5, 17-20, and 22-25, mechanical seals are or may be provided also at the vessel OD. FIG. 17 shows a full nozzle replacement assembly 250 which is a variation of the full nozzle replacement assembly 61b shown in FIG. 14 and incorporates an anti-rotation device and two leak paths features. The function of the leak paths is to channel any reactor coolant leakage which may occur through a corrosion resistant leak path out past the exterior surface of the vessel in a visible manner, e.g., to the oxygenated environment where the coolant can flash steam without eroding the vessel. Like full nozzle replacement assembly 220 of FIG. 14, the bore 30 of full nozzle replacement assembly 250 includes a smaller diameter cylindrical portion 226 and a larger diameter cylindrical portion 69b, which form a shoulder 65b at the interface thereof. The full nozzle replacement assembly 250 is similar to that of nozzle assembly 61b of FIG. 14 in that nozzle assembly 250 includes a nozzle body 62b and a threaded drive sleeve 68b, with the nozzle body 62b including a tapered portion 63b, a cylindrical portion 222 and a flange 224 at the interface of the cylindrical and tapered portions of the nozzle body. As in the full nozzle replacement assembly 61b of FIG. 14, instead of a tapered bore section, the full nozzle replacement assembly 250 includes a sleeve 204 having a cylindrical OD and a tapered ID positioned in the smaller diameter bore section 226 sized to engage the tapered nozzle portion 63b when the drive sleeve 68b is tightened to the bore. Full nozzle replacement assembly 250 (FIG. 17) includes the following additional elements: a first anti-rotation device 255 for preventing rotation of the sleeve 204 in the bore; a second anti-rotation device 258 for preventing rotation of the nozzle body 62b in the bore, circumferential vapor collection groove 282, first packing seal ring or gland 56a, flat washer 260, Belville washer 262; lantern ring 264; two concentric packing seal rings or glands 56b, c; axial vapor channel slot 280, gasket 56; channeling sleeve 268; channeling passages 270, and jam nut 271. Full nozzle replacement assembly 250 provides a primary seal at the tapered sections 226, 63b of the sleeve 204 and the nozzle body 62b, a secondary seal at the packing seal ring 56a, and a tertiary seal at the packing seal rings 56b, c. The anti-rotation device 255 (FIG. 17) comprises registered slot portions 271a, b, in the vessel and the flange 223 of the sleeve 204 and a key stock 208 in the registered slot portions. The anti-rotation device 255 prevents the sleeve 226 from rotating relative to the bore. The anti-rotation device 258 comprises registered slot portions 271a, b in the sleeve flange 223 and on the nozzle body flange 224, and a cylindrical key stock 208 in the registered slot portions 271a, b. The anti-rotation device 258 prevents the nozzle body 62b from rotating relative to the sleeve 204, and thereby from rotating relative to the vessel 20. The slots 271a, b and the key stock 208 for anti-rotation devices 255 and 258 may be of circular or rectangular cross-section. The flat washer 260 and the Belville washer 262 assist in securely compressively loading the nozzle body 62b in the bore, and line-load the seals. As mentioned, the packing seal ring 56a provides a secondary seal to prevent leakage. The packing seal rings 56b, c. provide a tertiary seal forming a primary leak path with the inclusion of lantern ring 264 that axially projects between the concentric packing seal rings 56b, c and includes a series of passages 265 spaced circumferentially about the lantern ring extending axially therethrough from one end to the other. The passages 265 are positioned to extend between the spaced concentric packing seal rings 56b, c into circumferential groove 276 where the coolant collects and passes through a series of passages 270 spaced circumferentially (i.e., each having a circumferential width of substantially less than 360 degrees) extending axially from one end to the other of drive sleeve 68b and allows the coolant to escape to the atmosphere. The finction of the slot 268 in the lantern ring 264 is to collect any fluid which may leak past an assumed failed seal at the tapered nozzle body 63b and the tapered sleeve 226 interface or the bore to tapered sleeve 226 interface and the secondary packing seal ring 56a, and to channel the leaked fluid to the axial passages 265 in the lantern ring 264. Grooves 269 perpendicularly intersecting circumferential grooves 268 channel the leaked fluid to the circumferential grooves 268. The function of the slot 276 in the sleeve end 272 is to collect any fluid which passes through the passages 265 in the lantern ring 264 and to channel that fluid to the axial passages 270 in the sleeve 68b. Thus, the primary leak path directs any fluid passing between the tapered nozzle body portion 63b and the sleeve 204, and/or between the sleeve 204 and the bore, which also leaks past a failed secondary packing seal ring 56a, to the passages 265 in the lantern ring 264 and the passages 270 in the sleeve 68b, from which the leaking fluid is discharged at the end 274 of the drive sleeve 68b outside of the vessel. Although not shown, the lantern ring 264, packing seal ring 56c, and sleeve 274 could easily be modified to channel the primary leak path between sleeve 274 and nozzle body 62b. The groove 276 is of sufficient depth to allow additional compression by tightening sleeve 68b should the seal shrink with time. Elements defining the secondary leak path include: an axial slot 280 in drive sleeve 68b which the channeling sleeve 268 overlays; and a radial groove 282 extending 360.degree. in the circumference of the drive sleeve 68b closed by the channeling sleeve 268. The radial groove 282 starts in the axial channel 280 and extends in the drive sleeve 68b past the channeling sleeve 268 where it opens to the outside of the vessel. Gasket 56 prevents in leakage on to the vessel. Thus, the secondary leak path directs any fluid passing between the tapered nozzle body portion 63b and the sleeve 204, and/or between the sleeve 204 and the bore, which also leaks past a failed first packing seal ring 56a and a failed second packing seal ring 56b and/or c to the radial channel 282, which then directs the fluid to axial groove 280 from which the leaking fluid is discharged outside of the vessel in a corrosion resistant means such that leakage will not erode the vessel. FIG. 18 shows a partial nozzle replacement assembly 300 which is a variation of the full nozzle replacement assembly 250 of FIG. 17, and which incorporates a feature or features from the partial nozzle replacement assembly 170 of FIG. 9, to provide a partial nozzle replacement assembly that includes an existing nozzle stub 152 adjacent the interior of the vessel 20. In this embodiment, like nozzle replacement assembly 170 (FIG. 9), a portion of the existing nozzle is removed leaving the nozzle stub 152 extending into the bore from the interior of the vessel. Both the nozzle stub 152 and its J-groove weld 34 are left in tact and substantially undisturbed. The entire portion 77f of the bore 30 extending from the end of the nozzle stub 152 to the exterior of the vessel is enlarged in diameter, including a portion 174 overlapping the end 162 of the nozzle stub 152. A shoulder 65b is formed in the interface of the smaller and larger diameter portions. The enlarged diameter bore portion 77f is threaded adjacent the exterior of the vessel. The replacement nozzle assembly 300 includes a nozzle body 62c and a drive sleeve 68c, similar to nozzle replacement assembly 250 of FIG. 17. However, the nozzle body 62c is configured like the one piece nozzle 71f of FIG. 9 in that it overlaps the existing nozzle stub 152. The ID of the replacement nozzle body 62c at the interior end 174 thereof is enlarged to receive therein the end 162 of the nozzle stub 152, as in replacement nozzle 170 of FIG. 9. A packing seal ring 56a is positioned between the bore shoulder 67b and the end 174 of the nozzle body 62c, surrounding the end 162 of the nozzle stub 152, and another packing seal ring 56d is applied at the flanged end 174 of the nozzle body 62c. The anti-rotation device 310 of the partial nozzle replacement assembly 300, which differs from the anti-rotation devices of full nozzle replacement assembly 250, includes an axial slot 171a in the existing nozzle stub 152 and a tab 208a machined into the nozzle body 62c and radially projecting therefrom into the slot 171a. If tortional loads are high a plurality of axial slots and tabs can be used. Alternatively, the anti-rotation device 310 may be formed by axial groove portion, circular or tapered in nature, in the existing nozzle stub 152 and a similar groove in nozzle body 62c into which the key stock is inserted and held (several grooves and key stocks could be used if tortional loads demanded such). The nozzle groove is of sufficient length to allow further compression loading from time to time The nozzle replacement assembly 300 is otherwise as described for nozzle replacement assembly 250 depicted in FIG. 17, and includes the primary and secondary leak paths as described for nozzle replacement assembly 250. FIG. 19 depicts a partial nozzle replacement assembly 320 similar to the partial nozzle replacement assembly 300 shown in FIG. 18, but which bolts the nozzle sleeve 68d to the vessel exterior with a flange 87a similar to the one used to bolt nozzle assembly 71c to the vessel as shown in FIG. 5. The nozzle body 62d is inserted into the bore, and the flange 87a along with channeling sleeve 268 (fabricated and fillet welded 324 to flange) is sealed against an annular taper guide 322 and the vessel OD by sealing (e.g., gasket) material 56, and bolted to the vessel exterior with threaded bolts 88c in threaded holes 90c in the vessel. Taper guide 322 provides alignment of the nozzle to the vessel such that the nozzle may be compressed into the vessel to form a tight interference fit to resist external bending loads that may be applied to the nozzle. (Though not shown for the other embodiments this may be applied to all of the embodiments in this application.) The drive sleeve 68d is then tightened to channeling sleeve 268 to compressively load the nozzle body 62d to the vessel. The jam nut 271 is then tightened to prevent the drive sleeve 68d from loosening. This embodiment also provides the same primary and secondary leak paths and anti-rotation device as in the partial nozzle replacement assembly 300, and is otherwise the same as the partial nozzle replacement assembly 300. FIG. 20 depicts a partial nozzle replacement assembly 330 also similar to the partial nozzle replacement assembly 300 shown in FIG. 18, but which bolts the drive sleeve 68e to the vessel exterior with an external flange 86a similar to the one used to bolt nozzle assembly 71b to the vessel as shown in FIG. 6. Exterior flange 86a is a separate piece from drive sleeve 68e (or in an alternate embodiment may be one-piece with the drive sleeve), and is threaded to the drive sleeve 68e. The external flange 86a is bolted to the vessel exterior by threaded bolts 88 in threaded holes 90 with the sealing (e.g., gasket) material 56 surrounding the taper guide 332 compressed between the end of a sleeve portion 334 of the flange 86a and the exterior of the vessel. Flange 86a is bolted on in this manner such that it does not have to conform to the curvature of the vessel as does flange 87a of FIG. 19 which can be difficult for heater nozzles inserted on the hillside of the spherical shaped head at the bottom of the pressurizer vessel The drive sleeve 68e is threaded to the flange 86a to compress the packing seal rings 56a-d. This embodiment is otherwise like the partial nozzle replacement assembly 300 (FIG. 18) and provides the same primary and secondary leak paths and anti-rotation device. Although only a single key stock is shown for each anti-rotation device, depending on the tortional loads experienced by the nozzle, multiple key stocks could be used in combination. In the embodiments described above, a structurally welded and seal welded nozzle may be full or partially removed by cutting or machining operations. As mentioned, the invention is also applicable to replacement of nozzles attached and sealed in other ways, (e.g., as described in the prior art discussed above and by mechanical attachment and sealing as described herein). A full or partial nozzle replacement, similar to a nozzle repair described below, may be of the precautionary type in which the existing nozzle has not failed in any way and the full or partial nozzle replacement by mechanical means is made for precautionary purposes. Typically, a precautionary replacement will also be made when the system is shut down for other reasons, such as scheduled refueling in a nuclear facility. Further, the invention is applicable to the repair of existing nozzles which do not require removal of the existing nozzle. FIGS. 21-25 disclose repair embodiments which leave the existing nozzle within the bore in tact, and further attach the existing nozzle to the vessel and mechanically seal the existing nozzle to the vessel. A nozzle repair may be used in the event that the existing nozzle leaks or as a precaution in which the existing nozzle has not failed in any way and the mechanical sealing and further attachment of the nozzle to the vessel are made for precautionary purposes. In the embodiments of FIG. 21-25, the existing nozzle is mechanically sealed to the vessel, and may be further attached to the vessel by a fillet weld between the existing nozzle and a sleeve which is mechanically attached to the vessel. FIG. 26 provides a system applicable to the embodiments of FIGS. 21-25 which enables the drive sleeve to be torqued even after the fillet weld is applied to the existing nozzle. In the embodiments of FIGS. 21-24, the existing nozzle is cut off outside of the vessel to enable mechanical sealing to proceed, and in the embodiment of FIG. 25, the existing nozzle is left entirely in tact by use of a split seal and flange. FIGS. 21-25 apply to nozzle repair, features and techniques disclosed herein and in the prior application for full and/or partial nozzle replacement. Typically, a precautionary repair will be made when the system is shut down for other reasons, such as scheduled refueling in a nuclear facility. FIGS. 21-24 depict a nozzle repair assembly 350, 360, 370, and 380 in which an existing nozzle 352 is left entirely in tact within the bore 30 but is cut off at some point 354 outside of the vessel 20. nozzle repair assembly 350 of FIG. 21 applies a technique similar to the partial nozzle replacement assembly of FIG. 9 to nozzle repair. Referring to FIG. 21, the diameter of the bore 30 is enlarged starting at a point spaced in the bore from the J-groove weld 34 to provide an enlarged diameter bore section 77f as in FIG. 9. The bore section 77f is threaded adjacent the exterior of the vessel 20 and a sleeve 71f is threaded to the bore, as in FIG. 9. Sealing (e.g., packing) material 56 positioned between the interior end 174 of the sleeve 71f and the shoulder 156 formed where the larger diameter bore section 77f meets the smaller diameter bore 30, and is compressed when the sleeve is threaded to the bore to provide the mechanical attachment and the mechanical seal. The nozzle repair assembly 350 is the same as the partial nozzle replacement assembly of FIG. 9, except that the entire existing nozzle 352 is left within the bore and no changes are made to the existing nozzle (and bore). The existing nozzle 352 may be welded to the sleeve 71f with a fillet weld 356 to provide for further attachment of the nozzle 352 to the vessel 20, functioning as an anti-ejection device or the sleeve 71f may be modified and a clamping device 406 used as shown in FIGS. 26 and 27 (described below) to allow the sleeve to be torqued from time to time to compress the mechanical seal and prevent leakage. In the nozzle repair assemblies 360, 370 and 380 depicted in FIGS. 22-24, respectively, the techniques of the partial nozzle replacement assemblies 300, 320 and 330 of FIGS. 18-20, respectively, are applied to repairing a nozzle. The nozzle repair assemblies 360, 370 and 380 are the same as the partial nozzle replacement assemblies 300, 320 and 330 of FIGS. 18-20, except that the entire existing nozzle 352 is left within the bore. Also, the existing nozzle 352 in nozzle repair assemblies 360, 370 and 380 may be welded to the respective sleeve with a fillet weld 356 to provide for further attachment of the nozzle 352 to the vessel 20, functioning as an anti-ejection device, or the respective sleeves may be modified as shown in FIG. 26 for the reason stated above in connection with FIG. 21. The nozzle repair assemblies 360, 370 and 380 of FIGS. 22-24 utilize the leak paths of FIGS. 18-20 and other elements shown in FIGS. 18-20, respectively. Another difference between the embodiments of FIGS. 22 and 24 and those of FIGS. 18 and 20 and is that in FIGS. 22 and 24 the exterior mechanical seal between the respective sleeve or flange is made with packing material in a short enlarged bore diameter section 358 defining an annular groove between the bore and the sleeve at the OD of the vessel, while in FIGS. 18 and 20 the seal is made on the exterior surface of the vessel OD with gasket material. The nozzle repair assembly 390 depicted in FIG. 25 is a combination of the nozzle repair assembly 360 and 380 in FIGS. 22 and 24. Nozzle repair assembly 390 utilizes a flange 86c similar to the clamp device 86b of FIG. 24 and a portion 392 of a drive sleeve similar to the drive sleeve 68c of FIG. 22. In addition, the flange 86c is split and is clamped to the existing nozzle 352 by bolts 394. The split flange 86c frictionally engages the existing nozzle to mechanically attach the nozzle to the vessel. This enables the sleeve portion 392 and the clamp device 86c to be attached to the existing nozzle 352 and the vessel 20 without cutting the existing nozzle and without the need for a weld to provide additional structural support. The sleeve portion 392 mechanically seals the existing nozzle 352 in the same manner as the partial nozzle replacement assembly of FIG. 18 and the nozzle repair assembly of FIG. 22 are sealed, except that the clamp device 86c may be tightened to the vessel after initial assembly and tightening of the clamp device 86c, without breaking any weld, by tightening bolts 88 to compress the mechanical seals therein after initial installation. However, if desired, the nozzle 352 may also be fillet welded to the clamp device 86c to provide additional support. FIG. 26 depicts a technique for attaching the sleeve of FIGS. 21-24 to an existing nozzle 352 so that the sleeve may be tightened to the bore while attached to the existing nozzle. In the nozzle -to- sleeve attachment 400, the nozzle 352 is welded to a compression ring 402, and the sleeve 61 is rotatably coupled to the compression ring 402 rather than being welded to the existing nozzle 352. The compression ring 402 is internally threaded, and an externally threaded tubular sleeve 404 through which the existing nozzle 352 passes is threaded to the compression ring 402. The tubular sleeve section 404 has wrenching flats 59 is rotatably coupled to the drive sleeve 61 by a mechanical clamp arrangement 406 comprising circumferential grooves 408, 409 in adjacent ends of the drive sleeve 61 and the tubular sleeve section 404, and a split clamp 412 having annular projections 414, 415. The annular projections 414, 415 of the split clamp 412 are rotatably received and engaged in grooves 408, 409, respectively. A jam nut 418 is threaded on the tubular sleeve section 404 against the compression ring 402. After the drive sleeve 61 has initially been installed and tightened, the tubular sleeve section 404, compression ring 402 and jam nut 418 (loosened) are installed and clamped to the drive sleeve 61 with the clamp arrangement 406, as follows. The compression ring 402 is welded to the existing nozzle with the fillet weld 356. The tubular sleeve section 404 is turned in a counter clockwise direction such that tubular sleeve section 404 rotates out of compression ring 402 and compresses the nozzle, with minimal loading, into the compression ring 402 to load the drive sleeve 61. The jam nut 418 is then tightened. During service, the loading on the drive sleeve 61 may be adjusted without breaking the fillet weld 356 by backing off the jam nut 418 and rotating the tubular sleeve section 404 using the wrenching flats 59. The jam nut 418 is then re-tightened against the compression ring 402. Also, depending upon the torqued relationship of the compression ring 402, drive sleeve 404 and jam nut 418, the nozzle-to-sleeve attachment 400 may provide either tension or no load to the existing nozzle 352, or compression as described above. Where the existing nozzle 352 is either tensioned or not loaded, ring 402 would not be referenced to as a "compression" ring. FIG. 27 depicts a technique for attaching the sleeve of FIGS. 21-24 to an existing nozzle 352 so that the sleeve may be tightened to the bore while attached to the existing nozzle. In the nozzle -to- sleeve attachment 400 (FIG. 26), the nozzle 352 is welded to a compression ring 402, and the sleeve 61 is rotatably coupled to the compression ring 402 which is welded to the existing nozzle 352. In the embodiment of FIG. 27 no weld to the nozzle is employed, and instead the nozzle is frictionally clamped to the sleeve 61 by a split clamp 420. The split clamp 420 includes the annular projection 414 of clamp 412 which is rotatably received and engaged in the groove 408 of the sleeve 61. The other end of the split clamp 420 includes a split ring 424 extending into frictional engagement with the nozzle 352. Tightening the bolt 394 compresses the split ring 424 around the nozzle 352 to frictionally clamp the nozzle. The sleeve 61 may be tightened simply by loosening bolts 394 on the split clamp 420. Once sleeve 61 is tightened, bolts 394 are retorqued. As mentioned above, after implementing a repair according to the embodiments of FIGS. 21-25, a full or partial nozzle replacement may be made later as described herein for the applicable replacement, e.g., FIGS. 17-20. For example, after a repair according to FIGS. 22, 23 or 24 was made, a partial nozzle replacement according to FIGS. 18, 19 or 20, respectively, may be made, and after the repair shown in FIG. 22 was made, the full nozzle replacement shown in FIG. 17 may be made. Other variations will be apparent to those of skill in the art from the disclosure herein. As also mentioned above, the implementations described herein of the nozzle replacements and repairs avoid all or some of the problems discussed above, and the embodiment of FIG. 17 avoids all of the problems discussed above. The partial nozzle replacements, such as the embodiment of FIG. 18, avoid all the concerns except that of continued cracking. Partial nozzle replacements that do not have sealing between the new and existing nozzles, such as the embodiment of FIG. 11, may not over come all the corrosion concerns if they are placed in an environment that is not stagnant. While further analysis, evaluation and information is necessary to determine whether all of the embodiments disclosed herein are acceptable for the life of the facility, they are all believed to be acceptable for at least short term, most likely long term, and in most cases the life of plant given that they are installed in the locations for which they were designed, e.g., those embodiments with a gap between the existing and new partial nozzle are in stagnant environments, and repaired nozzles include fracture mechanic evaluations justifying that the nozzle will not continue to crack beyond the seal. However, the repair embodiments and the partial nozzle replacement embodiments not employing seals between the existing and new partial nozzle (in environments which may not be suitable for long term corrosion concerns) could at least be used for an interim period until such time it was determined necessary or prudent to replace the repaired nozzle or partial nozzle with one of the other full or partial nozzle replacement embodiments described herein. In certain circumstances, such as particular plant modes of operation, one repair or replacement embodiment or technique may be more suitable than another given these conditions. Those experienced in the art are capable of making the determination as to whether or not a particular embodiment or technique used for a particular application can be considered short term, long term, or a life of plant repair or replacement. In the event the repair or replacement is considered as interim and it becomes necessary, those experienced in the art can select the appropriate full or partial nozzle replacement that corresponds to the repair or partial nozzle previously used. As indicated above, rather than replace all or part of an existing nozzle or repair an existing nozzle, the invention can be applied to provide a plug, particularly for a heater sleeve. For example, in the case of a failed heater and heater sleeve, the heater sleeve may be fully or partially removed and replaced as described above but with a plug or capped nozzle rather than a nozzle due to the fact that a new heater may not be available. All of the embodiments depicted in FIGS. 3-26 may be used in large bore piping, as well as pressure vessels in general, and do not require entry into the vessel or pipe or a remote system to install them. In the claims, the term vessel is used in a broad sense and, unless otherwise indicated, is meant to include, but not to be limited to, vessels, piping, etc., of different types which may operate under pressure and which may be used in different nuclear and non-nuclear ASME pressure vessel applications, and the term nozzle is used in a broad sense and is meant to include, but not be limited to, nozzles, sleeves, large bore pipes, pipe portions, etc. Also, where applicable in the claims, "nozzle" encompasses a nozzle assembly, which may include a nozzle having one or more parts, seals, one or more anti-rotation devices, and one or more leak paths. While the invention has been described and illustrated in connection with preferred embodiments, many variations and modifications, as will be evident to those skilled in this art, may be made without departing from the spirit and scope of the invention. For example, it will be apparent that one or more features in one embodiment may be applicable to other embodiments, or applied to a full or partial nozzle replacement, or to a repair. Although many examples are described above, specific reference of the applicability of a feature described in connection with one embodiment is not made in every other applicable embodiment One such example is that the thrust bearing 202 in FIG. 13 may be used in the embodiments depicted in many of the other figures. Other examples will be apparent to those of skill in this art. Therefore, the invention, as set forth in the appended claims, is not to be limited to the precise details of construction set forth above, as such variations and modifications are intended to be included within the spirit and scope of the invention as defined in the appended claims. |
052727327 | claims | 1. A permanent seal for providing an effective water barrier over an annular space between a nuclear reactor vessel flange and a surrounding annular ledge comprising: an annular space-spanning deck structure; a first annular seal member for providing a support of the deck and a seal between said deck structure and the surrounding annular ledge; and a second flexible annular seal member for providing a seal between said deck structure and the vessel flange and wherein a portion of said second seal member is straight walled and another portion of said second seal member is force absorbing. an annular space-spanning deck structure; a first annular seal member for providing a seal between said deck structure and the surrounding annular ledge; and a second flexible annular seal member for providing a seal between said deck structure and the vessel flange and wherein a portion of said second seal member is a straight walled cylinder and another portion of said second seal member is arcuate. 2. The seal according to claim 1 wherein said force absorbing section of said second seal member is arcuately shaped wall section. 3. The seal according to claim 1 wherein said arcuate section of said second seal member is composed of at least one coil shaped section. 4. The seal according to claim 3 wherein said at least one coil shaped section is two coil shaped sections. 5. The seal according to claim 1 wherein said cylindrical section of said second seal member is sealed to said deck structure and said force absorbing portion of said second seal member is sealed to the vessel flange. 6. The seal according to claim 1 wherein said deck structure further comprises hatches which may be opened to provide access to and ventilation of the reactor vessel cavity below the vessel flange and surrounding annular ledge. 7. The seal according to claim 1 wherein the material for the second seal member is metal. 8. The seal according to claim 7 wherein said metal is stainless steal. 9. The seal according to claim 7 wherein the material for said deck structure and said first seal member is metal. 10. The seal according to claim 9 wherein the material for said deck structure, said first seal member and said second seal member is stainless steal. 11. A permanent seal for providing an effective water barrier over an annular space between a nuclear reactor vessel flange and a surrounding annular ledge comprising: 12. The seal according to claim 11 wherein said arcuate section of said second seal member is a coil and ring shaped structure. 13. The seal according to claim 11 wherein said arcuate section of said second seal member is composed of at least one coil shaped section. 14. The seal according to claim 13 wherein said at least one coil shaped section is two coil sections. 15. The seal according to claim 11 wherein said cylindrical section of said second seal member is sealed to said deck structure and said arcuate portion of said second seal member is sealed to the vessel flange. 16. The seal according to claim 11 wherein said deck structure further comprises hatches which may be opened to provide access to and ventilation of the reactor vessel cavity below the vessel flange and surrounding annular ledge. 17. The seal according to claim 11 wherein the material for the second seal member is metal. 18. The seal according to claim 17 wherein said metal is stainless steal. 19. The seal according to claim 17 wherein the material for said deck structure and said first seal member is metal. 20. The seal according to claim 19 wherein the material for said deck structure, said first seal member and said second seal member is stainless steal. |
054611857 | abstract | The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide. |
description | Embodiments of this disclosure relate generally to radiation apparatuses and methods. In particular, various embodiments of methods of designing radiation shields for radiation machines and systems are described. Medical linear accelerators (LINACs) are useful in producing high energy radiation to treat patients with cancer. Depending on the type of cancer, position, size of the tumor and its surrounding critical organs, and the patient size, medical LINACs operating at energies from ˜4 to ˜20 MeV range are used for radiation therapy procedures. To ensure safety however, protective measures must be taken to limit unwanted radiation to patients outside the planned treatment field and to radiotherapists and the general public to an acceptable level. Electronic components sensitive to radiation exposure in the system also need to be protected from excessive radiation to prolong their useful life span. Radiation in directions other than the direction toward the intended places, such as tumor, is undesirable. Hence proper shielding is required. Unwanted radiation is called radiation leakage which in general have three major aspects: (1) Leakage to patients. This results in higher risk of patients getting secondary cancers, hence the lower the better. (2) Leakage to general public. Linac leakage to the operators or others is shielded by the treatment room. Reduced leakage can result in lower cost of the treatment room. (3) Leakage to the linac system itself. There are many PCBs and components located inside the treatment room and many of their performance degrade over time due to radiation damage. Reduced leakage to these parts can reduce the maintenance resource and service cost. Because of above, a well-designed shielded system is very important. Conventionally, machine shielding is developed based on a trial-and-error approach, which requires costly schedules, budgets and resources, and results in heavy and costly shielding parts and assembly. Furthermore, machine leakage performances are unknown until prototypes are constructed and actual measurements are made. In addition, unintended leakage hot spots may exist, incurring more expensive room shielding and short hie span of electronic machine components. This application describes a systematic method to design and optimize the shielding that reduces the cost of the shielding itself and overall cost in above aspects. Embodiments of methods for designing LINAC head shields and LINAC system shields are described. Also described are LINAC head shields and system shields constructed using the design methods. Other embodiments are described further herein. Various embodiments of methods for designing LINAC head shields and system shields are described. It is to be understood that the disclosure is not limited to the particular embodiments described as such may, of course, vary. An aspect described in conjunction with a particular embodiment is not necessarily limited to that embodiment and can be practiced in any other embodiments. For instance, various embodiments are presented using LINACs producing x-ray radiation. It will be appreciated that the disclosed methods can be implemented in other types of radiation machines and systems producing other types of radiation such as gamma rays. All technical and scientific terms used herein have the meaning as commonly understood by one of ordinary skill in the art unless specifically defined otherwise. As used in the description and appended claims, the singular forms of “a,” “an,” and “the” include plural references unless the context clearly dictates otherwise. The term “or” refers to a nonexclusive “or” unless the context clearly dictates otherwise. In the following description, well known components or steps may not be described in detail in order to avoid unnecessarily obscuring the embodiments of the disclosure. As used herein, the phrase “angular distribution of radiation” refers to a distribution of radiation propagating from a source as a function of spatial angle. By way of example, FIG. 1A schematically shows the trajectories of electrons before and after bombarding a target in producing radiation. FIG. 1B schematically shows the trajectories of photons produced. FIG. 1C schematically shows the trajectories of both electrons and photons, combining FIGS. 1A and 1B. As shown in FIGS. 1A-1C, when energetic electrons are stopped in a target, they generate Bremsstrahlung radiation (photons) which propagates largely in a forward direction. There is a considerable angular spread in the resulting photons. Some photons backscatter, i.e., propagate at 180 degrees with respect to the electron beam direction. The distribution of photons generated is a function of spatial angle, and may be effected by the energy and/or energy spectrum of the electrons striking the target, the geometric configuration of the target, the material composition of the target, and the surrounding target holder. As the TVL changes with angle, it is desirable to provide the disclosed method, as will be described in greater detail below, to design linac shield with minimum material and cost to achieve intended design per leakage specification. As used herein, the phrase “photon fluence” refers to the number of photons per unit area. The phrase “energy photon fluence” refers to a summation of photon fluence weighted by the corresponding photon energy. The phase “radiation dose” or “dose” refers to absorbed dose or energy deposited per unit mass. In this disclosure, the phrase “angular distribution of radiation dose” may be used interchangeably with the phrase “dose lobe.” The phrase “angular distribution of energy photon fluence” may be used interchangeably with the phrase “energy fluence lobe.” The phrase “angular distribution of photon fluence” may be used interchangeably with the phrase “photon fluence lobe.” In this disclosure, the phrase “angular distribution of radiation” includes angular distribution of photon fluence, angular distribution of energy photon fluence, and/or angular distribution of radiation dose. In general, less material is needed at large angle due to reduced dose lobe intensity and reduced TVL. As used herein, the phrase “primary radiation” refers to the portion of radiation emitted directly from a target source that did not suffer any interactions or scattering events with any of the radiation collimation components and is confined to the treatment field. The phrase “secondary radiation” refers to radiation scattering off from objects such as the patient, machine parts including components in the collimation system or radiation shield, and walls of the treatment room etc. Secondary radiation is in general undesirable and can be significant to critical organs near the treatment field and/or to electronic components or devices in the radiation system sensitive to radiation. As used herein, the phrase “Tenth-Value-Layer” (“TVL”) refers to the thickness of a material which, when introduced into the path of a given radiation beam, attenuates the radiation intensity to one-tenth of its original value. By way of example, the first TVL of a material attenuates radiation intensity to one-tenth of its original value, the second TVL attenuates it to one-hundredth, and the third TVL attenuates it to one-thousandth, etc. The phrase “Half-Value-Layer” (“HVL”) refers to the thickness of a material which, when introduced into the path of a given radiation beam, attenuates the radiation intensity to one-half of its original value. By way of example, the first HVL of a material reduces radiation intensity to half of its original value, and the second HVL reduces it by a factor of four, etc. The TVL, HVL or the similar, of a material depends on the attenuating property of the material such as the density and atomic number of the material. In general, high density materials have higher radiation stopping power than lower density materials with the same thickness. The TVL, HVL or the similar, of a particular material varies depending on the angular location of the material with respect to the source. In general, the TVL, HVL or the similar, of a particular material placed in a path at 0° with respect to the direction of the electron beam striking the target is greater than the TVL, HVL, or the similar, of the material placed in a path at 180° or opposite to the electron beam direction. Hence less material is required at a larger angle (in relative to the electron beam direction) and an optimized shield design can be based on TVL value at different angles. FIGS. 7 and 9, which will be described in greater detail below, show how the TVL values change with the angle for pure tungsten and lead respectively. Therefore, as used herein, the phrase “angular function of thickness of a material in attenuating radiation to a certain level of its original value” refers to the function or dependence of Tenth-Value-Layer (“TVL”), Half-Value-Layer (“HVL”), or the similar, of a material with respect to the angular location of the material relative to the radiation source. As used herein, the term “angle,” “angular,” or other grammatical equivalents refers to the deflection angle off the z-axis, which points forward along the direction of the electronic beam striking a target in producing radiation. As used herein, the term “shield” refers to any suitable material that attenuates radiation and is configured to reduce or minimize radiation leakage from a source to protect a point of interest (POI) or to ensure that the POI has a leakage lower than the predetermined threshold. For example, the phrase “head shield” refers to a shield around a target source to confine radiation produced by the target in a particular direction and reduce radiation to a predetermined threshold in a particular direction with respect to the beam's propagation direction. The phrase “shield for a point of interest” refers to a shield configured to protect a point of interest, such as a radiation sensitive component or device in a radiation system, by shielding off radiation originated as leakage from a source or patient scatter etc. A shield can be constructed by a single piece or by a combination of two or more pieces. A shield may be configured to perform one or more functions including radiation shielding, beam collimating etc. As used herein, the phrase “leakage from a source” or “source leakage” refers to radiation leaking through a shield of the radiation source. As used herein, the term “patient” refers to human, animal, or any artificial object (e.g. phantom) in the direct path of the radiation beam or undergoing radiation exposure. Linac Head Shield Design Leakage through linac system shielding components can be a significant source of unwanted radiation to the patient, other working personnel or electronic components in the radiation room. Conventionally, linac system shielding is developed based on a trial-and-error approach, which requires costly schedules, budgets and resources, and often times results in heavy and costly shielding parts and assemblies. Further, linac system leakage performance is unknown until a prototype is constructed and actual measurements are made. In a method of constructing a head shield according to one aspect of the disclosure, the angular distribution of radiation propagating from a source and the angular function of thickness of a material in attenuating the radiation to a certain level of its original value are determined. Based on the angular distribution of radiation from the source and the angular function of thickness of the material in attenuating the radiation, the thicknesses of the material at a plurality of angular locations around the source can be calculated in order to achieve a radiation level to or less than a pre-established threshold value. A shield around the source can be then constructed based on the calculated thicknesses of the material for any angle around the radiation source. The angular distribution of radiation intensity propagating from a source and the angular function of thickness of a material in attenuating the radiation to a certain level of its original value can be determined using Monte Carlo methods. Monte Carlo methods are known in the art and thus their detailed description is omitted herein in order to avoid obscuring description of this disclosure. In general, Monte Carlo methods are statistical simulation methods. They are a numerical solution to a problem that models objects interacting with other objects or their environment based upon simple object-object or object-environment relationships. They represent an attempt to model a system through direct simulation of the essential physics interactions of the system in question. Various aspects of Monte Carlo methods are described in A. Bielajew, “Fundamentals of the Monte Carlo Method for Neutral and Charged Particle Transport,” The University of Michigan, Ann Arbor, Mich., (2001) (hereafter the “Bielajew publication”). The disclosure of the Bielajew publication is incorporated herein by reference in its entirety. The use of Monte Carlo simulation will be illustrated below in conjunction with the description of exemplary methods for designing head shields and system shields. Although Monte Carlo can model the intended design directly, a separate simplified computing program was created, using MATLAB and a CAD program (such as SolidWorks). This reduces the total simulation time in design iteration and optimization. The thicknesses of a material needed for attenuating the radiation to or less than a threshold value can be calculated using various algorithms or computer software (hereafter “ShieldTool”). Parameters such as the angular distribution of radiation generated and propagating from a source, the angular function of thickness of a material in attenuating the radiation to a certain level of its original value, or other Monte Carlo simulation data or empirical data can be used as inputs into ShieldTool, which can then calculate the shield thickness needed for attenuating radiation to a specified value for a specified direction. The calculated 3-dimensional (3D) shielding information can be used in constructing a shield at any point around the radiation source. One exemplary ShieldTool may include a CAD program (such as SolidWorks, ProEngineering, CREO or similar) coupled to a MATLAB script and graphical user interface (GUI). SolidWorks and MATLAB are known in the art and thus their detailed description is omitted herein in order to avoid obscuring description of this disclosure. In general, the SolidWorks CAD program or an Add-on of the CAD program may generate raw data based on 3D models. Designated post-processing data analysis, such as MATLAB scripts with GUI, can be used to import the ShieldTool data and compute 3D leakage distributions around the radiation system. The ShieldTool CAD Add-on may generate a line length of the material at each angle around the source for each part of the assembly. The line lengths for all the parts are combined in post processing to calculate how much radiation attenuation will occur before any test point. ShieldTool is generally considered a “ray tracing” program in that it looks at straight lines out from the source and generally does not incorporate secondary sources such as radiation scatter off of materials. Therefore, to incorporate the effect of e.g. patient scatter and its contribution to the dose accumulated at a particular POI, the patient scatter is modeled as a secondary point source. Exemplary embodiments of methods for designing a head shield will now be described with reference to the figures. It should be noted that some figures are not necessarily drawn to scale. The figures are only intended to facilitate the description of specific embodiments, and are not intended as an exhaustive description or as a limitation on the scope of the disclosure. Linac Head Shield Design A Monte Carlo simulation was performed to provide the angular dependence of the dose lobe as a result of the physical interactions of 6 MeV electrons incident to a target button and target holder of a radiation system. The tenth-value-layers (TVLs) were also calculated for tungsten and lead. Additional TVLs for materials other than lead and tungsten can be obtained by scaling with the corresponding density. The obtained angular dependence of the dose lobe and TVLs were used to design a shield around the target. The results showed uniform leakage below a pre-established leakage threshold of 0.01% (100 ppm) for the entire angular spectrum, assuming azimuthal symmetry. Monte Carlo Simulation Methods and Materials Monte Carlo simulations were conducted using Geant4 (version 9.4.p02). The input electrons were based on an independent simulation of the electron propagation through the linear accelerator using Parmela. The electron input phase space contained ˜105 particles. The energy spectrum of the input Parmela electrons is shown in FIG. 2A-2C. It is characterized by a maximum energy Emax of 6.2 MeV. FIG. 3 shows a simplified target configuration 100, comprising a target 102 and a target holder 104. In this exemplary configuration, tungsten target 102 and target holder 104 had azimuthal symmetry. The target 102 may be a transmission target. FIG. 4 shows a particle fluence scorer setup 110 to account for all particles reaching at a 1 m radius spherical detector using Geant4. The center of the scorer was placed on the top surface of the target button 102. Filters that allowed scoring of particles with energy in the interval Ek∈[Ek, Ek+dE] and azimuthal direction in the interval θj∈[θj, θj+dθ] were included. The energy interval [0, Emax] was divided in 100 bins, hence the energy bin size was dE=0.062 MeV. The azimuthal angle increment was set to dθ=0.5 degree. Zero (0) degree points towards the patient (isocenter). Dose Lobe and TVL Calculation In the model described above, the particle fluence was converted to dose-to-water in order to obtain the dose lobe dependence on angle θ. The particle fluence was converted to dose-to-water according to the following equation: D ( cGy ) = 1.6 ⨯ 10 - 8 R 2 Φ E μ en ( E ) ρ , [ 1 ] where E is the particle energy, Φ is the particle fluence, μ en ( E k ) ρis the energy-mass absorption coefficient for water (in cm2/g), R is the distance from the origin to the detector surface (i.e. 1 m) and 1.6×10−8 is a constant to convert MeV/g to cGy. This formula is valid for monoenergetic beam; for non-monoenergetic beams, it becomes Eq. [2]: D ( θ j ) = 1.6 ⨯ 10 - 8 R 2 Σ 0 Emax E k S ( E k , θ j ) μ en ( E k ) ρ , [ 2 ] where S(Ek, θj) is the probability to find a particle with energy Ek∈[Ek, Ek+dE] and angle θj∈[θj, θj+dθ], and the sum is over all the detected particles with energy below Emax. To perform this conversion, the energy-mass absorption coefficient for water (i.e. tissue) was used. To calculate the transmission factor T(θj, x) through a shield of thickness x, Eq. [3] was used: D ( θ j , x ) = 1.6 ⨯ 10 - 8 R 2 Σ 0 Emax E k S ( E k , θ j ) μ en ( E k ) ρ B ( E k , μ x ) e - μ x , [ 3 ] where B(Ek, μx) is the dose buildup factor and e−μx is the linear attenuation factor. The dose buildup factor can be expressed using the Berger equation: B(Ek, μx)=1+C(Ek)μxeD(Ek)μx, where the dose buildup C and D coefficients for tungsten and lead were extracted from prior studies and interpolated for all energy sampling points. Buildup coefficients for various materials are disclosed in D. K. Trubey, A survey of empirical functions used to fit gamma-ray buildup factors, Oak Ridge Nat. Lab. (1966) (hereafter the “Trubey publication”). The disclosure of the Trubey publication is incorporated herein by reference in its entirety. The linear attenuation coefficient pt for tungsten and lead, respectively, was also interpolated to cover the entire set of energy sampling points. In the above equations, μ also depends on the particle energy Ek. The transmission factor T(θj, x) for a shield of thickness x corresponding to the azimuthal angle θj is expressed as: T ( θ j , x ) = D ( θ j , x ) D ( θ j , 0 ) . [ 4 ] Simulation Results FIG. 5 shows a plot of the dose lobe. The dose at each point is normalized to the central axis dose maximum. For comparison purposes, the normalized energy fluence lobe is also shown in FIG. 5. The energy fluence lobe is a summation of particle fluence weighted by the corresponding particle energy, except that the energy-mass absorption coefficient is not factored in. FIG. 5 shows that the radiation intensity is the highest on the beam's z-axis (i.e. at 0 degree). There is less intense radiation propagating in the horizontal plane (at 90 degree) and much less intense radiation is backscattered (at 180 degree). This means that different shielding amounts are needed at different angles for attenuating radiation to a uniform predetermined threshold. FIG. 6 shows the transmission coefficient of tungsten (ρ=18.0 g/cm3) versusshield thickness for a family of scattering angles between 0 and 180 degrees. The boundary curve 200 represents the transmission coefficient of tungsten at 0 degree (i.e. straight forward photons). The boundary curve 300 represents the transmission coefficient of tungsten at 180 degree (i.e. straight backscattered photons). The transmission coefficient of tungsten at an angle between 0 and 180 degrees can be found between the curve 200 and curve 300. FIG. 6 shows that for tungsten the difference of 1st TVL at 0 degree and 180 degree is 0.8 cm (3.2 cm at 0 degree (reference 202) and 2.4 cm at 180 degree (reference 302)). The difference of the 2nd TVL at 0 and 180 degrees is 1.2 cm (6.6 cm at 0 degree (reference 204) and 5.4 cm at 180 degree (reference 304)). The difference of the 3rd TVL at 0 and 180 degrees is 1.4 cm (10.2 cm at 0 degree (reference 206) and 8.8 cm at 180 degree (reference 306)). FIG. 6 further shows that the difference of the 4th TVL of tungsten at 0 and 180 degrees is as much as 1.8 cm (13.8 cm at 0 degree (reference 208) and 12 cm at 180 degree (reference 308)). FIG. 7 shows the angular dependence of individual TVLs (TVL1, TVL2, TVL3 and TVL4) and average TVL for tungsten. FIG. 7 shows that the average TVL decreases from 34.4 mm at 0 degree to 30.2 mm at 180 degree. FIG. 7 also shows that the higher the TVL order, the greater the beam hardening effect, hence the more material is needed to achieve the desired radiation attenuation. Beam hardening effect refers to the mean photon energy increase obtained as the radiation passes through more and more shield due to the stopping of the low energy photons. FIG. 8 shows the transmission coefficient of pure lead (ρ=11.35 g/cm3) versus shield thickness for a family of scattering angles between 0 and 180 degrees. The boundary curve 400 represents the transmission coefficient of lead at 0 degree. The boundary curve 500 represents the transmission coefficient of lead at 180 degree. The transmission coefficient of lead at an angle between 0 and 180 degree can be found between the curve 400 and curve 500. FIG. 8 shows that for lead the difference of 1st TVL at 0 degree and 180 degree is 1.5 cm (5.0 cm at 0 degree (reference 402) and 3.5 cm at 180 degree (reference 502)). The difference of the 2nd TVL at 0 and 180 degrees is 2.0 cm (10.5 cm at 0 degree (reference 404) and 8.5 cm at 180 degree (reference 504)). The difference of the 3rd TVL at 0 and 180 degrees is 2.1 cm (15.6 cm at 0 degree (reference 406) and 13.5 cm at 180 degree (reference 506)). FIG. 8 further shows that the difference of the 4th TVL of lead at 0 and 180 degrees is as much as 2.5 cm (21.0 cm at 0 degree (reference 408) and 18.5 cm at 180 degree (reference 508)). FIG. 9 shows the angle dependence of individual TVLs (TVL1, TVL2, TVL3 and TVL4) and average TVL for lead. FIG. 9 shows that the average TVL decreases from 51.8 mm at 0 degree to 45.4 mm at 180 degree. Similar to FIG. 7, the beam becomes harder as additional TVLs are inserted in the path of the beam. Designing Head Shield Based on the simulated angular dose lobe and angular TVL for a specific shielding material, a shield around the target was constructed in order to lower the head leakage below a certain threshold. If the desired leakage threshold is Ispec (normalized to an open field of 10×10 cm2), then the shield thickness at a given angle (x(θj)) can be calculated according to the following equation:x(θj)=TVL(θj)*[log10 D(θj)−log10(Ispec)]. [5]where TVL(θj) is the average tenth-value-layer shown in FIG. 7 (for W) and FIG. 9 (for Pb). D(θj) is the value of the normalized dose lobe (shown in FIG. 5) corresponding to the angle θj. By applying this procedure, an “avocado-like” shield 600 was obtained, as shown in FIG. 10. For a preliminary shield design exercise, the desired spec or leakage threshold was 100 ppm (0.01%), hence log10(Ispec)=−4. An angle increment of 5° was used for θj. The primary and secondary collimators 602, 604, as well as the proximal and distal MLCs 606, 608, were truncated such that their outer envelope followed a predetermined angle e.g. 15°. For simplicity, the primary and secondary collimators 602, 604 were drawn as cones with circular apertures having an area proximately equal to a field of 30×30 cm2. Similarly, the MLC aperture, when open, had an area equal to a field of 10×10 cm2. The avocado shield 602 has an inner radius of 15 mm and the outer radius is given by Eq. [5]. Avocado-like shape came from the fact that less shielding material is needed in the directions that are opposite to the incoming electron particles. FIG. 11 shows plots of simulated dose lobes for the designed head shield when the MLCs 606, 608 are fully open and fully closed respectively. FIG. 11 provides information on leakage through the target shield versus angle. As described, 0.01% is a desired leakage threshold for the head shield. For angles larger than 20°, outside the in-field treatment area, the dose lobe has a substantially “flat” or uniform aspect, which is a very good output for this preliminary shield design. FIG. 11 further shows that the dose lobe is on average 16% larger than the desired value, with a “hump” at ˜85° where the dose lobe 52% above the desired limit (see inset of FIG. 11). To correct this, a more conservative TVL can be used instead of the average TVL value. As shown in FIG. 7, the second, third and fourth TVLs are all larger than the average TVL. Since the second TVL is on average ˜1.5 mm larger than the average TVL, modifying the shield thickness according to x(θj)=TVL2(θj)*[4+log10 D(θj)] would lower the dose lobe below the 0.01% specification. FIG. 12 shows that modifying the shield thickness using the second TVL can provide a uniform dose lobe with an average leakage of 94 ppm for angles between 20° and 180°, with the maximum value of 133 ppm at 85°. However, FIG. 12 also shows that building the head shield based on the 2nd TVL angular dependence did not eliminate the “hump” and the resulting shield was too conservative beyond ˜110° due to leakage being lower than the desired threshold of 100 ppm. Optimizing Head Shield Design In an alternative embodiment, the head shield was modified or adjusted according to the following equation:{tilde over (x)}(θj)=L(θj)*[log10 D(θj)−log10(Ispec)], [6]where L(θj) represents a modified or corrected angular TVL curve: ( θ j ) = x ( θ j ) log 10 D ( θ j ) - log 10 D coll ( θ j ) . [ 7 ] In Eq. (7), x(θj) is given by Eq. [5] and Dcoll(θj) is the dose lobe shown in FIG. 12. The modified or corrected angular TVL is shown in FIG. 13 (blue squares), along with the average TVL and 2nd TVL for lead. Note that the modified TVL shows a “hump” at 85° which was expected to eliminate the leakage hump previously seen in FIGS. 11 and 12 where the TVL values used were not large enough to reduce the leakage below the desired specification of 100 ppm. At angles smaller than 20° the average TVL values were used since, as shown in FIG. 11, they ensured a leakage well below 100 ppm. The modified TVL curve for all angles is shown by the black dot curve. This modification recipe can be applied iteratively to improve the TVL curve. For example, in a case where the leakage specification changed from 100 ppm to 400 ppm, a new head shield was constructed and the modification recipe was applied one more time according to the following equation: TVL ″ ( θ j ) = ( θ j ) * [ log 10 D ( θ j ) - log 10 I spec ″ ) ] log 10 D ( θ j ) - log 10 D coll ( θ j ) [ 8 ] The second modification led to the results shown in FIG. 14 by the curve of 2nd TVL correction. The TVL curve is less conservative than the previous dependence since the leakage requirement was relaxed from 100 to 400 ppm. This TVL correction recipe can be applied to any radiation source shield, leading to a cost-efficient, optimal solution while ensuring that the shield weight is only driven by the desired predetermined leakage threshold. Linac System Shield Design In a radiation system, there are electronic components or devices that are vulnerable to radiation exposure. Components or devices that are radiation sensitive include power supplies, various device controllers, computers, display panels, cameras, sensors, and various electronic components on printed circuit boards. The radiation sensitive components or devices may be located in a gantry or rotatable with the radiation source. The radiation sensitive components or devices may also be located in a stand, patient support (e.g. couch or other lifting platforms), or other structures which do not rotate with the radiation source when in operation. Local shielding for the radiation sensitive components or devices may be needed to ensure their normal functions or prolong their useful life span. In this disclosure, the phrase “point of interest” or “POI” may be used to refer to a component or device in a radiation system that is sensitive to radiation exposure and requires a shield to block or minimize radiation deposited on it. Radiation deposited on a POI may come from radiation leakage of the linac system and the patient scatter as a result of the primary beam hitting the patients. Patient scatter may be significant to some POIs and less to the others depending on if the POI rotates with the source, the distance and the angel to the source. The shielding design according to this disclosure considers patient scatter in addition to the leakage from the primary source. The distribution of radiation propagating from a primary radiation source or secondary source placed at the machine isocenter (also coincidental with the patient's tumor) is a function of the corresponding spatial angles. More radiation propagates in a forward direction and less radiation propagates in a horizontal plane relative to the source and much less radiation backscatters. As such, a radiation sensitive component or device may receive same or different amounts of radiation, depending on the location of the component or device in the radiation system. By way of example, a power supply fixedly located in a radiation system may receive different amounts of radiation from source leakage and patient scatter during operation of the system since both the distance and the angle between the power supply and the radiation source may change as the radiation source rotates around the patient. On the other hand, an electronic controller located in a gantry or rotatable with a radiation source receives the same amount of radiation during operation of the system because both the distance and the angle between the electronic controller and the radiation source and/or isocenter remain unchanged as the radiation source rotates assuming the patient is somewhat cylindrical-like, centered at iso-center of the system. Therefore, the “rotational factor,” whether or not a radiation sensitive component or device rotates with the primary source during operation, needs be considered in designing a local shield for the component or device. In modern radiation therapy, treatments are often delivered using intensity modulated radiotherapy (IMRT), which involves shaping treatment fields with multileaf collimators (MLCs). In IMRT, both the shape of the treatment field and radiation fluence may be changed or “modulated” such that the radiation fluence to healthy organs or tissue is as low as possible to avoid risks such as secondary malignancies. This means that the “modulation factor,” i.e., radiation fluence blocked by MLCs during IMRT, need be accounted for in the calculation of patient scatter contribution to the net accumulated radiation deposited for the component to be shielded. In general higher modulation factor indicates a higher radiation generated from the linac system, hence higher radiation leakage to POIs from the primary source. Therefore, modulation factor needs to be considered in shielding design. Patient scatter contribution to radiation deposited on a component is typically obtained by directing an open or unmodulated treatment field of a specific size to a given phantom. Therefore, in a method of designing a radiation shield for a point of interest in a radiation system, both the target source and scatter source are included in the ShieldTool framework to improve the overall shielding calculation. With the target lobe centered at the target and the patient scatter lobe centered at isocenter, radiation level at all points of interest can be calculated from the two sources respectively and the total radiation at certain point can be obtained by summing up the results. In addition, both the rotational factor and modulation factor (MF) may be accounted for in the shielding design. In one embodiment, a method of constructing a radiation shield for a point of interest (POI) at a location in a radiation system is provided. The radiation system comprises a movable source configured to generate radiation to be delivered to an isocenter in a patient. The POI is movable with the source. According to this embodiment, the radiation reaching at the location of the POI through source leakage and the radiation reaching at the location of the POI from patient scatter are determined respectively. The accumulated radiation at the location of the POI is calculated by summing the contributions. A shield for the POI is constructed based on the accumulated radiation at the location of the POI. In a further embodiment, a method of constructing a radiation shield for a point of interest (POI) at a location in a radiation system is provided. The radiation system comprises a movable source configured to generate radiation to be delivered to an isocenter in a patient. The POI is non-movable with the source. According to this embodiment, a plurality of locations of the source with respect to the location of the POI are selected to account for the 360 degree rotation of the radiation source around the isocenter. At each of the plurality of locations of the source, a first portion of radiation reaching at the location of the POI through source leakage, and a second portion of radiation reaching at the location of the POI through patient scatter are determined respectively. An average first portion of radiation and an average second portion of radiation are calculated. The accumulated radiation reaching at the location of the POI is calculated by summing the average first portion and the average second portion of radiation. A shield for the POI is constructed based on the accumulated radiation at the location of the POI. More locations (angels) from the fixed POIs to the rotating source can be added and averaged throughout 360 degree rotation, based on dynamic treatments which utilize many different gantry angles in relative to tumors. Exemplary methods of designing local shields for radiation sensitive components will now be described with reference to the figures. It should be noted that the specific examples and figures are only intended to facilitate the description of embodiments, and are not intended as an exhaustive description or as a limitation on the scope of the disclosure. Examples—Linac System Shield Design Monte Carlo Simulation Methods and Materials Monte Carlo simulations with an estimated treatment field size were performed in order to create a model for patient scatter. FIG. 15 shows an exemplary model layout for patient scatter simulations. A scoring plane 702 of 2×2 m2 at 500 mm from the isocenter 704 was rotated around the y-axis along the length of the patient 706. The patient 706 was modeled as a water cylinder, with a height of 165 cm and a diameter of 22.6 cm. These dimensions are representative for small patients since the scatter is more reduced for larger patients. The center of the cylinder was placed in the radiation machine isocenter 704 at 1000 mm from the target 708. The cylindrical symmetry of the water phantom 706 allowed a rotation of the planar scorer 702 around the phantom center 704, equivalent to rotating the source 708 around the patient 706 and keeping the scorer 702 fixed. The angle θ measured with respect to the beam's axis 710 is shown in FIG. 15. Although the patient scatter contribution depends on the size of the patient 706, previous simulations revealed that leakage in large patients is smaller than that in small patients (i.e. radiation attenuation is the dominant effect compared to patient scatter). Hence, a small patient was chosen in this example in order to avoid an over-optimistic model. Phase space files were recorded at 500 mm from the isocenter 704 and the angle θ was incremented from 15, 30, 45, 60, 90, 120 and 150 degrees. Angles below 15 degrees were not considered because they project inside the primary radiation beam. The data files were binned using a 10×10 mm2 bin size to calculate photon energy fluence. The energy fluence profiles were normalized to the on-axis output at 1000 mm from source for an open field of 10×10 cm2 with no patient in the beam line. The result of this calculation is also referred to as relative leakage. The input electrons were based on the particle-in-cell codes phase space with ˜105 particles, with an energy spectrum shown in FIGS. 2A-2C. Monte Carlo simulations with Geant4 (version 9.4.p02) were launched on Amazon EC2 computing cloud using the c1.x large cluster. For the MLCs, template was used to simulate an open 18 cm square field. FIG. 16 is a plot of a patient scatter lobe, showing the relative intensity (vertical axis) versus scattering angles (horizontal axis). The dose at each point is normalized to the central axis dose output. FIG. 16 shows that more patient scatter occurs in the path of primary beam (at 0 degree). There is less patient scattering in the horizontal plane (at 90 degree) and even less patient scatter contributions at angles larger than 90 degrees. FIG. 17 is a plot showing the angular dependence of the patient scatter TVL (vertical axis) versus scattering angles (horizontal axis) for lead. FIG. 17 shows that the TVL of lead decreases from 52 mm at 0 degree of scattering angle to approximately 10 mm at 90 degree of scattering angle and approximately 1 mm at 180 degree of scattering angle. Shielding Design for Rotational Components FIG. 18 schematically shows a radiation system 800 illustrating an exemplary method for designing a radiation shield for a rotational component according to some embodiments of the disclosure. As shown, the radiation system 800 includes a movable source 802 configured to generate radiation to be delivered to an isocenter 804 in a patient 806. By way of example, the radiation source 802 may be supported by a gantry 808 which may rotate around the patient 806 in a circular orbit. The head shield 810 around the source 802 is configured to collimate the radiation beam to the desired treatment field at isocenter and minimize radiation leakage in the other directions. A point of interest 812 e.g. an MLC controller or other radiation sensitive component may be rotatable with the source 802. To facilitate description, the location of the POI is referenced at (x, y, z) in a coordinate system with the target source 802 as the reference origin (0, 0, 0). In an alternative coordinate system with the isocenter 804 in the patient 806 as the reference origin (0, 0, 0), the location of the POI 812 can be referenced at (x′, y′, z′). There is a 1 m shift between the target source reference origin and the isocenter reference origin. The spatial angle (θ) of the POI 812 with respect to the target source 802 is measured from the beam direction 814 to a direction 816 from the target 802 to the POI 812. The patient scatter angle (α) is measured from the beam direction 814 to a direction 818 from the isocenter 804 to the POI 812. Still referring to FIG. 18, as the source 802 rotates around the patient 806 from location (1) to location (2), the spatial angle (θ) of the POI 812 with respect to the target source 802 remains unchanged. The patient scatter angle (α) also remains unchanged. Furthermore, the distance from the target source 802 to the POI 812 and the distance from the isocenter 804 to the POI 812 maintain constant. This configuration can be referred to as a “static mode.” The accumulated leakage L (x, y, z) received by the POI 812 can be determined by summing the leakage from the target source (L_TAR) and patient scatter (L_PS) contributions.L(x,y,z)=L_TAR(d,θ)+L_PS(d′,α)/MF [9]where (x, y, z) are the coordinates of the POI 812 with respect to the target reference system [d=(x2+y2+z2)0.5], (x′,y′,z′) are the coordinates of the POI 812 with respect to the patient scatter reference system [d′=(x′2+y′2=z′2)0.5], angle (θ) is the spatial angle of the POI 812 with respect to the target source 802, angle (α) is the patient scatter angle, and MF is the modulation factor, a correction factor applied to the patient scatter contribution (L_PS) to account for the average MU “wasted” per cGy delivered to the tumor. The modulation factor can be empirically determined. Modulation factor can be estimated from the total dose generated by the linac system divided by the delivered dose to the iso-center. Alternative method is to sum up patient scatter and primary source leakage (Eq. [9]) using an averaged modulated field size in the calculations (as well as the measurements). Here, MF=1, when an averaged modulated field size is used. The averaged modulated field size can be calculated by field size area divided by the modulation factor. For example, an averaged field size of 16 cm×16 cm with modulation factor of 4 from the dynamic treatment plan can be translated to averaged modulated field size of 8 cm×8 cm, in which modulation factor 1 is used in Eq. [9]. The target contribution and the patient scatter contribution of Eq. [9] are resulted from the field size of 8 cm×8 cm in this case. The target source leakage contribution to the accumulated leakage at the POI 812 can be determined according to the following equation:L_TAR(d,θ)=LTAR(θ)/d2, where d=(x2+y2+z2)0.5 [10] Eq. [10] shows that the target source leakage contribution depends on the spatial angle (θ) of the POI 812 with respect to the target 802 and the distance from the POI 812 to the target 802. Target leakage contribution decreases as the distance to the POI squared. The spatial angle (θ) appears in the multiplication factor LTAR(θ) which can be determined by Monte Carlo simulation and is given by the target dose lobe, as illustrated in FIGS. 5-9. The patient scatter contribution to the accumulated dose at the POI 812 can be determined according to the following equation:L_PS(d′,α)=LPS(α)/d′2, where d′=(x′2+y′2+z2)0.5 [11] Eq. [11] shows that the patient scatter contribution depends on the patient scatter angle (α) and the distance of the POI 812 to the isocenter 804. The patient scatter contribution decreases as the distance squared. The patient scatter angle (α) appears in the multiplication factor LPS(α) which can be determined by Monte Carlo simulation and is given by the patient scatter lobe, as illustrated in FIG. 16. A local shield around the POI 812 can be constructed according to the accumulated dose deposited at the POI 812 and the angular dependence of the TVLs of the material constructing the shield. Shielding Design for Non-Rotational Components FIG. 19 schematically shows a radiation system 900 illustrating an exemplary method of designing a shield for a non-rotational component according to some embodiments of the disclosure. As shown, the radiation system 900 includes a movable source 902 configured to generate radiation to be delivered to an isocenter 904 in a patient 906. By way of example, the radiation source 902 may be supported by a gantry 908 which may rotate around the patient 906 in a circular orbit. The head shield 910 around the source 902 is configured to collimate the radiation beam to the desired treatment field at isocenter and minimize radiation leakage in the other directions. A point of interest 912 e.g. a power supply or other radiation sensitive component may be fixedly located in the system 900 or do not rotate with the source 902. The location of the POI 912 can be referenced at (x, y, z) in a coordinate system with the target source as the reference origin (0, 0, 0). In an alternative coordinate system with the isocenter 904 in the patient 906 as the reference origin (0, 0, 0), the location of the POI 912 can be referenced at (x′, y′, z′). The spatial angle (θ) of the POI 912 with respect to the target source 902 is measured from the beam direction 914 to a direction from the target 902 to the POI 912. The patient scatter angle (α) is measured from the beam direction 914 to a direction from the isocenter 904 to the POI 912. Still referring to FIGS. 19A-19B, as the source 902 rotates during operation e.g. from location (1) to location (2), (3), or (4), both the spatial angle (θ) of the POI 912 with respect to the target source 902 and the distances from the target source 902 to the POI 912 change. Further, as the source 902 rotates from location (1) to location (2), (3), or (4), the patient scatter angles (αi), as measured from the beam direction 914 to a direction 918 from the isocenter 904 to the POI 912, also changes from α1 to α2, α3, or α4. This configuration can be referred to as a “360 degree mode” since the gantry can fully rotate around the isocenter. This also implies that a sampling procedure should be implemented to account for the spatial variations at POI as the radiation source orbits around the isocenter. To account for these variations, the target source leakage (L_TARi) and patient scatter (L_PSi) contributions can be averaged over incremental angles αi:L(x,y,z)=Σαi=0360°LTARi+LPSi [12]where (x, y, z) are the Cartesian coordinates of the POI 912 in the target coordinate reference system. The target source leakage contribution at a particular spatial angle (L_TARi) and patient scatter contributions at a particular scatter angle (L_PSi) can be determined using the equations described above in conjunction with the calculation for rotational components. Those skilled in the art will appreciate that various other modifications may be made within the spirit and scope of the invention. All these or other variations and modifications are contemplated by the inventors and within the scope of the invention. |
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039403149 | claims | 1. A nuclear reactor fuel element comprising fuel rods having outsides and spacers contacting said outsides and maintaining said rods in spatial relationship; wherein the improvement comprises said rods each comprising a casing having one continuous hollow tubular wall having a length and having at least one tubular section along said length where said wall is of increased thickness, said sections forming said outsides contacted by said spacers and said casings containing nuclear fuel. 2. A nuclear reactor fuel element according to claim 1, wherein said spacers are arranged in a circular mesh and are connected together by webs. 3. A nuclear reactor fuel element according to claim 1, wherein said spacers are arranged in a circular mesh and are connected together by concentric ring straps. 4. A nuclear reactor fuel element according to claim 1, wherein said spacers each have a plurality of projections therein for making contact with the reinforced wall thickness of the tubular casing of the fuel rod. 5. A nuclear reactor fuel element according to claim 1, wherein said spacers each have an insertable ring connected to the spacer for making contact with the reinforced wall thickness of the tubular casing of the fuel rod. 6. A nuclear reactor fuel element according to claim 2, wherein said spacer mesh is provided on its outer circumference with outwardly pointing pads. 7. A nuclear reactor fuel element according to claim 2, wherein the spacer mesh is surrounded by a strap and the strap is provided on its outer circumference with a plurality of pads. 8. A nuclear reactor fuel element according to claim 2, wherein the spacers are mounted on a support rod which is arranged in the center of the mesh. 9. A nuclear fuel element according to claim 8, wherein the spacers are mounted on support rods by means of welds. 10. A nuclear fuel element according to claim 8, wherein the spacers are mounted on a support rod in the center of the mesh by use of additional sheet metal straps above and below the spacers at the support rod. |
058964363 | claims | 1. A fuel assembly in or for a boiling water reactor, said fuel assembly comprising a plurality of vertical fuel rods extending between a top and a bottom tie plate and surrounded by a sleeve-formed casing wherein the fuel rods are positioned with the aid of a plurality of axially separated spacers, and wherein only one fuel rod is detachably arranged in the top tie plate and in the bottom tie plate and at least one of the other fuel rods is fixed to the bottom tie plate and adapted to extend above the top tie plate to obtain a tensile-force-transmitting connection with the bottom tie plate. 2. A fuel assembly in or for a boiling water reactor, said fuel assembly comprising a plurality of vertical fuel rods distributed among four sub-bundles and surrounded by a sleeve-formed casing, each sub-bundle being positioned with the aid of a plurality of axially separated spacers, and each sub-bundle extending between a top and a bottom tie plate, only one fuel rod (3) in each sub-bundle being detachably arranged in the top tie plate and in the bottom tie plate and at least one of the other fuel rods in the sub-bundle being fixed to the bottom tie plate and adapted to extend above the top tie plate to make a tensile-force-transmitting connection with the bottom tie plate. 3. A fuel assembly according to claim 1, wherein the fuel rod which is detachably fixed to the top tie plate fixes the axial position of the top tie plate. 4. A fuel assembly according to claim 1, wherein said other fuel rods rest on the bottom tie plate and freely run through the top tie plate. 5. A fuel assembly according to claim 2 comprising one top tie plate for each sub-bundle. 6. A fuel assembly according to claim 2 comprising a top tie plate which is common to four sub-bundles. 7. A fuel assembly according to claim 1, wherein the top tie plate is made of stainless steel. 8. A fuel assembly according to claim 1, wherein the top tie plate is made of inconel. 9. A fuel assembly according to claim 1 wherein the top tie plate is designed as a spacer. 10. A fuel assembly according to claim 1, wherein the top tie plate comprises a plurality of tubular cells joined together in an orthogonal pattern and the spaces formed between the cells support the upper ends of the fuel rods. 11. A fuel assembly according to claim 1, comprising two supporting extra long fuel rods. |
045487820 | claims | 1. Apparatus for heating a plasma, the plasma being confined in an apparatus of the type wherein the plasma is confined in a vacuum chamber by a magnetic field, comprising: means for producing a space-charge-neutralized, pulsed, ion beam; means for directing the ion beam into the magnetic field before the plasma is fully formed; and means for forming the remainder of the plasma around the beam, the beam transferring its energy to the plasma by classical collisions with the electrons and ions of the plasma. fast pulsed valve means for injecting a puff of gas into the vacuum chamber. (a) producing a space-charge-neutralized pulsed ion beam; (b) injecting the beam into the magnetic field before the plasma is fully formed; and (c) forming the remainder of the plasma around the beam, the beam transferring its energy to the plasma by classical collisions with the electrons and ions of the plasma. directing the beam into the magnetic field along a trajectory generally parallel to the lines of force of the magnetic field. injecting a puff of gas into the vacuum chamber. (d) inactivating external means for driving a current in the plasma subsequent to step (c) so that the entire plasma current is carried by the ion beam. 2. The apparatus recited in claim 1, wherein the ion beam is directed into the magnetic field along a trajectory generally tangential to the lines of force of the magnetic field. 3. The apparatus recited in claim 1, wherein a vacuum region is formed surrounding the plasma; and the ion frequency of the beam is very much larger than the square of the gyrofrequency of the beam in the magnetic field, thereby producing a polarization electric field in the beam when it enters the magnetic field, the polarization field enabling the beam to propagate across the vacuum region. 4. The apparatus recited in claim 1 wherein the plasma-forming means includes: 5. A method for heating a plasma confined in a magnetic field in a vacuum chamber comprising the steps of: 6. The method recited in claim 5 wherein step (b) includes: 7. The method recited in claim 5 wherein step (c) includes: 8. The method recited in claim 5 including the step of: |
description | This application is a continuation of U.S. application Ser. No. 13/258,349, filed Sep. 21, 2011, which is the U.S. national phase of International Application No. PCT/SE2010/050349, filed Mar. 30, 2010, which claims priority to Swedish Application No. 0950199-0, filed Mar. 30, 2009, the disclosures of which are incorporated in their entireties by reference herein. The invention relates to a radiation tolerant camera, comprising a camera module having an electronic image sensor and a radiation shielding enclosure, said enclosure having an opening for allowing passage of light into the image sensor. The camera is formed to be used mainly for monitoring purposes in environments with strong ionizing radiation, mainly neutron and gamma radiation. In the nuclear energy industry, it can be used in a reactor and containment surveillance system, fuel pool inspection, and inspection “missions” for decommissioning. It can also be used in the radiotherapy industry, for instance, for patient monitoring during radiotherapy. The invention is specifically directed to be operated in a neutron radiation environment. In many applications today, tube cameras are used in the environments mentioned above because they are more radiation tolerant compared to cameras that are provided with CCD or CMOS image sensors. It is normally possible to separate any required electronic control units from the radioactive environment and thus avoid or limit some severe effects of the radiation. The conditions of using other types of cameras, and specifically digital cameras, however, are different. Ionizing radiation affects and finally destroys electronic equipment, specifically low voltage and more compact circuits and circuits with high spatial resolution. The ionizing radiation mainly causes temporary damage, so-called soft errors or single-event damage, and permanent damage, so-called atomic displacement. Commercially available devices of today suffer from these effects and produce images of continuously deteriorating quality. The cameras and associated control logic will be broken or have a decreased performance level only after a short period of use in the above-described harsh environment. There is still a need for better image quality that can be achieved with digital image sensors and also a need for cameras that will last longer in such environments. In accordance with the invention, a digital camera module having an electronic image sensor is enclosed by a radiation shielding enclosure. An opening in the enclosure will allow passage of light into the image sensor. The enclosure is made from a material that has low-mass nuclei. In such a material, neutrons can transfer large amounts of their energy to the light nuclei through collisions. In numerous embodiments, boron is added to the enclosure material so as to capture thermal neutrons resulting from the collisions. The complete enclosure in one embodiment can be pivoted or tilted between various operating positions in which the opening is uncovered and directed towards an observed object and a resting position in which the opening is directed towards a shield of radiation shielding material. A backside of the enclosure will be efficiently protected by the shield in the operating position. The opening of the enclosure preferably is covered with a transparent front cover allowing transmission of light and allowing an image to be picked up by the image sensor. The size of the transparent front cover is sufficient for providing a desired viewing angle. Preferably, the front cover also is made from a material that has low-mass nuclei. To improve further the shielding against effects of the radiation, the camera module is thermally connected to a heat absorbing cooling element that will facilitate and improve dissipation of heat from the camera module. The cooling element can include a thermoelectric cooling module, such as a module using the Peltier effect. The cooling capacity of the cooling module can be further improved by heat dissipating means extending exterior of the enclosure from the cooling element. In one embodiment, the heat dissipating means comprises heat pipes. By cooling the camera module to lower temperatures, such as a few degrees above zero, or about 2° C. to 5° C., the image quality from the camera module will be substantially improved. The enclosure can have an average thickness of a few centimeters, such as about five centimeters. At such a thickness, the material will provide sufficient neutron radiation attenuation. In various embodiments, the camera module comprises a standard camera, including a sensor and associated electronics, that is mounted in an insulated and sealed housing. In various embodiments, the housing comprises a moisture-proof layer, so as to ensure that the moisture content within the housing is maintained at a low level. A further radiation shielding layer made of lead, tungsten, or another material having similar shielding and constructional properties can be arranged to shield the camera module from other radiation, such as gamma radiation. The embodiment shown in FIG. 1 comprises a camera module 10 including a conventional digital image sensor 11 and a lens package 13. The camera module 10 is embedded in an insulating body 12 made of a material having a low thermal conductivity. The insulating body 12 is enclosed by a housing 14. Preferably, the camera module fits close to the insulating body and the insulating body to the housing 14 so as to reduce the amount of air within the housing and to avoid condensation at any optical component. In some embodiments, any remaining air is replaced by another suitable gas, such as CO2 or N. In the embodiment shown in FIG. 1, a further radiation shielding layer 15 is provided between the camera module 10 and the insulating body 12. The housing 14 is made of an airtight and neutron radiation shielding material, such as plastic or similar material, and is completely sealed. The radiation shielding layer 15 is provided mainly as a gamma or X-ray shielding means. A front side of the housing 14, as well as of the radiation shielding layer 15, forms an opening which is closed by a transparent front panel 16. The insulating body 12 is formed with an orifice 17 facing the transparent front panel 16. The orifice 17 of the insulating body 12 is beveled from the position of the lens package 13 to a wider open space adjacent to the front panel 16. The housing 14 is enclosed in an enclosure 18 formed by a first box-like part and a second backside part, c.f. also FIG. 2. The outer dimension of the housing 14 corresponds very well with inner dimensions of the enclosure 18 so as to minimize free space there between. The enclosure provides an efficient neutron radiation shield. The image sensor 11 and also the camera module 10 as a whole are thermally connected to a heat absorbing cooling element 20. In the embodiment shown in FIG. 1, the cooling element 20 at one end extends from a backside of the camera module opposite the lens package. An opposite end of the cooling element engages through an electrically insulating pod (not shown) the image sensor or a circuit board supporting the image sensor. The heat absorbing cooling element 20 extends partly out of the housing 14 where it is thermally connected to a cooling device, c.f. FIG. 3, and a plurality of heat pipes 22. The heat pipes extend out of the enclosure and transfer heat very efficiently to a heat sink mounted to the outside of the enclosure, c.f. FIG. 3. The cooling element 20 is designed to maintain the temperature of the camera module and any associated electronics at a temperature below 5° C., preferably at temperatures between 2° C. and 5° C. As shown in FIG. 2, the enclosure 18 is divided into two separate parts. A first box-like part 24 encloses basically the complete housing 14 with the camera module 10. An opening 26 in a front portion with cut-in edge surfaces is dimensioned to receive the transparent front panel 16. The front portion is arched so as to be rotated into protection in a shield, c.f. FIG. 4. A second part of the enclosure forms a backside 28 engaging in a tight manner the box-like part 24. To further seal the connection between the box-like part 24 and the backside 28, both parts can be formed with ribs 30 and corresponding recesses (not shown). The backside further comprises a support block 32 arranged for supporting the heat pipes. One side section of the box-like part 24 is formed with indentations 34 receiving the heat pipes. FIG. 3 shows a backside section of the housing 14 and a cooling device 36, which is thermally connected to the cooling element 20. The cooling device 36 in one embodiment comprises a Thermoelectric cooling module (TEC) 38. It is possible to provide a plurality of interconnected cooling modules, should there be need for it for obtaining a required temperature difference. In alternative embodiments, the camera module is cooled by a conventional air and/or fluid cooling system. The cooling device 36 further supports the heat pipes 22, extending through the enclosure and into a heat sink 40 provided on the outside of the enclosure. The heat sink 40 can comprise a finned element and, if required, a fan. The heat pipes 22 also are thermally connected to the cooling device 36. The heat pipes 22 extend during operating and resting conditions in a horizontal direction. The horizontal orientation of the heat pipes results in a constant heat transmission capacity during operation and in different tilting positions. The enclosure 18 can be rotated around an axis 42 extending in a horizontal direction. In FIG. 4, the camera is shown in a resting and shielded position where the arched section of the front portion of the enclosure engages a shield 44 made of the same material or a similar radiation shielding material as the enclosure. As shown in FIG. 4, the shield 44 is formed with one concave arched side corresponding to the arched front portion of the enclosure 18. As a result, the complete camera can be rotated between adjustable active positions and a resting position in which the opening of the enclosure 18 is very well shielded by the shield 44 against radiation. The rotation between the active position and the resting position can be manually controlled by operating personnel or automatically controlled by a control system, for instance, based on a time schedule. A fully equipped camera is shown in FIG. 5. In this embodiment, the camera includes also lamps 45. The lamps preferably are mounted at the same side of the enclosure 18 as the heat pipes and heat sink so as not to restrain tilting movements of the camera. A pyrometer (not shown) can be mounted on a side of the enclosure or preferably inside the enclosure. The pyrometer can be provided with one or several laser indicators to indicate the direction and area of measurement to an operator of the camera. The complete camera can be mounted at a fixed position on a wall bracket or a commercially available pan and tilt unit. In the embodiment shown in FIG. 5, the camera is mounted on a motorized support 46 which allows independent tilting and panning movements. A base unit 47 comprises a transformer, required electronic means, and connecting means for connecting the camera and motorized support to a remotely arranged control position. Sensitive electronic devices used for controlling and steering the camera can be arranged in a remote position or within an extension 49 of the shield 44. In this embodiment, the radiation sensitive electronic devices, together with power regulating devices, are thermally connected to a cooling system formed by the heat pipes and the heat absorbing means. A microphone 48 is provided to obtain information about sound and noise appearing in the surrounding area. Preferably, the microphone is specifically designed and radiation hardened. The embodiment of the heat absorbing cooling element 20 shown in FIG. 6 comprises a body 21 made of a material having high thermal conductivity, such as a metal. The body 21 has rectangular side edges and a bottom side and a top side. The rectangular side edges have a height and width corresponding to the dimensions of the camera module 10. A base plate 23 extends from the bottom side of the body 21 and engages at least a substantial part of a bottom side of the camera module 10. The base plate will contribute to the heat transfer from the camera module. The body 21 has protrusions on two opposite sides. A first protrusion 25 abuts the camera module 10, and more specifically, a section of the camera module where a circuit board 27 supporting the digital image sensor 11 is situated. A second protrusion 29 is dimensioned to protrude through an opening of the housing 14 so as to ensure an efficient heat transfer out of the housing 14. The second protrusion 29 also will engage the cooling device 36 outside the housing 14 and inside the enclosure 18. The material used for the radiation shielding enclosure and shield may include or be based on hydrocarbon plastics (such as polyethylene, polypropylene, and polystyrene); natural and synthetic rubber (such as silicone rubber); and other plastics or resins containing atoms in addition to hydrogen and carbon (such as acrylic, polyester, polyurethanes, and vinyl resins). These organic polymers show a high effectiveness of shielding against neutrons due to a large concentration of hydrogen atoms in these materials. The combination of a radiation shielding enclosure comprising hydrocarbon plastics and the efficient cooling of the camera module results in a higher image quality, in a short term perspective as well as in a long term perspective. Fast neutrons that are slowed down by repeated collisions with light nuclei form thermal neutrons that can be absorbed by nuclear reactions. The total neutron shielding ability of polyethylene can be improved if a good thermal neutron absorbing material is added to it. An appropriate thermal neutron absorbing material is boron. The material used for the transparent front panel 16 preferably is fully transparent so as to produce a true image of any object in front of the camera module. In a preferred embodiment, materials having a high hydrogen content are used, for instance, Polymethylmethacrylate (PMMA), known as PLEXIGLAS. While certain illustrative embodiments of the invention have been described in particularity, it will be understood that various other modifications will be readily apparent to those skilled in the art without departing from the scope and spirit of the invention. Accordingly, it is not intended that the scope of the claims appended hereto be limited to the description set forth herein but rather that the claims be construed as encompassing all equivalents of the present invention which are apparent to those skilled in the art to which the invention pertains. |
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abstract | The invention comprises a charged particle cancer therapy system used to seal a periphery of a tumor, such as through use of a proton or carbon ion beam searing the outer edges of the tumor, which prevents/hinders nutrient delivery to the tumor resultant in stunted growth of the tumor, halted growth of the tumor, and/or starvation/necrosis of the tumor. Optionally, a tumor sealing layer is formed using multiple passes, of a treatment beam of the charged particle cancer therapy system, across a tumor/healthy tissue boundary layer or voxel. |
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claims | 1. A nuclear power generation system comprising:a supercritical water-cooled reactor (SCWR) containing a cooling water in supercritical state; andan emergency core cooling system of a supercritical water-cooled reactor (SCWR),wherein the emergency core cooling system of a supercritical water-cooled reactor (SCWR) comprising:an accumulation tank;a coolant containing neutron poison micro-particles uniformly dispersed in water which is automatically injected by the accumulation tank and wherein the water is in supercritical state while the emergency core cooling system is operated;a refueling water storage tank in which neutron poison micro-particles are contained; anda high-pressure injection pump for re-injecting refueling water and the water gathered in a drainage tank at the bottom of a nuclear reactor building,wherein the neutron poison micro-particles have an average diameter of 7 μm or less. 2. The nuclear power generation system as set forth in claim 1, wherein the neutron poison micro-particle is a material including one or more selected from the group consisting of boron (B), gadolinium (Gd), argentums (Ag) and cadmium (Cd). 3. The nuclear power generation system as set forth in claim 1, wherein the neutron poison micro-particle is boron carbide or gadolinium carbide. 4. The nuclear power generation system as set forth in claim 1, wherein the neutron poison micro-particle is boron oxide or gadolinium oxide. 5. The nuclear power generation system as set forth in claim 1, wherein the neutron poison micro-particle is boron nitride or gadolinium nitride. 6. The nuclear power generation system as set forth in claim 1, wherein the neutron poison micro-particle is metal boride. 7. The nuclear power generation system as set forth in claim 2, wherein the boron comprises an isotope content of 19.9% or more of 10B. 8. The nuclear power generation system as set forth in claim 3, wherein the boron carbide comprises boron with an isotope content of 19.9% or more of 10B. 9. The nuclear power generation system as set forth in claim 4, wherein the boron oxide comprises boron with an isotope content of 19.9% or more of 10B. 10. The nuclear power generation system as set forth in claim 5, wherein the boron nitride comprises boron with an isotope content of 19.9% or more of 10B. 11. The nuclear power generation system as set forth in claim 6, wherein the metal boride comprises boron with an isotope content of 19.9% or more of 10B. 12. The nuclear power generation system as set forth in claim 2, wherein the gadolinium comprises an isotope content of 16.65% or more of 157Gd. 13. The nuclear power generation system as set forth in claim 3, wherein the gadolinium carbide comprises gadolinium with an isotope content of 16.65% or more of 157Gd. 14. The nuclear power generation system as set forth in claim 4, wherein the gadolinium oxide comprises gadolinium with an isotope content of 16.65% or more of 157Gd. 15. The nuclear power generation system as set forth in claim 5, wherein the gadolinium nitride comprises gadolinium with an isotope content of 16.65% or more of 157Gd. 16. The nuclear power generation system as set forth in claim 1, wherein the neutron poison micro-particles have an average diameter of 10 nm to 1 μm. 17. The nuclear power generation system as set forth in claim 6, wherein the metal boride comprises TiB2. |
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abstract | A bonded article including a single crystal cathode, for use in projection electron beam lithography, such as the SCALPEL(trademark) system. Because of its single crystalline structure, the single crystal cathode has only slightly misoriented grains. As a result, the single crystal cathode has few structural non-uniformities, and therefore a uniform emission characteristic. The single crystal cathode may be made of at least one of tantalum, tungsten, rhenium, and molybdenum. A local bonding technique for bonding a single crystal cathode with a conventional member. The local bonding technique does not recrystallize a center of the single crystal cathode, and therefore produces a bonded article which is usable in a projection electron lithography system, such as the SCALPEL(trademark) system. The local bonding technique may be laser welding and the single crystal cathode may be made of at least one of tantalum, tungsten, rhenium, and molybdenum. The member may be a conventional filament and the filament may be made of one of tungsten, a tungsten-rhenium alloy, and a tungsten-tantalum alloy. |
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description | The present invention refers generally to welding of nuclear components, in particular nuclear components made of zirconium based alloys. Particular difficulties arise when welding the end plugs to the cladding tube of a fuel rod. The difficulties include the material of which the end plugs and the fuel rods are made, namely zirconium based alloys, such as Zircaloy-1, Zircaloy-2, Zircaloy-4, ZIRLO, ZIRLO-B, etc. Zirconium based alloys have a high disposition for oxidation at the melting temperature. A further difficulty is that no atmospheric air or other surrounding gases are allowed to enter the interior of the fuel rod in connection with the final filling of hyperbaric helium, sealing and welding of the fuel rod. There are two methods of solving this problem. The fuel rod may be filled with fuel pellets and the rod be closed with end plugs. The end plugs are welded to the cladding tube in their final position without hyperbaric helium. Thereafter, fuel rod is filled with helium gas through a so called fill hole through one of the end plugs. The fill hole is then closed by means of a final welding operation. Alternatively, the fuel rod is filled with fuel pellets and helium gas prior to the final positioning and welding of the end plugs. The pressure prevailing in the fuel rods filled with helium is typically 5-10 bars for boiling water reactors, BWR, and 30-70 bars for pressurized water reactors, PWR. The ability to weld the end plug under these pressures in helium constitutes a further difficulty but eliminates the need for a filling hole and a welding step to close such a filling hole. FR-A-2 625 022 discloses a method of welding an end plug to a cladding tube of a nuclear fuel rod. The known method comprises the steps of attaching a lower end plug, filling the interior of the fuel rod with fuel pellets and helium gas, positioning the upper end plug to abut the upper end of the cladding tube: at an interface, and applying a laser beam of a laser source. The proposed laser is a pulsed laser. The laser beam is directed to a welding zone at the interface to melt material of the end plug and the cladding tube at the interface. U.S. Pat. No. 5,231,261 discloses a pulsed laser welding equipment for welding of fuel rods and monitoring the laser beam. The monitoring of the laser beam is done before the beam passes through the protective lens. It can therefore not identify changes to the protective lens e.g. from soot emanating from the welding process. The welding equipment setup for the girth welding is not done under pressure. The welding under helium pressure is done in a separate setup for a filling hole, thus not eliminating the need for such a filling hole and welding process step. U.S. Pat. No. 5,958,267 discloses a method for welding fuel rods under high pressure and to control the laser position with a video system. The method claims to prevent soot accumulating on a laser window and to limit plasma formation. In practice this is difficult to achieve and the process, without control of the plasma, is possibly unstable. U.S. Pat. No. 6,670,574 discloses a laser weld monitoring system for monitoring the welding of a pulsed laser beam. The system comprises two sensors, one sensor for sensing infrared radiation and one sensor for sensing reflection of the light of the laser beam. The method proposes essentially a trial and error method with a plurality of welds to correlate the complicated sensor curves to welding properties. U.S. Pat. No. 5,651,903 also discloses a system for monitoring laser welding by means of an equipment comprising two sensors arranged beside the optical path of the laser beam. A first sensor senses infrared radiation of the temperature of the weld and a second sensor senses ultraviolet radiation of the plasma of the weld. The electrical signals are used to monitoring variations compared to a predetermined anomaly values obtained by empirical testing. U.S. Pat. No. 6,710,283 discloses further laser weld monitoring system. The system comprises two sensors arranged beside the optical path of the laser beam. A first sensor senses reflected light of the laser beam and a second sensor, called a plasmatic sensor, senses the light emitted from welding zone. The method uses the frequency spectrum to compare actual variations in the signal with predetermined threshold values. An object of the present invention is to provide an improved method of welding a fuel rod, especially of welding end plugs without fill hole to a cladding tube of the fuel rod. This object is achieved by the method of welding a nuclear fuel rod including two end plugs, a cladding tube and a pile of fuel pellets in the interior of the cladding tube, the method comprising the steps of: bringing one of the end plugs and the cladding tube together to abut each other at an interface, and welding the end plug and the cladding tube by means of a welding equipment by applying a laser beam of a laser source of the welding equipment, the laser beam having a wavelength and being directed along an optical path of the welding equipment to a welding zone at the interface to melt material of the end plug and the cladding tube at the interface,characterised by the further steps of:sensing the welding by sensing radiation from the welding zone comprising:sensing radiation within a first wavelength range, which includes the wavelength of the laser beam coming from reflections from the welding zone,sensing radiation within a second wavelength range different from the first wavelength range, which includes infrared radiation from melted material in the welding zone, andsensing radiation within a third wavelength range different from the first wavelength range and the second wavelength range, which includes radiation from plasma in the welding zone, andmonitoring the welding and melting of material by monitoring the sensed radiations. By sensing different wavelengths of the radiation from the welding zone it is possible to monitor the quality of the weld and the joint of the end plug and the fuel rod. The three different wavelength ranges are independent of each other and provides different information. A possible deviation from a normal value of the radiation within any one of the three wavelength ranges may be used as an indication of an improper welding process enabling the operator to adjust the welding equipment. For instance, the operator may adjust the power of the laser source within an allowed range. The monitoring provides direct feedback during the welding process, and thus a warning if anything goes wrong or if there is a trend to be act upon. The first wavelength range, which may be sensed by a first sensor, includes the wavelength of the laser beam coming from reflections from the welding zone. The intensity of this sensed radiation is an indication of soot or other changes of the transmission of the protective lenses through which the laser beams passes. It also indicates changes of incoming laser power and changes of reflection on the work piece. The second wavelength range, which may be sensed by a second sensor, includes infrared radiation from melted material in the welding zone. The intensity of this radiation is thus an indication of the temperature and size of the melted material. This indicates also the effectiveness of the welding, i.e. the penetration of the weld. The third wavelength range, which may be sensed by a third sensor, includes radiation from plasma in the welding zone. The intensity of this radiation is an indication of the amount and the extension of the plasma formed during the welding. An increased signal from the plasma also indicate decreasing effectiveness of the welding, i.e. the penetration of the weld. Compared to earlier methods of monitoring, this method has a comparatively simple and straight forward interpretation and does not need a lot of trial and error weldings to find a welding characteristic to compare the curves with. In principal only one good weld is needed as a base to compare actual welding curves. Laser welding for joining end plugs to fuel rods has a plurality of advantages compare to other welding methods, such as electron beam welding and TIG-welding. The investment costs are low, for instance since it is possible to enclose an end section of the fuel rod to be welded. A relatively small enclosure may thus be used, and thus a relatively small floor space will be occupied by the welding equipment. It is possible to use one single laser source for sequential welding of a plurality of fuel rods in several welding equipments. Laser welding enables achievement of a smooth weld surface which is important when introducing the fuel rods in spacers of the fuel assembly. The alloy depletion is low ensuring proper corrosion resistance. There is no risk of tungsten contamination from a TIG seal welding of a fill hole. The sensing and monitoring may take place during the whole time period of the welding of the joint between the end plug and the cladding tube. The fuel rod is rotated during the welding so that the laser beam is moved relatively the fuel rod along the interface. The welding may be performed during one, two or more revolutions of the fuel rod. Typically, the velocity of the rotation may be about 1 revolution per second, which means that the welding of the end plug to the cladding tube may last for about 2 seconds. According to an embodiment of the invention, said reflections also includes reflections, or partial reflections, of the laser beam in the optical path, including protective lenses through which the optical beam passes. According to a further embodiment of the invention, the radiation of at least one of the first wavelength range, the second wavelength range and the third wavelength range is sensed along a direction being coaxial with the optical path at least in the proximity of the welding zone. The radiation to the different sensors may thus be diverted from the optical path to the interface. According to a further embodiment of the invention, the method also comprises viewing of the welding zone before, and/or during, the welding and melting of material by means of a video camera. It is thus possible for the operator to inspect the interface before the welding is initiated. Advantageously, the method may also comprise controlling the laser beam position relative the interface by means of the viewed interface. Furthermore, the viewing of the welding zone may take place along a viewing direction being coaxial with the optical path at least in the proximity of the welding zone. According to a further embodiment of the invention, the method also comprises the step of controlling the power of the laser beam in response to the sensed radiations. According to a further embodiment of the invention, the setup of the laser equipment is optimized and/or controlled using any anomalies of the signal curves from the three different wavelengths. Uneven signal curves compared to a reference signal curve from earlier approved welding test indicate uneven interface or wobble or dirt in the welding area and may lead to pores or uneven welding quality. According to a further embodiment of the invention, the monitoring step comprises monitoring the intensity of the radiation of the first wavelength range as a first signal level over time to form a first signal curve, monitoring the intensity of the radiation of the second wavelength range, as a second signal level over time to form a second signal curve and monitoring the intensity of the radiation of the third wavelength range as a third signal level over time to form a third signal curve. The setup of the welding equipment may be optimized and/or controlled using the signals levels from the three different wavelengths. According to a further embodiment of the invention, the method comprises the step of optimizing and verifying the setup of the welding equipment including the power level of the laser beam and the alignment of optical path, including lenses and/or protective lenses, with the signal level of the first wavelength range that includes the radiation of the reflection of the laser beam. During welding the same signal level may be used to control any changes of the transmitted laser beam, e.g. due to soot on the protective lens just above the welding zone. According to a further embodiment of the invention, the method comprises the step of controlling the focus position, effectiveness and/or penetration of the laser beam by the signal level of the second wavelength range that includes the infrared radiation from the melted material. An increased infrared signal may correspond to a deeper penetration of the weld. According to a further embodiment of the invention, the method comprises the step of controlling the effectiveness and the penetration of the welding by the signal level of the third wavelength range that includes the radiation from the plasma. An increased plasma signal may correspond to less penetration of the weld. According to a further embodiment of the invention, the method comprises the step of monitoring any anomalies of any of the signal curves from the three different wavelength ranges compared to a reference signal curve, for instance formed by earlier approved welding tests, to indicate uneven interface or wobble or dirt in the welding zone and/or possible occurrence of pores or uneven welding quality. According to a further embodiment of the invention, the laser beam is a continuous laser beam. The continuous laser can be generated by e.g. a Yb:YAG-fibre pumped by InGaAs diodes. Typically 500 W laser power is used. The present method is further simplified when used with a continuous laser beam which gives simple stable signals compared to embodiments with a pulsed laser. According to a further embodiment of the invention, the wavelength of the laser beam lies in the range 1050-1090 nm, preferably in the range 1060-1080 nm, for instance 1070 nm. According to a further embodiment of the invention, the second wavelength range is 1100-1800 nm. According to a further embodiment of the invention, the third wavelength is less than 600 nm, preferably 50-600 nm, more preferably 100-600 nm. The signal of the plasma is improved compared to other embodiments that uses only ultraviolet light below 390 nm. Thus, a preferred range of the third wavelength could be 390-600 nm, or 400-600 nm. According to a further embodiment of the invention, the welding takes place in a closed enclosure containing an atmosphere of helium at a pressure above the atmospheric pressure. A gas flow of typically 50 liter per minutes may be advantageous for limiting the plasma and soot formation and thus protecting the lens above the welding zone. The gas flow preferably enter the equipment below the protective lenses and flow coaxial with the laser beam passing the welding zone. By performing the welding in such a closed enclosure, the welding may be performed in only two steps, one step for the bottom plug and one for the top plug, without the need of a fill hole to be welded after the welding along the interface. Elimination of the fill hole will cut costs and risks in different ways. There will be lower costs for top end plugs with no fill holes. No separate fill hole weld station is needed. No fill hole weld inspection equipment is needed. There is obviously no yield loss for seal welding. The risk of tungsten contamination is eliminated. The welding may also be performed with other protective gases e.g. argon. When welding the final end plug this can only be done after a secured and sealed attachment of the end plug to the cladding tube as the interior of the fuel rod must contain helium. Welding with argon is cheaper but the welding is less stable and the needed welding power is higher due to a larger formation of plasma in the argon atmosphere. According to a further embodiment of the invention, the closed enclosure encloses the end plug and an end section of the cladding tube, wherein the method may comprise the preceding steps of: evacuating the interior of the cladding tube and the closed enclosure to a certain vacuum level during a predetermined time period, and then filling the closed enclosure and the interior of the cladding tube with helium to a predetermined pressure. Furthermore the method may comprise the steps of: prepositioning the end plug on the cladding tube at a determined distance from the cladding tube before the evacuating step, thereby permitting a free flow of gas from and to the interior of the cladding tube, and final positioning of the end plug on the cladding tube after the filling step and before the welding step. Advantageously, the prepositioning of the end plug at the determined distance is made by means of a mechanical stop, which may be displaceable to be introduced into the distance between the cladding tube and the end plug, and withdrawn therefrom. FIGS. 1-3 disclose a fuel rod 1 including a cladding tube 2 and two end plugs 3, one of which is disclosed. The fuel rod 1 includes an upper end plug 3 at the upper end of the cladding tube 2, and a lower end plug at the lower end of the cladding tube 2. The fuel rod 1 also includes a pile of fuel pellets 4 in the interior of the cladding tube 2. The fuel pellets 4 rest directly on the lower end plug. A so called plenum spring 5 is provided between the upper end of the pile of fuel pellets 4 and the upper end plug 3 to maintain the fuel pellets in a proper position in the cladding tube 2 and to ensure a plenum 6 for containing helium and fission gases generated during the fissile process in the nuclear reactor. The nuclear reactor may be a boiling water reactor, BWR, or a pressurized water reactor, PWR. The initial pressure prevailing in the fuel rod 1 filled with helium is typically 5-10 bars for a BWR, and 30-70 bars for a PWR. In the embodiments disclosed, the cladding tube comprises an outer tube 2′ and an inner tube 2″ a so called liner. FIG. 4 discloses a welding equipment for welding the end plug 3 to the cladding tube 2. The welding equipment includes a chuck 10 for holding and rotating the fuel rod and a closed enclosure 11 into which an end section of the fuel rod 1 is introduced via a passage, i.e. an end section of the cladding tube 2 and one of the end plugs 3. The chuck 10 is configured to rotate the fuel rod at a rotary speed of for instance 1 revolution per second. The closed enclosure 11 includes or is formed by a pressure resistant wall 12. A sealing 13 extends through the wall 12 to seal the passage for the fuel rod 1. A first positioning device 14 is provided to extend through the wall 12 at an end opposite to the sealing 13. The first positioning device 14 includes a movable piston 15 acting on the end plug 3 of the fuel rod 1 along the longitudinal direction of the fuel rod 1. Furthermore, a second positioning device 16 is provided to extend through the wall 12. The second positioning device 16 includes a mechanical stop 17 provided in the enclosure 11 to be displaceable along a transversal direction y, being transversal to the longitudinal direction x of the fuel rod 1 between a passive position shown in FIG. 4 and an active position shown in FIG. 1. The mechanical stop 17 when in the active position maintains the and plug 3 at a determined distance from the cladding tube 2 so that gases may be evacuated from the interior of the fuel rod and helium may be filled into the fuel rod 1. When the mechanical stop 17 is withdrawn to the passive position the end plug 3 may be brought to the final position and tight abutment to the cladding tube 2 by means of the first positioning device 14. Furthermore, the enclosure 11 includes a first protective lens 21 forming a part of the wall 12 of the enclosure 11. A second protective lens 22 is provided within the enclosure inside the first protective lens 21. The first protective lens 21 is relatively thick and configured to withstand the pressure prevailing in the enclosure 11. The second protective lens 22 is thinner than the first protective lens 21 and configured to protect the first protective lens against soot formed during the welding. A gas supply device 23 is provided for supplying a flow of gas, in the embodiments described helium, to the enclosure 11. The gas supply device 23 comprises a supply conduit 24 and an annular nozzle provided in the enclosure 11. The annular nozzle 25 is provided between the second protective lens 22 and the fuel rod 1, and extends around the second protective lens 22. The flow of helium gas to the enclosure may be for instance about 50 liter per minute. The gas supply device 23 is configured to provide a gas pressure in the enclosure equal to the gas that is to be achieved in the fuel rod when both the end plugs are secured and welded to the cladding tube 2. The welding equipment also includes a laser source 30 configured to generate a continuous laser beam. The laser source 30 may, for instance, comprise a Yb:YAG fibre laser with a wavelength in the range 1050-1090 nm, preferably in the range 1060-1080 nm, for instance 1070 nm. The YB:YAG-fibre may be pumped by InGaAs diodes. The laser source 30 transmits the laser beam via a fibre 31 to a primary optic 32. The primary optic 32 transmits the laser beam to a primary mirror 33 via a secondary semitransparent mirror 34. From the primary mirror 33 the laser beam is reflected and directed to the fuel rod 1 and a welding zone 36 at an interface 37 between the end plug 3 and the cladding tube 2. The laser beam thus extends along an optical path from the laser source 30 to the welding zone 36. The laser beam, reflected by the primary mirror 33, passes through at least one optical focusing lens 35, the first protective lens 21 and the second protective lens 22 along the optical path. The welding equipment also includes a sensing device including a first sensor 41, a second sensor 42 and a third sensor 43. During welding radiation from the weld zone 36 is transmitted to the sensors 41, 42 and 43 along the optical path through the second protective lens 22, the first protective lens 21 and the optical lens 35. The radiation is the reflected by the primary mirror 33 and the secondary mirror 34 away from the optical path of the laser beam. The radiation from the welding zone 36 is thus extending along a direction that is coaxial with the optical path in the proximity of the welding zone 36 at least along a straight line from the welding zone 36 or to the secondary mirror 34. The sensing device may be operated at a sampling frequency of up to 20 kHz. Via a first semitransparent mirror 44 the radiation is reflected to the first sensor 41. Via second semitransparent mirror 45 the radiation is reflected to the second sensor 42. The radiation passes through the semitransparent mirrors 44, 45 and 46 to the third sensor 43. The first sensor 41 is configured to sense radiation from the welding zone 36 within a first wavelength range, which includes the wavelength of the laser beam coming from reflections from the welding zone 36, i.e. wavelengths in the range 1050-1090 nm, preferably in the range 1060-1080 nm, for instance 1070 nm. The reflections from the welding zone 36, that are reflected via a second semitransparent mirror 45, also includes reflections, or partial reflections, of the laser beam in the optical path, including the first and second protective lenses 21, 22 and the at least one optical lens 35. The second sensor 42 is configured to sense radiation from the welding zone 36 within a second wavelength range different from the first wavelength range. The radiations are reflected to the second sensor via a third semitransparent mirror 46. The second wavelength range includes infrared radiation from melted material in the welding zone 36. The second wavelength range is 1100-1800 nm. The third sensor 43 is configured to sense radiation from the welding zone 36 within a third wavelength range different from the first wavelength range and the second wavelength range. The radiations to the third sensor passes through the semitransparent mirrors 44, 45 and 46. The third wavelength range includes radiation from plasma in the welding zone 36. The third wavelength is less than 600 nm, preferably 50-600 nm, more preferably 100-600 nm. The welding equipment also includes a monitoring device configured to monitor the welding and melting of material by monitoring the sensed radiations. The monitoring device comprises a processor 50 and a display 51 communicating with the processor 50. The sensors 41-43 communicate with the processor 50 which receives signals of the radiation of the three wavelength ranges from the sensors 41-43. The monitoring device is thus configured to monitor to an operator on the display the intensities of the wavelength ranges, i.e. the intensity of the radiation of the first wavelength range as a first signal level (in volt) over time (in seconds) to form a first signal curve 56, as illustrated in FIG. 5, the intensity of the radiation of the second wavelength range, as a second signal level (in volt) over time (in seconds) to form a second signal curve 57, as illustrated in FIG. 6 and the intensity of the radiation of the third wavelength range as a third signal level (in volt) over time (in seconds) to form a third signal curve 58, as illustrated in FIG. 7. The signal curves 56-58 can be saved by the processor for future use as reference or quality assurance. The signal curves 56-57 may be inspected or monitored to lie within an upper limit line L1 and a lower limit line L2. The upper and lower limit lines L1, L2 may represent a deviation of 15%, or preferably of 10%, from a desired signal level. The signal curves 56-57 may also, or alternatively, be inspected or monitored in relation to a reference line R representing a desired signal level. The line R can alternatively be a saved reference curve from one good weld. The line R can alternatively be a saved reference curve from a mean value curve from several good welds. Moreover, the welding equipment includes a viewing device configured to enable viewing of the welding zone before, and/or during, the welding and melting of material. To that end the viewing device comprises a video camera 61, a processor 62 and a display 63. If the viewing device is to be used before welding, the enclosure 11 may be illuminated by e.g. LED light. The viewing of the welding zone 36 may take place along a viewing direction being coaxial with the optical path in the proximity of the welding zone 36 at least along a straight line from the welding zone 36 or up to the secondary mirror 34. Thanks to the viewing device, the laser beam position relative the interface 37 may be controlled manually by the operator when inspecting the interface 37 on the display 63 or automatically. The welding equipment may also include means for controlling the power of the laser beam in response to the sensed radiations by means of the processor 50 controlling the output of the laser source 30. The controlling may be performed manually by the operator when inspecting the signal curves shown on the display 51 or automatically. The controlling may, preferably as an initial measure, comprise the step of verifying the setup of the welding equipment including the power level of the laser beam and the optical path with the signal level of the first wavelength range that includes the radiation of the reflection of the laser beam. During welding, or between the welding of the fuel rods 1, the controlling may include the steps of: controlling the focus position of the laser beam by the signal level of the second wavelength range that includes the infrared radiation from the melted material, and/or controlling the effectiveness and the penetration of the welding by the signal level of the wavelength that includes the radiation from the plasma. During welding it is also possible to monitor any anomalies any of the signal curves from the three different wavelength ranges compared to a reference signal curve to indicate uneven interface or wobble or dirt in the welding zone and/or possible occurrence of pores or uneven welding quality. The method and the welding equipment enable achievement of a smooth and uniform weld. The shape of the finished weld W is illustrated in FIG. 3. As can be seen, the surface of the surrounding weld W is even with the surface of the cladding tube 2 and the end plug 3. Welding of a fuel rod 1 may include the following steps: prepositioning a bottom end plug to a bottom end section of a cladding tube 2, introducing the bottom end section into the enclosure 11 and holding the fuel rod 1 by means of the chuck 10, activating the first positioning device 14 to press the bottom end plug against the cladding tube 2, evacuating the enclosure 11 to a certain vacuum level during a predetermined time period, rotating the fuel rod 1 by means of the chuck 10, inspecting the position of and positioning the interface 37 with the aid of the viewing device, initiating the welding by the laser source 30, monitoring the signal curves 56-58 illustrating the intensity of the three wavelength ranges, removing the fuel rod 1 from the enclosure 11, prepositioning a top end plug 3 to a top end section of a cladding tube 2, introducing the top end section into the enclosure 11 and holding the fuel rod 1 by means of the chuck 10, activating the second positioning device 16 to bring the mechanical stop 17 into contact with the top end section to ensure the determined distance between the top end plug 3 and the cladding tube 2, evacuating the enclosure 11 a certain vacuum level during a predetermined time period, filling the enclosure 11 and the interior of the fuel rod 1 with helium to a desired predetermined pressure, removing the mechanical stop 17, activating the first positioning device 14 to press the top end plug 3 against the cladding tube 2, rotating the fuel rod 1 by means of the chuck 10, inspecting the position of and positioning the interface 37 with the aid of the viewing device, initiating the welding by the laser source 30, monitoring the signal curves 56-58 illustrating the intensity of the three wavelength ranges, and removing the fuel rod 1 from the enclosure 11. The present invention is not limited to the embodiments and descriptions given above but may be varied and modified within the scope of the following claims. |
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claims | 1. A method for computer-aided analysis of a reliability of a technical system with a plurality of technical components, comprising:describing the reliability of each of the plurality of technical components by a component function ƒ for each component under consideration of the plurality of technical components, wherein the component function is described by a distribution function comprising parameter t representing an operating time of the technical system, parameter t0 representing a maintenance interval for the technical component under consideration, and parameter λ representing a parameter that describes the reliability of the component under consideration and wherein the distribution function comprises an exponential function represented as ƒ(t, t0)=1−e−λ(tmodt0) wherein tmodt0 represents a modulo function;determining a system reliability for the technical system in accordance with a system reliability function F comprising a mathematical combination of the reliabilities of the components;determining for each of at least some of the components a change magnitude which is a measure of the change in the system reliability as a function of the change in the maintenance interval for the component under consideration; anddetermining, using a computer, a factor for the effect of the component concerned on the system reliability via the change magnitude for each of at least some of the components,wherein, based on the factors determined, at least one component is identified having a greater effect on the system reliability than the other. 2. The method as claimed in claim 1, wherein the maintenance interval for a component under consideration differs from maintenance intervals for other components. 3. The method as claimed in claim 1, wherein the reliabilities of the components are variables which characterize the probability of a fault and/or failure and/or the availability of the component concerned. 4. The method as claimed in claim 1, wherein the component functions comprise at least one probability distribution function. 5. The method as claimed in claim 4, wherein the probability distribution function is selected from the group consisting of: a Weibull distribution, a gamma distribution, a lognormal distribution, an exponential distribution and combinations thereof. 6. The method as claimed in claim 1, wherein the mathematical combination of the reliabilities of the components comprises using a Boolean algebra. 7. The method as claimed in claim 1, wherein each of the change magnitudes comprises the derivative of the system reliability with respect to the maintenance interval concerned. 8. The method as claimed in claim 1, wherein the factors for a plurality of the components comprise in each case an integral, with respect to the maintenance interval of the change magnitude for the component concerned. 9. The method as claimed in claim 1, wherein the factors for each of the plurality of the components comprise the maximum value of the absolute value of the change magnitude for the component concerned in an interval of the operating time of the technical system. 10. The method as claimed in claim 1, wherein the factors for each of the plurality of the components comprise the absolute value of the change magnitude for the component concerned at a predefined value of the operating time of the technical system. 11. The method as claimed in claim 1, wherein the reliability of a power station, and/or of a technical control system is analyzed. |
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abstract | A seal plate for repairing damaged areas in a pressure vessel cladding includes a first side and an opposing second side and a sealing portion located on the second side, with the sealing portion circumscribing the seal plate. The seal plate also includes a seal lip extending from a periphery of the sealing portion and a cavity located in the second side, with the cavity circumscribed by the sealing portion. The seal plate further includes at least one purge port extending from the first side to the cavity. Each purge port is in fluid communication with the cavity. |
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summary | ||
051006081 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIGS. 1 through 5 are views illustrating an embodiment of in-core nuclear instrumentation for fast breeder reactors in accordance with the present invention, in which FIG. 1 is view illustrating a backup shutdown rod and shows in detail the installation location of a neutron flux measuring unit, FIG. 2 is a view illustrating a control rod guide tube, FIG. 3 is a view illustrating a control rod element, FIG. 4 is a view showing the overall construction of in-core nuclear instrumentation for fast breeder reactors in accordance with the invention, and FIG. 5 is a view showing the arrangement of a core in a fast breeder reactor. Shown in the Figures are a backup shutdown rod 1, a neutron flux measuring unit 2, a control rod guide tube 3, a protecting tube 4, a buffer (dashpot) 5, a buffer (dash ram) 6, a lower spacer pad 7, a latch spring 8, an entrance nozzle 9, a control rod guide tube handling head 10, an upper spacer pad 11, a control rod handling head 12, a connecting shaft 13, an intermediate spacer pad 14, a control rod element 15, an instrumentation cable guide tube 16, an upper end cap 17, a cladding tube 18, a plenum spring 19, a spacer 20, a neutron absorber pellet 21, a wire spacer 22, a reactor vessel chamber wall 23, a core 24, an upper mechanism 25, a reactor vessel 26, a control rod drive mechanism 27, a fine rod 28, a coarse adjustment rod 29, a core fuel assembly 30, a breeding fuel assembly 31, and a neutron shield 32. The general features of the construction of the present invention will now be described with reference to FIGS. 4 and 5. As shown in FIG. 5, the core 24 is so designed that the control rods, such as the fine rods 28 and coarse rods 29, are distributed uniformly within the fuel rods of the core and have substantially the same shape as the core fuel assemblies 30 and breeding fuel assemblies 31. The neutron shield 32 is provided outside these control rods. The backup shutdown rod 1, fine rods 28 and coarse rods 29 are all control rod assemblies for controlling the reactivity of the reactor. Ordinary start-up and shutdown of the reactor is performed by the fine and coarse rods. When it is impossible for both the fine and coarse rods to effect emergency shutdown of the reactor, this is carried out by the backup shutdown rod 1, the system whereof is different from that of the fine and coarse rod and which is capable of performing shutdown independently. As shown in FIG. 4, these control rod assemblies are operated by the control rod drive mechanism 27 via the core upper mechanism 25 at the upper part of the core. In accordance with the present invention, the neutron flux measuring unit is arranged in one or a plurality of the control rods such as the backup shutdown rod, coarse rods and fine rods in order to measure neutron flux. By adopting such an arrangement, a change in the neutron flux within the core assemblies can be ascertained with a quick response since the control rods themselves occupy a single assembly portion of the fast breeder reactor and are distributed throughout the core. The location at which the neutron flux measuring unit 2 is installed will now be described in detail with reference to FIGS. 1 through 3. Each control rod element 15 (FIG. 3) comprises the cladding tube 18 accommodating a stack of the neutron absorber pellets 21. These control rod elements 15 are clustered within the protecting tube 4 (FIG. 2). As shown in FIG. 1, the connecting rod 13 is attached to the upper portion of the protecting tube 4 for being connected to the control rod handling head 12, which is adapted to be connected to the control rod drive mechanism 27 that carries out insertion and withdrawal. A dash ram 6 serving as a buffer is connected to the lower portion of the protecting tube 4. A dashpot 5 serving as a buffer is provided in the lower portion of the control rod guide tube 3, which accommodates the protective tube 4 that separates and drops from the control rod drive mechanism 27 in the event of an emergency shutdown. The dash ram 6 and dashpot 5 decelerate and stop the fall of the protecting tube 4, which includes the control rod elements 15. The neutron flux measuring unit 2 is mounted inside the dash ram 6. The control rod handling head 12, which is for being attached to the control rod drive mechanism 27 in order to insert and withdraw the control rod guide tube 3, is connected to the upper portion of the rod guide tube 3. The entrance nozzle 9, which is for introducing a reactor coolant flow into the control rod guide tube 3, is attached to the lower portion of the guide tube 3. Each control rod element 15 comprises the cladding tube 18, which is made of stainless steel, in which are stacked the neutron absorber pellets 21 consisting of boron or the like. The upper portion of the control rod element 15 is provided with the upper end cap 17, and the lower portion is provided with the lower end cap 17' and the wire spacer 22. These end caps 17, 17' are for fixing and supporting the control rod elements in the protecting tube 4 and serve to plug the ends of the cladding tube 18. The plenum spring 19 and spacer 20 for setting and retaining the neutron absorber pellets 21 are provided inside the control rod element 15. In the embodiment described above, the neutron flux measuring unit 2 is arranged inside the buffer (dash ram) 6. However, the invention is not limited to this embodiment, for it is possible to select any position in the axial direction for measurement purposes by appropriately choosing the location of installation. It is possible to grasp the in-core neutron flux distribution with sufficient accuracy not only in a homogeneous core but also in an axially heterogeneous core and diametrically heterogeneous core. Furthermore, though the measurement position can be fixed as much as possible by disposing the neutron flux measuring unit in the backup shutdown rod 1, the invention is not so limited. The neutron flux measuring unit can be disposed in other control rod assemblies such as the coarse rods or fine rods, or neutron flux measuring units can be dispersed throughout the core by making joint use of a plurality of these control rods. The neutron flux measuring unit used should at the very least be one which does not readily deteriorate during the control rod assembly exchange interval. As shown in FIG. 2, an instrumentation cable for the neutron flux measuring unit is passed through the instrumentation cable guide tube at the center of the backup shutdown rod and is drawn out at the upper mechanism of the core. This makes it possible to minimize the amount of neutron irradiation received by the instrumentation cable. Thus, in accordance with the present invention as described above, the control rods themselves occupy a single assembly portion in the core and one or a plurality of the neutron flux measuring units are arranged in the control rod assemblies dispersed within the core. As a result, it is possible to reliably and accurately grasp, with a quick response, a change in in-core neutron flux even in a large-size heterogeneous core of a demonstration-class reactor. Furthermore, by using a control rod assembly as a backup shutdown control rod, the measurement position can be fixed to the greatest extent possible. Moreover, the amount of neutron irradiation sustained by the instrumentation cable can be minimized by passing the cable through the center of the control rod assembly. As many apparently widely different embodiments of the present invention can be made without departing from the spirit and scope thereof, it is to be understood that the invention is not limited to the specific embodiments thereof except as defined in the appended claims. |
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description | The United States Government has rights in this invention pursuant to Contract No. W-7405-ENG-48, between the United States Department of Energy and the University of California for the operation of the Lawrence Livermore National Laboratory. The embodiments of the present invention relates generally to radioisotopic sources and, more particularly, to a source of gamma rays and neutrons for providing a radiation spectrum similar to that of weapons-grade plutonium without the use of special nuclear material. Radiation detection technology is being deployed worldwide to address concerns regarding the illicit movement of radiological and nuclear materials. Equipment of different types and from various manufacturers is being distributed to operators with varying levels of training and different backgrounds. There is a need to reliably exercise and demonstrate the capabilities of these detectors and responders. In particular: (1) many detector developers, manufacturers, and vendors do not have weapons-grade plutonium, WGPu, for testing their hardware or isotope identification algorithms; (2) since the identification of shielded or masked plutonium depends on the plutonium radiation intensity and spectrum, a high-fidelity surrogate exhibiting the full WGPu spectrum is needed to test the effects of shielding and masking in different shielding configurations; (3) fixed-site radiation detection equipment (ports, border crossings, etc.) requires in situ testing capability, and (4) nuclear incident response exercises require credible materials. The use of Nuclear Explosive-Like Assemblies (NELAs) is not always an attractive option for the stated applications, since NELAs typically contain actual SNM combined with inert materials (or conversely, high-explosives combined with non-radioactive materials), and their use is limited to secure facilities. The use of a NELA is prohibitive due to cost, safety and security concerns for all but the most pressing needs. By contrast, a non-SNM surrogate can be transported and deployed without the substantive administrative controls required for SNM. Accordingly, it is an object of the embodiments of the present invention to provide a radiation surrogate having a neutron and gamma-ray signature which is representative of the neutron and gamma-ray spectrum of weapons-grade plutonium at an energy resolution of 5% without the use of special nuclear material. Another object of the embodiments of the present invention is to provide a radiation surrogate having a neutron and gamma-ray signature which is representative of the gamma-ray spectrum of weapons-grade plutonium at an energy resolution 5% over an energy range of 59 keV to 2614 keV without the use of special nuclear material. Still another object of the embodiments of the present invention is to provide a radiation surrogate having a neutron and gamma-ray signature which is representative of the gamma-ray spectrum of weapons-grade plutonium at an energy resolution 5% over an energy range of 59 keV to 2614 keV without the use of special nuclear material, and having low α-particle emission. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects, and in accordance with the purposes of the present invention, as embodied and broadly described herein, the radiation surrogate for weapons-grade plutonium, includes in combination: the radioisotopes Ba-133 having all activity of between about 5 and about 5.5 μCi, Cf-252 having an activity of between about 4 and about 5 μCi, Cs-137 having an activity of between about 10.2 and about 10.4 μCi, Gd-153 having an activity of between about 350 and about 550 μCi, Lu-177 m having an activity between about 40 and about 50 μCi, Sn-113 having an activity between about 13.5 and about 30 μCi, and Zr-95 having an activity between about 1 and about 6 μCi. The embodiments of the present invention overcome the disadvantages and limitations of the prior art, and benefits and advantages thereof include, but are not limited to, providing a neutron and gamma ray source that represents the gamma-ray spectrum of weapons-grade plutonium at 5% energy resolution between 59 keV and 2614 keV without containing special nuclear material and α-particle emitters, and in a form which is easier to deploy than nuclear explosive-like assemblies or small quantities of plutonium while meeting Department of Transportation Limited Quantity requirements. The embodiments of the invention do not require replacement of radioisotopes more frequently than about three-month intervals. Briefly, embodiments of the present invention includes an apparatus for providing a neutron and gamma-ray source that represents the gamma-ray spectrum of weapons-grade plutonium at 5% energy resolution between 59 keV and 2614 keV without containing special nuclear material and without significant α-particle emission, and which meets Department of Transportation Limited Quantity requirements, while reliably yielding plutonium isotope identification by current and next-generation identification equipment and algorithms. Reference will now be made in detail to the present embodiments of the inventions, examples of which are illustrated in the accompanying drawings. In the FIGURES, similar structure will be identified using identical callouts. Turning now to FIG. 1A, illustrated is a schematic representation of an embodiment of the surrogate plutonium radiation source of the present invention, 10, showing cylindrical vessel, 12, having a bore, 13, containing radioactive disk sources, 14, and spherical source holder, 16, having base, 18, into which cylindrical vessel 12 is disposed. FIG. 1B is a schematic representation of the embodiment of the present invention illustrated in FIG. 1A hereof showing cylindrical vessel 12 removed from spherical source holder 16. Spherical source holder 16 has a cylindrical passage, 20, therein adapted to receive cylindrical vessel 12. Base 18 has flat surface, 22, concave surface, 24, and counterbored or countersunk holes, 26, which align with threaded holes, 28, in spherical holder 16 such that base 18 can be secured thereto by the use of screws (not shown in FIG. 1A and FIG. 1B). After cylindrical vessel 12 is disposed inside of and secured to spherical holder 16 using screws (not shown in FIG. 1A and FIG. 1B), inserted through counterbored holes, 30, in flange portion, 32, and into matching threaded screw holes, 34, in flat portion, 36, adapted to receive said screws, and base 18 is attached thereto, the plutonium surrogate 10 may be rotated by 180° for deployment such that flat surface 22 of base 18 may be placed on a flat surface. Use of cylindrical vessel 12 permits radioisotope sources 14 to be replaced when the half-lives of the radioisotopes employed no longer permit acceptable activities to be obtained therefrom. FIG. 2 shows a schematic representation of cylindrical vessel 12 illustrated in FIGS. 1A and 1B with one of radioactive wafer sources 14 removed and disassembled. Open end, 38, of cylindrical vessel 12, is closed, after radioactive sources 14 are placed therein, using screw cap, 40, having threaded portion, 42, which screws into matching threaded portion 44 of vessel 12. Wafer sources 14 include cylindrical radioactive disk, 46, which is disposed in the inner opening of washer, 48, for support, and lower and upper spacers, 50, and, 52, respectively, provide spacing between disks 46. Solid portion, 54, of cylindrical vessel 12 is adapted to engage lower spacer 50 of the lowest radioactive disk 14 such that the group of radioactive sources 14 is approximately centered within spherical holder 16 when cylindrical vessel 12 is inserted therein, As will be described hereinbelow, the stack of radioactive wafer sources was wrapped with a thin tungsten foil (not shown in FIG. 2) before insertion into bore 13 of cylindrical vessel 12. FIG. 3 shows a comparison gamma-ray spectrum of an embodiment of the surrogate radiation source of the present invention (thin line) as a function of the photon energy, compared with a plutonium spectrum (heavy line), both spectra being measured using a handheld sodium-iodide radioisotope identification device. Neutron fluxes appropriate for the gamma-radiation fluxes are provided by fission neutrons from the Cf-252 radioisotope. Individual radioisotopes were commercially obtained from Isotope Products Laboratories as sealed, Type D Disks having similar geometries. Disks 46 having about a 1-in. diameter and a thickness of approximately 0.25 in, were stacked in 1.8-in. outer diameter×6-in. long transparent polycarbonate cylindrical vessel 12, using washers 48 and spacers 50 and 52 fabricated from 0.25-in. thick polycarbonate to adjust source spacing and to prevent the movement of the individual sources. A spherical geometry was chosen for uniform attenuation of the gamma-ray spectrum (assuming that radioisotope sources are placed at the center of the sphere. Requirements for dose rate, radiation attenuation, weight, and transparency were satisfied by a sphere having radius of 4.25-in. composed of clear or transparent plastic (thermoset) resin. That is, this diameter provides the appropriate “stand-off” distance from the radioactive sources to the surface of the ball to achieve a contact dose rate below 5 mRem/h, and the spherical shape approximates the design of an isotropically shielded surrogate source. The transparent material allows visual confirmation of the presence of the sealed sources. As an example, an inexpensive, clear, commercially available, off-the-shelf bowling ball 16 (the Lane Hawk “Clear Ball”) was employed, since it provides the added benefits of the easy re-supply and replacement, and the wide availability of different types of hard and soft carrying cases. Further, the costs of annealing and machining are low, and a simple removable handle made from high-quality nylon webbing, and a rubber grip can be used to carry the surrogate. A 2-in. diameter, 6-in. long cylindrical radial bore 20 was machined into the sphere. As stated hereinabove, inner sleeve 12 and radioisotope sources 14 were placed in bore 20 and the bore sealed with a plastic cylinder plug 40 secured with shoulder bolts. Different configurations of radioisotopes were examined for the surrogate. The four combinations shown in TABLES 1-4 were found to closely satisfy the requirements of a useful surrogate weapons-grade plutonium radiation source. TABLE 1Radioisotopic sources used for surrogate configuration 6.Activity whenIsotopeInitial activity (μCi)Half-life (days)tested (μCi)Ba-1335.413836.155.36Cf-2525.00965.434.90Cs-1375.2110975.555.19Cs-1375.1810975.555.16Gd-153515.40240.40508.02Lu-177m47.17160.4041.62Sn-11321.24115.0915.07Sn-11320.46115.0914.52Zr-9510.2664.025.54 TABLE 2Radioisotopic sources usedfor surrogate configuration 8 including tungsten foil wrapping.Activity whenIsotopeInitial activity (μCi)Half-life (days)tested (μCi)Ba-1335.413836.155.30Cf-2525.00965.434.68Cs-1375.2110975.555.17Cs-1375.1810975.555.14Gd-153515.40240.40424.88Lu-177m47.17160.4031.84Sn-11321.24115.0910.37Sn-11320.46115.099.99Zr-9510.2664.022.83 TABLE 3Radioisotopic sources usedfor surrogate configuration 9 including tungsten foil wrapping.Activity whenIsotopeInitial activity (μCi)Half-life (days)tested (μCi)Ba-1335.413836.155.24Cf-2525.00965.434.48Co-5752.65271.7933.10Cs-1375.2110975.555.15Cs-1375.1810975.555.12Gd-153515.40240.40354.32Lu-177m52.89160.4046.06Sn-11321.24115.097.10Sn-11320.46115.096.84Zr-9510.2664.021.43 TABLE 4Radioisotopic sources usedfor surrogate configuration 10 including tungsten foil wrapping.Activity onAug. 2, 2006IsotopeInitial activity (μCi)Half-life (days)(μCi)Ba-1335.413836.155.24Cf-2525.00965.434.48Co-5752.65271.7933.10Cs-1375.2110975.555.15Cs-1375.1810975.555.12Gd-153515.40240.40354.32Lu-177m52.47160.4045.69Sn-11321.24115.097.10Sn-11320.46115.096.84Th-2284.20697.734.07Zr-9510.2664.021.43 A tungsten foil (0.05-mm thickness) was used to simulate the 59.5 gamma-rays from americium-241 (a daughter product due to the beta-decay of plutonium-241), since elemental tungsten emits 59.3-keV fluorescence x-rays if stimulated by higher energy photons. The use of foil reduces the self-attenuation of the fluorescence x-rays in the tungsten. In the present case, 80-keV gamma-rays emitted by the barium-133 source provide a means for inducing the x-ray fluorescence response. It might be beneficial to consolidate some of the individual radioisotope sources into single sealed-source, based on similarity of half-lives. For example, Sn-113 might be combined with Lu-177m, and Gd-153 with Co-57. Based on their relatively long half-lives, Eu-155, Ba-133 and Cs-137 might be combined. Typically, Cf-252 is sealed in a different manner than gamma-beta sources, and may not be practically combined with the other isotopes. Measurements were performed at a distance of 1 m for 55 s from the center of spherical source holder 16 of plutonium surrogate 10. Detectors were positioned in the “equatorial plane” of the spherical source holder, relative to the vertical axis of the source cylinder. In each set, 10 individual measurements were made with each radioisotope identification device (GR-135 and ThermoElectron IdentiFinder-U). The results are set forth in TABLE 5. TABLE 5Summary of radiation measurements for two surrogate configurations.Configuration 6Configuration 8No. ofNo. ofoccurrencesoccurrencesDetectorIdentification(out of 10)Identification(out of 10)GR-135Pu-23910Pu-2397Unknown10Unknown6IdentiFinderPu-23910Pu-2399Ga-678Cs-1379Not in Library1 Plutonium was identified in the majority of the measurements (between 70 and 90%). Radiation measurements were made using surrogate configuration 6 which yielded indications of Pu-239 accompanied by an “unknown” on 10 consecutive measurements using the GR-135 detector. On the same day, measurements using the same surrogate but with the IdentiFinder detector yielded indications of plutonium on 10 consecutive measurements, eight of which were accompanied by indications of the presence of gallium-67. Two months later configuration 8 yielded three indications of Pu-239 only, four indications of plutonium-239 and “unknown” and two indications of “unknown” only using the and the GR-135 detector. On the same day, measurements of surrogate configuration 8 yielded nine instances of indication of plutonium-239 accompanied by cesium-137, and a single instance in which the IdentiFinder detector indicated “Not In Library.” As may be observed in FIG. 3, the surrogate spectrum also yielded a good visual approximation of the weapons-grade plutonium spectrum across relevant energies. Measurements of the surrogate (configurations 6 and 8) were also performed using an adaptable radiation area monitor (ARAM) employing a 4-in.×4-in.×16-in. Nal, gamma-ray detector, He-3 tubes, and the autoGadRas isotope identification software. The surrogate was rolled past the ARAM at a distance of closest approach of about one meter, which consistently yielded an identification of plutonium for the surrogate. Spectra from the surrogate for various configurations were also measured using an ORTEC Detective which employs high-purity germanium (HPGe). These measurements were intended to confirm the actual isotopic composition and activities of the surrogate. In configurations 8 and 9 which included the surrounding layer of tungsten foil to produce 59.3 photons, yielded an indication of plutonium-239 on the Detective after 2-3 min, of measurement time at a distance of about 30 cm. The dose rate from the prototype has been modeled in full, three-dimensional geometry, including disc sources, plastic spacers and spherical resin sphere. At a radius of 30 cm from the center of the sphere, the dose rate is estimated at a conservative maximum value of 2.8 mRem/h which is below the desired limit of 5 mRem/h. Approximately one-third of the dose is imparted by neutrons. It is reasonable to estimate that the 30 cm standoff from the surface of the spherical container is equivalent to the dimensions of the shipping container that will be used with the prototype. Therefore, in order to affect a dose rate less than 0.5 mRem/h at the surface of a shipping container, the dose rate must be attenuated by a factor of one-sixth using shielding materials alone. This attenuation is approximately equivalent to two mean-free paths of any chosen shielding material (The upper limit on dose rate is determined by situating the particular source disks that contribute the most doses at the outside of the disk stack to minimize self-shielding.). TABLE 6 shows a sample determination of whether a surrogate meets DOT Limited Quantity requirements. TABLE 6Typical spreadsheet entry to determine ifconfiguration meets DOT Limited Quantity requirements.FractionalProposedcontributionIsotopeA2 (Ci)A2/1000 (Ci)activity (mCi)to limitBa-133810.0810.005204916.426E−05Cf-2520.0810.0000810.0042787945.282E−02Co-572700.270.0302763761.121E−04Cs-137160.0160.0051361433.210E−04Cs-137160.0160.0051065513.192E−04Eu-155810.0810.2825595163.48SE−03Gd-1532400.240.315484221.315E−03Lu-177m0.540.000540.0195795493.626E−02Sn-113540.0540.0057502761.065E−04Sn-113540.0540.0055391081.026E−04Th-2280.0270.0000270.0041.481E−01Zr-95220.0220.0009785424.44SE−050.6838939872.431E−01Total mCiConsignmentA2 Fraction In general, alpha-emitting radioisotopes are assigned lower regulatory limits on activity. In TABLE 6, this is apparent in the large fraction of the consignment activity fraction attributable to Th-228, even though the activity of the thorium is relatively low when compared with other isotopes. The remainder of the consignment fraction is largely attributable to the high activities of Gd-153 and Eu-155. The total consignment A2 fraction is approximately 25% which indicates that the activity of the surrogate could be increased by a factor of up to four and still meet DOT Limited Quantity requirements. 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, and obviously many modifications and variations are possible in light of the above teaching. The embodiments were chosen and described in order to best explain the principles of the invention and its practical application to thereby enable others skilled in the art to best utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto. |
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description | This application is a divisional of U.S. patent application Ser No. 12/153,397, filed May 19, 2008, which is a divisional of U.S. patent application Ser. No. 10/511,396, filed on Mar. 23, 2005, now U.S. Pat. No. 7,391,036, issued Jun. 24, 2008, which is a §371 of International Application No. PCT/JP03/04910 filed on Apr. 17, 2003, which claims the priority of Japanese Application Nos. 114756/2002 and 129515/2002, filed on Apr. 17, 2002 and May 1, 2002, respectively, all of which are incorporated herein by reference. The present invention relates to a surface inspection apparatus and method for effectively performing evaluations, observations, analyses and the like such a structure inspection, an observation in an enlarged view, a material evaluation, an electric conduction and so forth of a sample surface. More particularly, the present invention relates to a surface inspection apparatus and method which can detect defects on high density patterns having, for example, a minimum line width of 0.15 μm or less on a sample surface, with high accuracy, high reliability and high throughput. The present invention also relates to a semiconductor device manufacturing system which incorporates such a surface inspection apparatus, and a semiconductor device manufacturing method which employs such a surface inspection apparatus to inspect patterns in the midway of a semiconductor device manufacturing process and/or after completion of the process. Conventionally, when a surface of a sample, such as a substrate or a wafer, is observed or subjected to material inspection, or is inspected or evaluated with respect to a structure or electric continuity of electric circuits formed thereon, it is known to use a surface inspection apparatus in which a defect in a surface of a sample is detected by emitting a charged particle beam (a primary charged particle beam), such as an electron beam, to the surface for scanning, detecting secondary charged particles emitted from the surface, producing image data from the detected secondary charged particles, and comparing the produced image data with the image data for each die (or chip). Such surface inspection apparatuses including one that utilizes a scanning electron microscope (SEM), and surface flattening apparatuses for flattening a surface of a sample, such as a substrate, exist as independent apparatuses and have been conventionally used. In such a conventional SEM-based system as described above, and a system which simultaneously illuminates a wide area on a wafer, as a wafer under inspection is irradiated with an electron beam, the wafer is charged. The overcharged wafer Would cause distorted image data and obscure images. In addition, a normal pattern may be erroneously evaluated as defective. Another grave problem in the prior art is damages to a sample. Specifically, when an electron beam is irradiated to a surface of a wafer, the surface is charged by the irradiated beam, permitting acquisition of an image representative of a potential contrast. However, the wafer may be charged in a different condition depending on insulating material, metal conductive materials, circuit resistance, and the like. Therefore, occasionally, an extreme potential difference may be produced on a boundary of patterns, resulting in failed acquisition of secondary electrons emitted from the wafer surface or in arc discharge. The above-mentioned problems will be described in greater detail. Secondary electron emission characteristics differ from one sample to another depending on the energy of an incident irradiated beam thereon and the characteristic of the sample surface. FIG. 1 is a graph showing an exemplary relationship between beam energy and secondary electron emission efficiency η when an insulating material is irradiated with an electron beam. With η larger than one, more electrons than incident electrons are emitted from the insulating material, and hence the surface of the insulating material is positively charged (Region P). On the other hand, with η smaller than one, the surface is negatively charged (Region Q). This may cause damages such as breakdown in some samples, the characteristics of which depend on the circuit configuration and layered structure thereon. Specifically, when an insulating material is applied with an electric field equal to or larger than a breakdown strength (for example, 50-1000 kV/mn), the insulating material loses the insulating property to cause a breakdown, resulting in a current flowing therethrough. On the other hand, when an excessive amount of charge is accumulated on the insulating material, the field strength exceeds a breakdown voltage, thereby resulting in a breakdown. Also, once a breakdown occurs, an excessive current flows to break the circuit, and the insulating material may no longer restore its insulating property. Further, when the magnitude of a beam current is decreased so as to minimize distortion of the image data due to the electric charge, an S/N ratio of a signal resulting from a secondary charged particle beam, which is emitted from the surface of the sample by emission of a primary charged particle beam, becomes undesirably low. This increases the possibility of false detection. The problem of lowering of the S/N ratio may be lessened and the possibility of false detection may be reduced, by effecting scanning at a plurality of times and conducting an averaging operation. However, the throughput in such an inspection apparatus becomes low. Further, in order to detect fine defects, a large-current emission beam is required. For example, when it is assumed that the amount of signal required for determining a defect having a 2×2 pixel size of a CCD is 1, the amount of signal required for determining a defect having a 1×1 pixel size is 4. That is, for detecting a fine defect by using the same detector, the magnitude of a beam current must be increased so as to increase the amount of secondary electrons. However, when the magnitude of a beam current is increased, as is described above, the electric charge on the sample becomes high, thus increasing distortion of the image. In order to solve the above-mentioned problems, the Applicant proposed, in Japanese Patent Application No. 2000-340651 (published as Japanese Patent Public Disclosure (Kokai) No. 2002-148227), a method in which a resistive film is coated on the surface of a sample before emission of a charged particle beam to the surface. In this method, however, it is difficult to form a thin resistive film having a uniform thickness on each sample. Further improvements are required to be made. Although a technique of coating a resistive film or coat on the surface of the sample before inspection was proposed by the forgoing application, no proposals have been made with respect to an inspection apparatus which enables a series of operations such as surface flattening, resistive film coating and emission of a charged particle beam for inspection to be efficiently conducted. Conventional independent apparatuses, such as a flattening apparatus for flattening a surface of a sample, a cleaning apparatus for cleaning a sample, and a drying apparatus, are individually placed with a resistive film coating apparatus, and each operation is conducted by using these apparatuses. With this arrangement, however, it is difficult to efficiently conduct the above-mentioned series of operations. The reason for this is as follows. That is, each of the above independent apparatuses comprises a sample loading station and a loading/unloading robot. A sample conveyed by a conveyor apparatus for conveying a sample between the apparatuses is temporarily loaded on the sample loading station. The loading/unloading robot moves the sample from the sample loading station to a work position, i.e., a stage device, and removes the sample from the stage device. A plurality of such independent apparatuses are individually placed and a conveyor apparatus for conveying a sample between the independent apparatuses is further provided. This results in a complicated structure and an increase in overall size of a surface inspection apparatus. Further, the time required for conveying the sample is long and the throughput in the entire apparatus is low. Further, the possibility of contamination and oxidation of a surface of a sample increases, leading to deterioration of product quality. The present invention has been made to solve the problems in the prior arts as mentioned above, and it is an object of the invention to provide a sample surface inspection apparatus and method, which is capable of reliably performing inspections such as observation, detection of defects, analysis, and the like on a surface of a sample without damaging the sample. It is another object of the invention to provide a method for forming a resistive film on a sample and a resistive film forming apparatus enabling an amount of electric charge on a sample to be appropriately controlled and enabling clear image data having minimum distortion to be obtained with respect to a surface of the sample, by making the thickness of the resistive film thin. It is another object of the invention to provide a surface inspection method and a surface inspection apparatus, wherein even when scanning is conducted using a high-current beam so as to increase a throughput, an amount of electric charge on a sample can be appropriately controlled and clear image data having minimum distortion can be obtained with respect to a surface of the sample, thus enabling a defect to be detected by scanning with high reliability. It is another object of the invention to provide a method and an apparatus for inspecting a surface of a sample with high reliability, wherein the thickness of a resistive film formed on the surface of the sample can be precisely controlled so that a resistive film having a desired thickness can be formed uniformly with respect to each sample, to thereby enable clear image data having minimum distortion to be obtained with respect to the surface of the sample. It is a further object of the invention to provide a surface inspection system having a simple structure enabled by improving mechanisms for flattening, cleaning and drying a sample, a mechanism for coating a resistive film on a surface of a sample, and a mechanism for conveying a sample between the above mechanisms, while maintaining a high throughput. It is another object of the invention to provide a surface inspection system which enables a loading/unloading robot of each independent apparatus to be eliminated, thus simplifying the apparatus and a process for conveyance of a sample, and which is capable of maintaining a high throughput. It is another object of the invention to provide a semiconductor device manufacturing method which employs the surface inspection apparatus to inspect a semiconductor wafer in the middle of or after completion of a process. In order to achieve the objects, the present invention provides a method of forming a resistive film on a surface of a sample, which comprises the steps of: rotating the sample at a rotational speed with the sample being held in a substantial horizontal situation; dropping a liquid film material on the sample surface while the sample is being rotated, to form a resistive film thereon; and dropping a solvent which solves the resistive film formed on the sample surface while the sample is being rotated at a rotational speed, thereby dissolving a part of the resistive film to obtain the resistive film having a desired level of thickness. In the above method, it is preferable that the desired level of the thickness of the resistive film is 0.1 nm to 10 nm, and the resistive film is water-soluble. The present invention further provides an apparatus for forming a resistive film on a surface of a sample, which comprises: a spin coater which drops a liquid film material on the sample surface while the sample is rotated at a rotational speed with the sample being held in a substantial horizontal situation, thereby forming a resistive film on the sample surface; and a film thickness uniformalizing mechanism which makes the thickness of the resistive film formed on the sample surface thin and uniform by dissolving a part of the resistive film with a solvent. In the above apparatus, it is preferable that the film thickness uniformalizing mechanism includes a solvent dropping device which drops a solvent dissolving the resistive film while the sample is rotated. The present invention further provides a method of inspecting a surface of a sample, which comprises the steps of; coating the sample surface with a resistive film having an arbitrarily determined thickness; dissolving a part of the resistive film, to thereby reduce the thickness of the resistive film to a desired level which is thinner than that of the arbitrarily determined thickness; and irradiating a charged particle beam to the sample surface coated with the resistive film, to thereby conduct inspection of the sample surface. In the above method, it is preferable that the desired level of the thickness of the resistive film is 0.1 nm to 10 nm, the resistive film is water-soluble, and the method further comprises the step of removing the resistive film from the sample surface by cleaning it with pure water or ultrapure water after the inspection of the sample surface. The present invention further provides a system for inspecting a surface of a sample, which comprises: a surface flattening mechanism for flattening the sample surface; a resistive film coating mechanism for coating a resistive film on the sample surface after the surface is flattened by the surface flattening mechanism, the resistive film having an arbitrarily determined thickness, and then dissolving a part of the resistive film in a solvent, to thereby reduce the thickness of the resistive film to a desired level; an inspection mechanism for emitting a charged particle beam to the sample surface having the resistive film coated thereon, to thereby conduct inspection of the sample surface; and a conveyor mechanism for conveying the sample between the mechanisms. In the above system, it is preferable that the system further comprises a cleaning mechanism and a sample drying mechanism so that the sample in a clean and dry state is introduced into and removed from the surface inspection apparatus, wherein the surface flattening mechanism, the resistive film coating mechanism, the inspection mechanism, the cleaning mechanism and the sample drying mechanism are disposed so as to surround the conveyor mechanism. The present invention further provides a mechanism for inspecting a surface of a sample, comprising: an electromagnetic wave irradiation apparatus comprising an electromagnetic wave source, and a device for guiding an electromagnetic wave generated from the electromagnetic wave source onto a sample surface; a detector for detecting electrons emitted from the sample surface which is irradiated with the electromagnetic wave to output an electric or optical signal; and a processing unit for processing the electric or optical signal from the detector for evaluation of the sample surface. In the above mechanism, it is preferable that the electromagnetic wave source is a source which emits an ultraviolet or X-ray laser, or an ultraviolet ray or X-ray having a wavelength of 400 nm or less. The mechanism could further comprises: an electron beam irradiation apparatus comprising an electron beam source, and a device for guiding an electron beam generated from the electron beam source onto the sample surface; and an apparatus for driving one or both of the electromagnetic wave irradiation apparatus and the electron beam irradiation apparatus to irradiate the sample surface with one or both of the electromagnetic wave or the electron beam. The mechanism could further comprises an imaging optical system for guiding the electrons emitted from the sample surface to the detector. The present invention further provides a method of manufacturing a semiconductor device comprising the step of inspecting a semiconductor wafer in the middle of a manufacturing process and/or after completion of the manufacturing process, using the inspection apparatus or system as above or by the inspection method as above. Referring to the drawings, description is made with regard to embodiments of the present invention. FIG. 2 is an enlarged cross-sectional view showing a sample S to be inspected by a surface inspection method of the present invention. In this embodiment, as the sample S, a silicon wafer (hereinafter, referred to simply as “a wafer”) W during a semiconductor device manufacturing process, on which an arbitrarily determined pattern p such as an electronic circuit has been formed, is taken as an example. In an inspection process or some processes for manufacturing semiconductor circuit, an electron beam is emitted to a surface of the wafer W [in the following explanation of this embodiment, an electron beam (a primary electron beam) is taken as an example of a charged particle beam emitted to a wafer], and the resulting electron beam (a secondary electron beam) emitted from the wafer W is detected, to thereby conduct inspection of the surface of the wafer W. Such inspection is conducted to detect foreign matter, a defect in electric continuity, a defect or loss of a pattern, determine a wafer condition or determine the types of wafers for sorting. However, when the magnitude of a current of an electron beam emitted to the wafer W is increased, the electric charge on the wafer W increases, thus generating distortion of image data produced from detected secondary electrons, and preventing accurate detection. In order to solve this problem, in the above-mentioned patent application, the Applicant proposed a method of controlling the amount of electric charge on the wafer by coating a resistive film f on the wafer. However, it was found that the amount of electric charge on the wafer cannot be precisely controlled simply by coating a resistive film on the wafer. That is, when a resistive film is formed by using a so-called spin coater, as is indicated in FIG. 3, the thickness of resistive film becomes small as the speed of rotation (revolutions per minute) of the wafer increases, but the thickness of resistive film does not decrease when the rotation speed of the wafer exceeds a certain level, namely, about 2,000 rpm. In fact, it is impossible to obtain a resistive film having a thickness less than about 20 nm. As is explained with reference to FIG. 1, with a high beam energy such that η larger than 1, the number of electrons emitted from the insulating material is larger than that of incident electrons. Therefore, the surface of the insulating material is positively charged. The value of resistance and the thickness of a resistive film are herein important. When the value of resistance is as small as that of a metallic film, a potential contrast of the image becomes low, and distortion of the image is suppressed. However, pattern recognition becomes poor and detection of a defect becomes difficult. When the value of resistance is too large, the image is distorted to a large extent. Secondary electrons emitted from a part of the surface cannot be obtained, and arc discharge occurs. In the present invention, as in conventional techniques, the surface of the wafer W (in this embodiment in which the pattern p is formed on an upper surface of the wafer W, the surface of the layer of pattern p) is flattened, by means of a surface flattening mechanism, such as a CMP (Chemical Mechanical Polishing) apparatus having a structure known in the art or a reactive ion flattening apparatus using plasma (FIG. 4(A)). Subsequently, as shown in FIG. 4(B), a resistive film material m in liquid form is sprayed over the flattened surface of the wafer W while the wafer W is rotated, by a conventional spin coater, to thereby form a resistive film f. Coating is conducted so that the thickness of the resistive film f becomes a thickness t1 which is larger than a desired thickness to be finally obtained. As a resistive film material, use is made of, for example, a metal-containing polymer material or a polymer compound of a thienyl alkane sulphonate type. The value of resistance is about 1×106 to 100×106Ω (per cm2, for example; a film thickness: 0.1 nm to 100 nm). A resistive film made of a polymer compound of a thienyl alkane sulphonate type is water-soluble, which is advantageous in a film-removing process described later. Instead of a polymer compound of a thienyl alkane sulphonate, similar coating materials of an acrylic type or a chemical amplification type may be used. According to the present invention, by appropriately selecting the value of resistance, image distortion can be minimized and surface inspection can be conducted efficiently while reducing the risk of false detection. When inspection is conducted during an LSI manufacturing process, removability of the resistive film is important because after the inspection, the resistive film must be removed before subjecting the wafer to the subsequent process. Therefore, use of a water-soluble resistive film is of great importance in the present invention. As is described above, the thickness of a resistive film which can be obtained by a spin coater is limited. In fact, it is impossible to form a resistive film having a thickness less than 20 nm. Therefore, conventionally, a wafer coated with a resistive film having a thickness of 20 nm or more is subjected to inspection. This is problematic, however, in the following point. That is, when acquiring secondary electrons emitted from the wafer by emission of a primary electron beam, an error in acquisition due to the thickness of the resistive film occurs. For example, when electrons pass through the resistive film, due to lowering of a electric field for extraction, secondary electrons are dispersed, leading to a low contrast and a blur of the image. To prevent these problems, it is desired to further reduce the thickness of the resistive film. However, as is described above, conventional coating methods are unsatisfactory. Therefore, in the present invention, as shown in FIG. 4C, a solvent (pure water or ultrapure water n in this embodiment) for the resistive film f is uniformly sprayed over an upper surface of the resistive film f having the thickness t1 formed by the spin coater, to thereby dissolve an upper-side portion of the resistive film f so that the remaining resistive film has a uniform thickness t2. After the film thickness t2 is obtained, by using an inspection mechanism described later and a method known in the art, a primary electron beam, which is a type of a charged particle beam, is emitted to the surface of the wafer W as the sample S, on which the pattern p is formed. A secondary electron beam is emitted from the wafer W and acquired for inspection and evaluation. Experiments were conducted, in which a resistive film was formed on a wafer by a spin coater (the thickness of the resistive film at this time was 20 nm), and then ultrapure water is sprayed to thereby dissolve the resistive film. As a result, the relationship between the thickness of the resistive film remaining on the wafer and the supply amount of ultrapure water, such as that shown in FIG. 5, was obtained. From FIG. 5, it is apparent that a final thickness of the resistive film can be controlled by appropriately controlling the supply amount of a solvent for dissolving the resistive film, such as pure water or ultrapure water. FIG. 6 shows a sample S′ on which steps are formed on a surface of a layer of pattern p (which steps are formed by a plurality of lines l). In this embodiment also, coating of a resistive film and control of a film thickness can be conducted by the same method as explained above in connection with FIG. 3. In this embodiment, the lines l are spaced equidistant from each other. In the case where distances between the lines l are substantially different, surface electrification becomes different, depending on the distance between the lines l. However, a method of the present invention of forming a thin resistive film having a uniform thickness is effective in obtaining a uniform surface potential. The water solubility varies according to a type of a resistive film to be applied, a molecular structure thereof, a type of an addition agent, a concentration of a solution and the like. FIG. 7 is a graph representing a relationship between a thickness of a film and a supply amount of pure water, observed under a condition different to that employed in FIG. 5. In an embodiment of a resistive film forming process, a resistive film solution of a polymer compound of a thienyl alkane sulphonate or isothianaphthenediyl-sulphonate can be employed to form a resistive film. The relationship shown in FIG. 7 represents a film thickness feature in a case that a polymer compound of isothianaphthenediyl-sulphonate and surfactant is utilized as a resistive film solution, the a sample is placed in a spin coater and a resistive film solution is dropped on the sample. Next, the sample is coated with a resistive film, on the spin coater at a rotational speed of, for example, 5000-7000 rpm. Thereby, a resistive film having a thickness of approx. 20 nm is formed. In dissolving the thus formed resistive film, the sample is placed in the spin coater and pure water is dropped on the sample while rotating the spin coater. A rotational speed of the coater is 5000-10,000 rpm. The resistive film is dissolved with pure water or ultrapure water. FIG. 7 shows a thickness of the resistive film remaining on the sample after dissolution. As is clear from the graph, a thickness of a resistive film can be controlled by adjusting the amount of pure water or ultrapure water to be supplied to dissolve the resistive film. In conducting an inspection to detect wafer defects, at first a resistive film is formed on the wafer by a process similarly to the above-described embodiment of a resistive film forming process. For example, a Si wafer on which an LSI device structure has been formed is prepared and a resistive film of a polymer compound of isothianaphthenediyle-sulphonate is deposited to be 1 nm in thickness on the Si wafer. Since a thickness of the resistive film formed on the wafer is controllable by the above-mentioned process, a clear electron image with suppressed charge-up, small distortion and minor contrast deterioration can be obtained. Since a contrast of an electron image varies according to a material and structure of a surface, an even clearer electron image having minimum distortion can be obtained by controlling for a resistive film to have an appropriate thickness. Further, a resistive film forming process can be executed on a photo mask and/or reticle mask. FIGS. 8(A) and 8(B) illustrate cross-sectional views of a photo mask and the same with a resistive film coated thereon. A material of a base structure of the photo mask is silicon glass or quartz glass having a thickness of 1 mm and a material of a pattern is Cr, as shown in FIG. 8(A). A size of the mask is 8×8 inches, a size of a portion of a pattern being 20×20 mm and a size of a minimum pattern being approximately 0.5-0.3 μm. The Cr pattern formed on the substrate has a thickness of 0.1 μm. A resistive film is applied to the surface of the mask by means of spin coating to have a film thickness of 20 nm. Subsequently, the film is dissolved with ultra pure water to have a thickness of 5 nm. Next, an inspection to detect defects such as pattern defects on the photo mask with the resistive film is conducted by using a scanning electron microscope. Instead of using a scanning electron microscope, an mapping projection method may be employed to conduct an inspection to detect a defect by an electron beam. If any defect is not found in the inspection, the resistive film is removed by cleaning with ultrapure water in a cleaning mechanism. After the resistive film is removed, the mask can be used as a proper photo mask again for an exposure system. Although the above example employs a photo mask, the same processing as set forth above may be performed on a reticle mask. In the case of a reticle mask, a pattern reduction rate would be ½-⅕ and a transferred minimum line width would be approximately 0.1-0.5 μm. In a photo mask or reticle mask, a thin film pattern of metal such as Cr and the like is generally deposited on a surface of a glass material. Since a base material is an insulation glass material such as silicon glass, quartz glass and the like, the surface is allowed to charge-up heavily when an electron beam is applied, which makes it difficult to conduct inspection using an electron beam. In the present invention, a photo mask or reticle mask is coated with a resistive film to thereby stabilize the potential on the surface of the mask. Thus, even though an electron beam is applied to the mask, charge-up of the mask is obviated and inspection by means of irradiation of an electron beam becomes possible. This is applicable to a scanning electron microscope, projection electron microscope, an inspection apparatus employing an electron beam and the like. In the resistive film forming method according to the present invention, the thickness of a resistive film on a sample can be precisely controlled so that a resistive film having a desired thickness can be formed uniformly with respect to each sample. Therefore, it is possible to precisely control the amount of electric charge generated by emission of a beam, such as an electron beam or an ion beam, so that a clear potential-contrast image having minimum distortion can be obtained. Conventionally, the thickness of a resistive film cannot be reduced to a level lower than 20 nm. However, a resistive film having a thickness of 0.1 to 1 nm can be obtained in the present invention. Thus, the electric charge on a sample substrate can be appropriately controlled, so that an electron image which is most satisfactory in terms of a potential contrast and distortion can be obtained. When an electron beam or an ion beam is emitted to a sample substrate, in the case where the beam has an energy as high as about 2 keV or more, there is a possibility of the substrate or device circuits formed thereon being damaged. This possibility can be reduced by effecting observation or inspection according to the present invention. With a high beam energy that will cause device damage in conventional techniques, such damage does not occur and an electron image having minimum distortion can be obtained in the present invention. In the above resistive film forming process, the desired level of the thickness of the resistive film may be 0.1 nm to 10 nm. The resistive film may be water-soluble. In this case, after inspection of the sample surface, the resistive film may be removed by cleaning with pure water or ultrapure water. FIG. 9 is a general plan view of an embodiment of a surface inspection system for carrying out a surface inspection method of the present invention. A surface inspection system 100 in this embodiment comprises a central conveyor mechanism 200, and a flattening mechanism 300, a cleaning mechanism 400, a drying mechanism 500, a resistive film coating mechanism 600, an inspection mechanism 700 and a wafer stock interface 800 surrounding the conveyor mechanism 200. As shown in FIG. 10, the conveyor mechanism 200 comprises a movable table 210 movable horizontally (in a lateral direction in FIG. 10) and a plurality of (two in this embodiment) articulated conveyor robots 220a and 220b attached to the movable table 210. The conveyor robots 220a and 220b may have the same structure which is known in the art. Each conveyor robot comprises a base 230, an articulated arm 250 and a chuck 260. The base 230 is vertically movable relative to the movable table 210 and pivotable about an axis 240-240. The articulated arm 250 is connected to the base 230. The chuck 260 is connected to a forward end arm of the articulated arm 250 so as to hold a wafer W. The articulated arm 250 may have a known structure which comprises a plurality of (three in this embodiment) arm members connected so as to be pivotally movable relative to each other. The chuck 260 also may have a known structure. Therefore, detailed description of structures and operations of the articulated arm 250 and the chuck 260 is omitted. The conveyor mechanism 200 conveys a wafer between the above mechanisms 300, 400, 500, 600 and 700, and between each mechanism and the wafer stock interface 800. A wafer held by the chuck 260 of the conveyor robot 220a or 220b is directly loaded on a work station of each mechanism, that is, a stage device, and a wafer is unloaded from the stage device directly by the conveyor robot 220a or 220b. Therefore, differing from conventional independent apparatuses which conduct the same operations as the above-mentioned mechanisms, a loading/unloading robot for each mechanism can be eliminated. Further, in the conveyor mechanism 200 in this embodiment in which a plurality of conveyor robots 220a and 220b are provided, when a wafer is loaded on or unloaded from, for example, the coating mechanism 600, a new wafer held by one conveyor robot 220a is moved to a position close to the coating mechanism 600, and a wafer after coating is removed from the coating mechanism 600 by the other conveyor robot 220b, and the new wafer held by the conveyor robot 220a is placed on a stage device of the coating mechanism 600. Thus, a wafer is loaded on or unloaded from a stage device of each mechanism directly by the conveyor mechanism 200. Therefore, the loading/unloading robots for the respective mechanisms can be eliminated, thus eliminating operations of these robots and simplifying a conveying process. A basic arrangement of the flattening mechanism 300 may be the same as those of a CMP apparatus having a known structure, a reactive ion flattening apparatus using plasma, etc. Therefore, detailed description of a structure and an operation of the flattening mechanism 300 is omitted. The flattening mechanism 300 differs from a conventional independent flattening apparatus such as a conventional CMP apparatus in that a wafer is loaded on a work station, namely, a stage device 310, directly by means of the conveyor robot 220a or 220b of the conveyor mechanism 200, and that a wafer loading station and a loading/unloading robot for conveying a wafer between the conveyor mechanism 200 and the stage device 310 are not provided. A basic arrangement of the cleaning mechanism 400 may be the same as that of a conventional cleaning apparatus adapted to clean a wafer using pure water or ultrapure water, except that a loading/unloading robot and a wafer loading station are eliminated. Therefore, detailed description of a structure and an operation of the cleaning mechanism 400 is omitted. A wafer is loaded on a stage device 410 directly by the conveyor mechanism 200. The drying mechanism 500 may be a conventional drying apparatus of a type which blows dry air or a nitrogen gas against a wafer, a vacuum drying type or a heat-drying type. Therefore, detailed description of a structure and an operation of the drying mechanism 500 is omitted. The drying mechanism 500 differs from a conventional independent drying apparatus in that a wafer is loaded on a stage device 510 directly by the conveyor robot 220a or 220b of the conveyor mechanism 200, and that a wafer loading station and a loading/unloading robot for conveying a wafer between the conveyor mechanism 200 and the stage device 510 is not provided. The resistive film coating mechanism 600 comprises a conventional coating apparatus, such as a spin coater, and a film thickness uniformalizing mechanism 650 for dissolving a part of a resistive film to thereby reduce and uniformalize the thickness of the resistive film. Therefore, in the resistive film coating mechanism 600, a mechanism for coating a resistive film is the same as that of a conventional spin coater and is therefore not described in detail. Only a basic arrangement of the film thickness uniformalizing mechanism 650 is explained, with reference to FIG. 11. Referring to FIG. 11, the film thickness uniformalizing mechanism 650 comprises a micro syringe 660 movable to a position above a rotor 610 of the resistive film coating mechanism 600. The micro syringe 660 is connected through a control valve 680 and a flexible pipe to a supply source 670 of pure water or ultrapure water. In the film thickness uniformalizing mechanism 650, a wafer W is held on the rotor 610 of a spin coater under force of vacuum, and a resistive film material in liquid form is sprayed from a film material dropping head 620 over the wafer W, to thereby conduct coating. At this time, a resistive film having a thickness t1 is formed on the wafer W. Thereafter, the film material dropping head 620 is moved away from the position above the rotor 610, and the micro syringe 660 is moved to that position, in place of the film material dropping head 620. Then, while the wafer W is rotated at a rate of, for example, 3,000 rpm to 7,000 rpm by the rotor 61, pure water or ultrapure water n from the supply source 670 is jet-sprayed or shower-sprayed through a spray opening 6610 of the micro syringe 660 uniformly over an upper surface of the resistive film f coated on the wafer W. Because the resistive film f is water-soluble, an upper-side portion of the resistive film which makes contact with pure water or ultrapure water is dissolved. The thickness t2 of the resistive film to be left on the wafer W is controlled by determining the supply amount of pure water or ultrapure water in consideration of the surface area of the wafer W and the initial film thickness t1. The thickness t1 of the resistive film at the time of completion of coating by a spin coater can be made 20 nm, which is a lower limit of a film thickness obtained by using a conventional spin coater. However, the thickness t1 may be larger than 20 nm. The thickness t2 of the resistive film f remaining on the wafer W after dissolution of a part of the resistive film f in a solvent, such as pure water or ultrapure water, is preferably 0.1 nm to 10 nm. When the film thickness t2 is smaller than 0.1 nm, an effect of suppressing charge-up cannot be obtained, thus preventing high-precision observation. When the thickness t2 is larger than 10 nm, the number of detected electrons (such as secondary electrons) emitted from the resistive film f itself increases, and electrons (such as secondary electrons) emitted from the surface of a sample (such as a wafer) underneath the resistive film are scattered within the resistive film, thus generating a change in an electron orbit and deterioration of an acquisition value (a blur). The thickness t2 of the resistive film f is more preferably 0.5 nm to 2 nm. FIG. 12(A) shows another example of a film thickness uniformalizing mechanism. In this example, while a wafer W is placed on the rotor 610 and rotated (at a rate of, for example, 1,000 to 10,000 rpm), a supply head 660a is disposed at a position above the wafer W (at a height h from the wafer W). The supply head 66a extends along a diameter passing through the axis of rotation of the rotor 610. The supply head 66a includes a number of flow apertures 6610a formed in a lower surface thereof, which are arranged at predetermined spaced intervals in a longitudinal direction of the supply head 660a. Pure water or ultrapure water flows from the flow apertures 6610a, to thereby dissolve a part of the resistive film f in pure water or ultrapure water. In this case, the thickness t2 can be controlled by controlling a total amount of pure water or ultrapure water supplied from the supply head 66a through the flow apertures 661a, by means of the control valve 680. Instead of the supply head 66a having the flow apertures 6610a, a supply head 660b as shown in FIG. 12(B) may be used. The supply head 660b includes a plurality of (five in this example) nozzles 6610b formed on a lower surface thereof, each having a cross-section such as that shown in FIG. 12(C). In this case, pure water or ultrapure water is jet-sprayed from the nozzles 661b. In either case, the distance h between the flow apertures 6610a or the nozzles 6610b and the surface of the resistive film f on the wafer W should be appropriately selected so that pure water or ultrapure water can be sprayed uniformly all over the upper surface of the resistive film f. FIG. 12(D) shows a further example of a film thickness uniformalizing mechanism. In this example, while a wafer W is rotated by the rotor 610, a supply head 660c is moved reciprocally or in one direction between an axis O-O of rotation of the rotor 610 and a position at an outer periphery of the rotor 610, along a diameter passing through the axis O-O of rotation of the rotor 610. Thus, pure water or ultrapure water is sprayed from nozzles 6610c of the supply head 660c, to thereby dissolve the resistive film f. FIG. 13 shows a still further example of a film thickness uniformalizing mechanism. In this example, a supply head 660d is disposed above the rotor 610. The supply head 660d includes a plurality of nozzles 6210d for spraying a film material in liquid form over a wafer and a plurality of nozzles 6610d for spraying a solvent such as pure water or ultrapure water over a resistive film. The nozzles 6210d are connected to a supply source 630d of a film material in liquid form, and the nozzles 6610d are connected to a supply source 670d of a solvent. With this arrangement, it is unnecessary to move the supply head for coating a resistive film or for supplying a solvent. The thickness t2 of the resistive film remaining on the wafer after dissolution of an upper-side portion of the resistive film is measured by an optical measurement method using an ellipsometer, or a method using a step height measurement device, such as a surface roughness measurement device or an atomic force microscope. If desired, measurement of a film thickness can be conducted more precisely by using Auger spectroscopy, SIMS, or characteristic X-ray analysis measurement. Measurement of a film thickness may be conducted with respect to each wafer, although a cumbersome operation is required and a throughput lowers. Normally, suitable conditions are preliminarily determined by conducting experiments so as to obtain the relationship between the film thickness t2 (obtained after dissolution of a part of a resistive film) and various conditions, such as the supply amount of pure water or ultrapure water, the rotation speed of a wafer and the film thickness t1 prior to dissolution. Uniformalization of the film thickness is conducted based on this relationship, and reproducibility in the film thickness due to variation in the conditions is examined by sampling. Therefore, the surface inspection system in this embodiment does not include a device for measurement of a film thickness. After uniformalization of the film thickness, the wafer is introduced into the inspection mechanism (or surface inspection apparatus) 700. As to the inspection mechanism 700, embodiments thereof will be explained later. The wafer stock interface 800 (in FIG. 9) includes a plurality of (at least two) cartridges, each accommodating a plurality of wafers arranged vertically in a spaced relationship. A wafer can be removed from the cartridge by means of the conveyor robot 220a or 220b of the conveyor mechanism 200. This arrangement is known in the art. Therefore, detailed description of the wafer stock interface 800 is omitted. By providing the components of the surface inspection system 100 (i.e., the central conveyor mechanism 200, the flattening mechanism 300, the cleaning mechanism 400, the drying mechanism 500, the resistive film coating mechanism 600, the inspection mechanism 700 and the wafer stock interface 800) in one large chamber, and maintaining the inside of the chamber in a vacuum or an inert gas atmosphere, contamination of wafers during conveyance can be prevented. Using the surface inspection system 100 as illustrated in FIG. 9, a wafer is processed in the following sequence of steps. 1) Pre-Processing: a wafer is pre-processed. 2) Wafer Conveyance: after pre-processing, the wafer is supplied to a wafer station 810 of the wafer stock interface 800. 3) Wafer Loading: the wafer is loaded on the conveyor robot 220a or 220b of the conveyor mechanism 200. 4) Wafer Cleaning: the wafer is conveyed to the cleaning mechanism 400 by the conveyor robot 220a or 220b, and cleaned. 5) Flattening: the wafer is conveyed to the flattening mechanism 300 by the conveyor robot 220a or 220b, and flattened. 6) Cleaning: the wafer is conveyed to the cleaning mechanism 400 by the conveyor robot 220a or 220b, and cleaned. 7) Drying: the wafer is conveyed to the drying mechanism 500 by the conveyor robot 220a or 220b, and dried. 8) Resistive Film Coating and Film Thickness Control: the wafer is conveyed to the resistive film coating mechanism 600 by the conveyor robot 220a or 220b, and coated with a resistive film having a desired thickness t1. 9) Inspection: the wafer is conveyed to the inspection mechanism 700 by the conveyor robot 220a or 220b, and an electron beam or electromagnetic wave beam is emitted to the wafer for inspection. 10) Cleaning: the wafer is conveyed to the cleaning mechanism 400 by the conveyor robot 220a or 220b, and the resistive film on the wafer is completely removed. 11) Drying: the wafer after cleaning is conveyed to the drying mechanism 500 by the conveyor robot 220a or 220b, and dried. 12) Wafer Unloading: the wafer is conveyed to a wafer station 820 of the wafer stock interface 800 by the conveyor robot 220a or 220b. 13) Wafer Conveyance: the wafer is conveyed to a post-process. 14) Post-Processing: the wafer is post-processed. A plurality of wafers can be processed at the same time in the above operating sequence, using the surface inspection system of the present invention. For example, while one wafer is being flattened, another wafer can be coated with a resistive film and inspected. Thus, it is possible to reduce the time required for processing wafers for device manufacturing. Therefore, processing time for each sample can be reduced and a defect in each process can be early detected. The surface inspection system illustrated in FIG. 9 may be additionally provided with a repairing mechanism, a pattern forming mechanism, a developing mechanism, a resin film coating mechanism and the like. In this way, it is possible to flatten, form patterns, evaluate pattern forming masks for defects, and repair defective masks. Embodiments of the inspection mechanism 700 will next be explained with reference to FIGS. 14-20. It should be noted that these embodiments can be employed dependently and independently on the system shown in FIG. 9. FIG. 14 shows a first embodiment of the inspection mechanism 700, which is an electron beam apparatus. The electron beam apparatus 700 shown in FIG. 14 is an imaging projection type electron beam apparatus which comprises an electron gun 71 for emitting a primary electron beam (a shaped beam) 72 shaped by a square opening, a primary electron optical system (hereinafter, referred to simply as “the primary optical system”) 73 for emitting the primary electron beam 72 to a wafer W, a secondary electron optical system (hereinafter, referred to simply as “the secondary optical system”) 75 for acquiring secondary electrons 74 emitted from the wafer W by the irradiation of the primary electron beam 72, and a detector 76 for detecting the secondary electrons 74. In this electron beam apparatus 700, the primary electron beam 72 emitted from the electron gun 71 is reduced by two lens systems 731 and 732 of the primary optical system 73, and an image is formed in a size of 1.25 square mm at a center plane of an ExB separator 733. The electron beam deflected by the ExB separator 733 is reduced to ⅕ by lenses 736 and 737, and projected on the wafer W. The secondary electrons 74 having pattern image data emitted from the wafer W pass through the lenses 736 and 737. The secondary electron beam is then magnified by magnification lenses 751 and 752 of the secondary optical system 75, and forms a secondary electron image at the detector 76. In this electron beam apparatus, the ExB separator 733 is adapted to deflect the primary electron beam 72 emitted from the electron gun 71, while allowing the secondary electrons 74 from the surface of the sample to pass straightforward through the ExB separator 733, and the primary electron beam 720 to become incident at a right angle on the surface of the sample. Of the four magnification lenses 736, 737, 751 and 752, the lenses 736 and 737 form a symmetrical doublet, and the magnification lenses 751 and 752 also form a symmetrical doublet, thus providing a lens system having no distortion. However, slight distortion will be generated due to contamination of electrodes. Therefore, periodically, a reference pattern is placed at the sample plane, and measurement of distortion is conducted to calculate a parameter for correction of distortion. When inspection is conducted by the imaging projection type electron beam apparatus shown in FIG. 14, with respect to a wafer on which an oxide film or a nitride film is selectively formed, the problem of distortion cannot be effectively avoided by only correcting distortion of an optical system. After acquisition of image data, selected points on a pattern edge are compared with a data image, to thereby correct distortion. Thereafter, comparison is made between dies (or chips), or between image data and a data image, to thereby detect a defect. In the inspection mechanism 700 comprising an electron beam apparatus in this embodiment, a conveyor robot for conveying a wafer from the conveyor mechanism 200 to a stage 770 of the electron beam apparatus of the inspection mechanism 700 is not provided. FIG. 15 shows a second embodiment of a sample surface inspection apparatus or mechanism 700. The inspection apparatus in this example comprises a scanning type electron beam apparatus. This scanning type electron beam apparatus 700 comprises an electron gun 71a for emitting a primary electron beam, a primary electron optical system (hereinafter, referred to simply as “the primary optical system”) 73a for illuminating a wafer W with the primary electron beam which has been emitted from the electron gun 71a and shaped, and a detector 76a for detecting electrons, such as secondary electrons, emitted from the wafer W. In this electron beam apparatus, primary electrons emitted from the electron gun 71a are accelerated by anode, and pass through an aperture formed by an aperture plate 731a of the primary optical system 73a, to thereby form an electron beam 72a, which in turn passes through lens systems 732a and 733a and illuminates the wafer W on which the resistive film f is formed. A scanning operation and magnification of the primary electron beam 72a are controlled by a scanning coil 734a and a lens system 735a. Secondary electrons, backscattered electrons or reflected electrons emitted as a result of irradiation of the primary electron beam 72a are detected by the detector 76a, such as a photomal, and form a secondary electron image. The wafer W is attached to a movable stage 770, and continuously moved in an X- or Y-direction at a rate corresponding to an image magnification. Thus, a continuous image can be obtained, by using a liner sensor. Using this secondary electron image, comparison is made between dies (or chips) or between image data and a data image, thus detecting a defect in the wafer W. FIGS. 16-20 respectively show third through seventh embodiments of an inspection mechanism 700. In FIGS. 16-20, the same reference numerals denote the same or similar components. First, the concept of these embodiments of the apparatus will be described below. Each of these embodiments is characterized in that an electromagnetic wave is irradiated onto a surface of a sample, photoelectrons generated thereby is detected, and an image is formed on the basis of the electrons. Particularly, electromagnetic waves having wavelengths of 400 nm or less such as ultraviolet rays, X-rays and the like, are suitable to utilized in the invention. Such electromagnetic waves can limit the amount of charge on the sample surface, so that less distorted images can be obtained. Also, the electromagnetic wave can be uniformly irradiated even when a potential distribution occurs on the sample surface. It is therefore possible to acquire a uniform and clear potential contrast image in the field of view. Also, since the electromagnetic waves are consistent in irradiation characteristic irrespective of the properties of materials irradiated therewith, they can be irradiated to any material. For example, the electromagnetic waves can be similarly irradiated to such samples as semiconductor, LSI, metal, insulating material, glass, living material, polymer material, ceramic, or a composite material thereof, to evaluate the sample surface. The surface inspection apparatus comprises an electron beam generator together with an electromagnetic wave generator, so that they can be selectively driven for irradiation or simultaneously driven for irradiation. For example, for evaluating a surface of a semiconductor waver on which a variety of circuit configurations are formed over multiple layers, an electromagnetic wave can be irradiated to a circuit and a layered structure which must be protected from damages, while an electron beam can be irradiated to a circuit and a layered structure which need not be protected from damages. When damages need not be taken into consideration, a sample can be irradiated with an electron beam having electron irradiation energy selected approximately in a range of 0 to 4 keV to provide best images. For example, when an electron beam having 2-3 keV is irradiated to a wafer having simple L/S metal wires and insulating materials such as SiO2, a less distorted image can be produced than when it is irradiated with an electron beam having lower energy. When damages must be taken into consideration, the breakdown may be increased for gate oxide films and insulating thin films. In such a case, less distorted accurate images can be produced with less damages by selecting an appropriate electromagnetic wave for irradiation. By the irradiation of electromagnetic wave, photoelectrons are emitted from a surface of a sample. Because of low emission energy, the electrons can more readily achieve a focusing condition by an electo-optical system. For example, in an imaging optical system, it is possible to achieve a resolution of 0.1 μm or less, and a maximum image height of 50 μm or more in the field of view. As to an imaging optical system, it is composed such that electron beams are irradiated on a field of view or evaluation range of a sample surface to obtain an image of the field of view, enlarging the image by an electro-optical system, and inspecting the sample on the basis of the enlarged image. Namely, a system in which electron beams are irradiated on a field of view extending at least one dimension on a sample (but not only one point) to obtain an image, is generally called a system of “an imaging type”. As described above, the present invention permits a selection from the following three approaches: (1) Photoelectrons emitted from a sample surface irradiated with an electromagnetic wave such as ultraviolet ray are detected to generate an electron image of the sample surface; (2) Secondary electrons emitted from a sample surface irradiated with an electron beam are detected to generate an electron image of the sample surface; and (4) Photoelectrons and secondary electrons emitted from a sample surface irradiate with both an electromagnetic wave such as ultraviolet ray and an electron beam to generate an electron image of the sample surface. An appropriate approach is selected for use from the foregoing three approaches on the basis of a charging characteristic and anti-damage characteristic associated with a material, structure and the like of a particular sample. In this way, it is possible to efficiently acquire an electron image of a sample surface with high quality while preventing damages on the sample. For example, a semiconductor wafer can be effectively evaluated for detecting defects. For the electromagnetic wave generator, a lamp light source such as mercury lamp, a deuterium lamp, an excimer lamp, or a laser light source may be used as an ultraviolet ray source. Fourth-order harmonic waves of ArF, KrF excimer laser, Nd:YAG laser may be used as a laser. The ultraviolet ray or laser is guided by a light guide or an optical fiber, such that it can be irradiated onto a surface of a sample placed in a vacuum chamber. Optical lenses may be used to control an irradiated region and an optical density. Also, a resist film may be coated on a sample surface for preventing the surface from being charged. While the irradiation of an electromagnetic wave does not cause large charging, the resist film may be coated to generate high quality images which are less distorted when an image is distorted even with slight charging, or when small distortions must be corrected for generating a precise image. A resist film may be coated on a sample which is irradiated with an electron beam, in which case a less distorted image can be generated. With the realization of the foregoing features, it is possible to provide electron emission characteristics more suitable for the properties of particular circuits and layered structures and efficiently generate high quality images while preventing damages and distorted images. In the example mentioned above, ultraviolet rays, ultraviolet laser, X-rays, X-ray laser may be used as the electromagnetic waves. Alternatively, photoelectrons may be emitted using visible light or the like which has a wavelength longer than ultraviolet rays and X-rays. The latter sources can be applied when a sample has a small work function, when multiple photons are absorbed, and the like. It is also possible to employ an imaging optical system for detecting electrons emitted from a sample surface in order to improve a throughput for evaluating the sample surface. For example, a detector using MCP/screen/relay lens/CCD structure can be used for acquiring images in multiple portions in a step & repeat system. Alternatively, a detector using MCP/screen/relay lens/TDI structure can be used for sequentially acquiring images. Since the use of such an imaging optical system allows simultaneous detection of electrons which are two-dimensionally emitted from a sample surface, a two-dimensional electron image can be rapidly acquired. A sample can be evaluated for detecting defects and the like at a high throughput, by utilizing this feature and the aforementioned detector. Since a sample can be inspected or evaluated at a high throughput in the middle of and after completion of a process using the aforementioned surface inspection apparatus, semiconductor devices can be necessarily manufactured at a high throughput. A semiconductor device manufacturing system can be built to have a surface inspection apparatus using an electromagnetic wave, or a surface inspection apparatus using both of an electromagnetic wave and an electron beam, and a surface flattening mechanism. For example, the semiconductor device manufacturing system can be formed to associatively arrange a wafer carrying mechanism, a surface inspection mechanism, a surface flattening mechanism, and a sample drying mechanism in such a manner that a dried semiconductor wafer is introduced into the surface processing apparatus and removed therefrom in a dried state. The system may be additionally provided with a resin film coating mechanism, a pattern forming mechanism, a developing mechanism, and a defect repairing mechanism. FIG. 16 explanatorily shows a third embodiment of the inspection mechanism or apparatus 700. The inspection apparatus in FIG. 16 comprises a stage 770 on which a sample S is carried; an ultraviolet ray source 30 for generating an electromagnetic wave; an imaging optical system 40 having three enlarging lens systems 41-43; a detector 50 for detecting photoelectrons; and an image formation/signal processing unit 60. Each of the lens systems 41-43 comprises two or more lenses for a high resolution and a high zooming ratio. The sample S is, for example, a wafer in which circuit patterns are formed on silicon substrates, in the middle of or after completion of a semiconductor device manufacturing process. While the semiconductor device manufacturing process includes a variety of steps, the ultraviolet ray source 30 irradiates ultraviolet rays to the surface of the wafer to emit photoelectrons from the surface of the sample S, which are detected by the detector 50 through the imaging optical system 40. An electric signal or an optical signal output from the detector 50 is processed in the image formation/signal processing unit 60 for evaluating the presence or absence of debris, defective conduction, defective patterns, drops and the like, state determination, classification and the like in all steps which include those in the semiconductor device manufacturing process. FIG. 17 illustrates a fourth embodiment of the inspection mechanism or apparatus 700. The apparatus of the fourth embodiment correspond to the third embodiment shown in FIG. 16, provided that each of the lens systems 41-43 of the third embodiment is replaced with a single lens. In FIG. 17, an aperture plate 44 is disposed between the second lens 42 and the third lens 43 of the imaging optical system 40. The aperture plate 44 is utilized for noise cutting, transmittance control and aberration amount control. Though not shown in FIG. 16, a similar aperture plate is also provided in the third embodiment. In the fourth embodiment, a sample S is a silicon wafer having a size of 8-12 inches is carried on an X-Y-θ control stage 770. The wafer is formed with circuit patterns of LSI in the middle of manufacturing. A detector 50 used herein can be composed of an electronic amplifier, an opto-electronic converter and a TDI. An ultraviolet ray source 30 used herein is a mercury lamp which irradiates ultraviolet rays into a vacuum chamber (not shown) by an appropriate light guide (not shown) onto the surface of the sample S. An ultraviolet irradiation region is adjusted by optical lenses or the like (not shown) such that the region has a diameter approximately in a range of 0.01 mm to 10 mm. Photoelectrons are generated from the surface of the wafer by the irradiation of ultraviolet rays. The photoelectrons are guided to the detector 50 by the imaging optical system 40. Since the imaging optical system 40 can provide a focus scaling ratio of approximately 50-500 by the three lenses 41-43, the detector 50 two-dimensionally detects a pattern on the wafer, resulted from a difference in photoelectron generation characteristics (such as a working function) on the wafer surface, and the image formation/signal processing unit 60 forms images and processes signals. Through the signal processing, defects are detected and classified. Using the detector 50 and stage 770 in synchronism, wafers can be evaluated in sequence. The TDI detector can also detect electron images in sequence while the stage 770 is moved in sequence. The use of this strategy can save a loss time involved in stop/move and loss time until a moving speed is stabilized, as compared with the step & repeat system, thereby efficiently performing evaluations such as detection of defects. While the third and fourth embodiments have shown examples in which the electromagnetic rays are used as ultraviolet rays, X-rays may be irradiated instead. FIG. 18 illustrates a fifth embodiment of the inspection mechanism or apparatus 700. In the fifth embodiment, ultraviolet rays and electron beam can be selectively irradiated to a sample surface. The inspection apparatus illustrated in FIG. 16 additionally comprises an electron beam irradiation system including an electron source (electron gun) 31 in addition to the ultraviolet irradiation system including the ultraviolet ray source 30 in the inspection apparatus 700 illustrated in FIG. 16. An ExB deflector (filter) 45 is also provided for deflecting a primary electron beam. A sample S is a silicon wafer of 8-12 inches which is carried on an X-Y-Z-θ control stage 770. The wafer is formed with circuit patterns of LSI in the middle of manufacturing. A detector 50 used herein can be composed of an electronic amplifier, an opto-electronic converter and a TDI. The ultraviolet ray source 30 used herein is an Nd:YAG laser based UV laser source for generating fourth-order harmonics (251.5 nm). A UV laser is introduced into a vacuum chamber (not shown) by an appropriate light guide (not shown) for irradiation onto the wafer S within the chamber. An ultraviolet irradiation region is adjusted by optical lenses or the like (not shown) such that the region has a diameter approximately in a range of 0.01 mm to 10 mm. Photoelectrons are generated from the surface of the wafer by the irradiation of ultraviolet rays. The photoelectrons are guided to the detector 50 by the imaging optical system 40. When the electron beam irradiation is used, electrons generated from the electron source 31 are guided to the surface of the wafer S through a plurality of lenses, aligners and apertures (none of them is shown), and the ExB deflector 45 for deflecting the direction of an electron beam. The electron beam source 31 used herein can be an electron gun which comprises, for example, an LaB6 cathode, a Wehnelt and an anode. A primary electro-optical system can be comprised of an aperture, an aligner, an aperture, a quadruple lens for forming an irradiation beam, an aligner, the ExB deflector and the like. The electron beam thus introduced can be controlled by the primary electro-optical system to have a field of view, the diameter of which is approximately in a range of 10 μm to 1 mm. The irradiation field of view can be formed at an aspect ratio not equal to one, for example, ¼ (=x/y). The energy of the electron beam irradiated to a sample can be selected from a range of approximately 0 to 4 kV. In the fifth embodiment, the imaging optical system 40 can implement a focus scaling ratio approximately in a range of 50 to 500 by the three lens systems (a total of six lenses). In this event, electrostatic lenses used herein can satisfy the Wien condition in which electrons emitted from the wafer surface travel straight, while an irradiated electron beam is deflected toward the surface. The detector 50 acquires through the imaging optical system 40 a pattern produced from a difference in the photoelectron generation characteristics (working function and the like) through the irradiation of ultraviolet rays onto the surface of the wafer S, or secondary electrons emitted from the surface of the wafer S irradiated by an electron beam. The electrons from the surface of the wafer S are two-dimensionally detected by the detector 50, and applied to the image formation/signal processing unit 60 for image formation and signal processing. Through the signal processing, the wafer S is evaluated for detecting defects, classifying the defects, and the like. Likewise, in the fifth embodiment, the surface of the wafer S can be inspected or evaluated sequentially by operating the detector 50 and stage 770 in synchronization. The TDI detector can also detect electron images in sequence while the stage 770 is moved in sequence. The use of this strategy can save a loss time involved in stop/move and loss time until a moving speed is stabilized, as compared with the step & repeat system, thereby efficiently performing evaluations such as detection of defects. When the inspection mechanism 700 comprises an ultraviolet (or ultraviolet laser) irradiation device and an electron beam irradiation device as in the fifth embodiment, either of the irradiation devices is selected for use with a particular wafer sample, depending on differences in material and structure. For example, when a semiconductor wafer has a structure near the surface which is susceptible to breakdown, such as a gate oxide film, the ultraviolet rays (laser) should be irradiated to the wafer for evaluation. On the other hand, in an evaluation of a wafer Which is flattened after a wiring material has been embedded, when a gate oxide film or the like is away from the surface to an extent enough to be free from damages, an electron beam may be irradiated to acquire a high contrast electron image for evaluating the wafer. In this event, incident energy to the wafer is preferably in a range of 2 to 3 keV. For materials other than a semiconductor wafer, for example, living samples, polymer materials and the like, ultraviolet ray (laser) irradiation mechanism should be used if an irradiated electron beam would cause charging so strong that the sample could be broken. Both of the ultraviolet rays and electron beam can be irradiated to a single sample. This method can minimize a potential difference between metal and insulating materials both which are on a sample, to reduce distortions in image. Further, irradiating both the ultraviolet rays and electron beam is also preferable when a larger amount of electrons is desirably acquired from the surface of the sample. In this event, the sample may be coated with a resist film. When the sample is irradiated with a highly intense electron beam and ultraviolet rays (laser), the resist film coated on the sample can effectively prevent charging, thereby providing a high quality electron image with less distortions. X-rays or an X-ray laser may be used instead of the ultraviolet rays or ultraviolet laser as well in the fifth embodiment. FIG. 19 illustrates a sixth embodiment of the inspection mechanism or apparatus 700. In the sixth embodiment, both of ultraviolet ray irradiation and electron beam irradiation are employed. An excimer lamp is used for the ultraviolet ray source 30, and ultraviolet rays from the lamp is irradiated to a surface of a sample S through an appropriate optical system (optical lenses and optical fiber). Electrons emitted from the sample S are guided to the detector through an imaging projection type lens system. An electron beam emitted from the electron beam source 31 having an electron gun is reshaped by a square aperture, reduced in size by two lenses 33, 34, and focused at the center of a deflection plane of the ExB deflector 45 in the shape of a square, one side of which is 1.25 mm, for instance. The electron beam deflected by the ExB deflector 45 is reduced to ⅕ by lenses 8, 9 (lens system 41) and irradiated onto a sample S. Electrons (including photoelectrons and secondary electrons) having information of a pattern image, emitted from the sample S, are enlarged by four electrostatic lenses 9, 8, 12, 13 (corresponding to lens systems 41, 42), and detected by the detector 50. The lenses 9, 8 make up a symmetric doublet lens, while the lenses 12, 13 also make up a symmetric doublet lens, thereby providing distortionless lenses. However, since slight distortions may occur due to stains on the electrodes of the lenses, it is preferable that a sample having a standard pattern is used to measure distortions and a parameter is calculated for correcting the distortions. When the sample S is a wafer Which is selectively formed with an oxide film and a nitride film, a correction for distortions in the optical system is not sufficient. Therefore, when the detector 50 acquire image data, a representative point should be selected from a pattern edge for comparison with the image data for correcting distortions. Then, defects should be detected through a die-by-die comparison or an image-by-image comparison. In FIG. 19, a control electrode 15 is applied with an appropriately selected voltage to control the field strength on the surface of the sample S, thereby reducing aberration of electrons (secondary electrons and the like) emitted from the surface of the sample S. FIG. 20 illustrates a seventh embodiment of the inspection mechanism or apparatus 700. In the sixth embodiment, an image detecting system based on laser scanning is employed. The ultraviolet ray source 30 used herein generates a double wave (350-600 nm) or a triple wave (233-400 nm) of a Ti:Al2O3 laser. Generated laser light is guided into a vacuum chamber (not shown) through an optical lens system 35, and is two-dimensionally scanned by a mirror 36 such as a polygon mirror. With the use of the double wave or triple wave of the Ti:Al2O3 laser, a wavelength can be arbitrarily selected in a range of 233 to 600 nm. It is therefore possible to select a wavelength which presents a high photoelectron acquisition efficiency for irradiation in accordance with the material and structure of a sample. Photoelectrons are generated from portions irradiated with the ultraviolet ray laser and detected by a detector 50 which in turn generates an electric signal or an optical signal. In this event, an image can be generated from the electrons emitted from the sample S in the following two modes: 1) Mode Based on Laser Scan and Imaging Optical System: Photoelectrons emitted from the surface of the sample S through scanning of the laser light are focused on the detector 50 through the imaging optical system. In this event, the photoelectrons focused on the detector 50 are enlarged by a factor of 50-500 by the imaging optical system while maintaining the coordinates of positions at which the photoelectrons have been two-dimensionally emitted. The detector 50 can be comprised, by way of example, of MCP/FOP/TDI configuration, MCP/fluorescent plate/relay lenses/TDI configuration, EB-TDI configuration, and the like. Also, a CCD may be substituted for the TDI. When the TDI is used, electron images can be acquired while the sample S is moved in sequence by the stage 770. When the CCD is used, electron images are acquired on a step & repeat basis because the CCD acquires still images. Defects are detected in the images thus acquired in a die-by-die (or chip-by-chip) comparison or a image data based comparison. 2) Mode Based on Laser Scan and Second Detector: A second detector 51 is an electron detector including a scintillator or a photo-multiplier, which comprises an electrode 52 for attracting photoelectrons. This electrode 52 permits scanning of laser light to be synchronized with an electronic signal, so that a signal from the second detector 51 can be processed into an electron image. Defects on the sample is detected on the electron images produced from secondary electrons acquired by the second detector 51 by the image formation/signal processing unit 60 through a die-by-die comparison or a image data based comparison. Next, explanation will be made on a method of manufacturing semiconductor devices which includes procedures for inspecting semiconductor wafers in the middle of a manufacturing process or after the process, using such a surface inspection system as shown in FIG. 9 or such an inspection apparatus as shown in any of FIGS. 14-20. As illustrated in FIG. 21, the method of manufacturing semiconductor devices, when generally divided, comprises the following steps: 1. A wafer manufacturing step S1 for manufacturing wafers (alternatively, a wafer preparing step); 2. A wafer processing step S2 for processing wafers as required; 3. A mask manufacturing step S3 for manufacturing photo masks or reticle masks required for exposure; 4. A chip assembly step S4 for dicing chips formed on a wafer one by one and bringing each chip into an operable state; and 5. A chip testing step S5 for testing finished chips. Each of the steps S1-S5 may include several sub-steps. In the respective steps, a step which exerts a critical influence to the manufacturing of semiconductor devices is the wafer processing step S2. This is because designed circuit patterns are formed on a wafer, and a multiplicity of chips which operate as a memory and MPU are formed in this step. It is therefore important to evaluate a processed state of a wafer executed in sub-steps of the wafer processing steps which influences the manufacturing of semiconductor devices. Such sub-steps will be described below. First, a dielectric thin film serving as an insulating layer is formed, and a metal thin film is formed for forming wires and electrodes. The thin films are formed by CVD, sputtering or the like. Next, the formed dielectric thin film and metal thin film, and a wafer substrate are oxidized, and a mask (photo mask) or a reticle mask created in the mask manufacturing step S503 is used to form a resist pattern in a lithography step. Then, the substrate is processed in accordance with the resist pattern by a dry etching technique or the like, followed by injection of ions and impurities. Subsequently, a resist layer is stripped off, and the wafer is tested. The wafer processing step S3 is repeated the number of times equal to the number of required layers to form a wafer before it is separated into chips in the chip assembly step S4. FIG. 22 is a flow chart illustrating the lithography step which is a sub-step of the wafer processing step S2 in FIG. 21. As illustrated in FIG. 22, the lithography step includes a resist coating step S21, an exposure step S22, a development step S23, and an annealing step S24. After a resist is coated on a wafer formed with circuit patterns using CVD or sputtering in the resist coating step S21, the coated resist is exposed in the exposure step S22. Then, in the development step S23, the exposed resist is developed to create a resist pattern. In the annealing step S24, the developed resist pattern is annealed for stabilization. These steps S21 through S24 are repeated the number of times equal to the number of required layers. In the process of manufacturing semiconductor devices, a test is conducted for defects and the like after the processing step which requires the test. However, the electron beam based defect testing apparatus is generally expensive and is low in throughput as compared with other processing apparatuses, so that the defect testing apparatus is preferably used after a critical step which is considered to most require the test (for example, etching, deposition (including copper plating), CMP (chemical mechanical polishing), planarization, and the like). In the above-mentioned semiconductor device manufacturing method, a defect in flattening of a sample can be immediately detected and determined, thus enabling efficient process control. The surface inspection apparatus is capable of detecting not only a defect in flattening, but also a failure or a defect in a process prior to flattening. Therefore, process control can be conducted with respect to flattening and other processes conducted prior to flattening. Based on information obtained by this process control, it is possible to efficiently detect, rectify and improve a defect in each process which has been carried out. Further, an entire structure of an inspection apparatus capable of conducting a series of operations between flattening and inspection can be simplified, and samples can be smoothly conveyed using the system illustrated in FIG. 9, thus ensuring a high throughput. In summary, the present invention is advantageous in the following points. (1) According to the present invention, a defect in a wafer surface can be correctly measured by suppressing distortion of a measurement image due to electrification. Therefore, it is possible to conduct a scanning operation using a high current, while the electric charge on the wafer can be appropriately controlled, and a large number of secondary electrons can be detected. Therefore, a detection signal having a desirably high S/N ratio can be obtained, thus enabling a defect to be detected with high reliability. (2) A high S/N ratio can be obtained, and image data can be produced even when scanning is conducted at a high speed. Therefore, a throughput in an inspection apparatus can be increased. (3) When an electron beam or an ion beam is emitted to a wafer, in the case where the beam has energy as high as about 2 keV or more, there is a possibility of a substrate or device circuits formed thereon being damaged. This possibility can be reduced by effecting observation or inspection according to the present invention. While high beam energy causes device damage in conventional techniques, such damage does not occur in the present invention and an electron image having minimum distortion can be obtained. (4) A defect in flattening of a wafer can be immediately detected and determined, thus enabling efficient process control. The surface inspection apparatus of the present invention is capable of detecting not only a defect in flattening, but also a failure or a defect in a process prior to flattening. Therefore, process control can be conducted with respect to flattening and other processes conducted prior to flattening. Based on information obtained by this process control, it is possible to efficiently detect, rectify and improve a defect in each process which has been carried out. Further, a yield of a device manufacturing process can be increased, and a cost of manufacture of devices can be reduced. (5) In order to provide the above advantage, it is possible to use an image processing system capable of determining the types of defects generated in wafer surfaces. This image processing system is capable of determining, for example, deposition of a foreign matter, contact failure in circuits, and a defect in a pattern form. Therefore, it is possible to determine, from an acquired electron image, in which manufacturing process or which place in an apparatus a defect was generated. Thus, a cause of a defect can be efficiently improved and rectified. (6) According to the present invention, it has become possible to manufacture devices by a method in which a defect in a wafer during a process is detected by using the above-mentioned inspection apparatus and method. (7) Loading/unloading robots, such as those used in a conventional independent apparatuses, can be eliminated, thus simplifying an entire structure of the apparatus and reducing a size of the apparatus. Further, a process for conveyance of a wafer can be simplified, thus increasing a throughput. (8) The respective mechanisms of the inspection apparatus are disposed on a single base. Therefore, it is easy to conduct adjustment of a work position of a wafer and a loading/unloading position, which is difficult in a conventional system in which a plurality of independent apparatuses are placed. (9) The number of utilities and control valves for supplying compressed air, electricity, cooling water, pure water or ultrapure water, a nitrogen gas, etc, can be reduced. Therefore, the inspection apparatus can be simplified. (10) The possibility of exposing a wafer to an external environment can be reduced, thus preventing contamination of wafers. Although the present invention has been described above in detail with reference to the drawings, the foregoing description is for explanatory purposes and not intended to limit characteristics. It should be understood that the foregoing description merely illustrates and explains preferred embodiments, and all modifications and changes within the scope of the spirit of the present invention are protected. |
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047160128 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to the internals of a nuclear reactor and more specifically to a strainer which prevents debris from lodging between the secondary lower core support and the hemispherical lower head on the pressure vessel. 2. Prior Art A pressurized water reactor includes an upright cylindrical pressure vessel with a hemispherical lower head section and a removable hemispherical head bolted to the upper end. A cylindrical core barrel is suspended inside the pressure vessel from a flange extending around the inside of the upper end of the pressure vessel. The core of fissile material is supported inside this core barrel. An upper core support, commonly referred to as the top hat, is clamped down on top of the flange from which the core barrel is suspended by the removable hemispherical upper head. In order to accommodate for tolerances, an annular spring is placed between the flanges on the core barrel and the upper core support. Reactor coolant is introduced into the pressure vessel near the top of the core barrel, flows downward through the annular space between the core barrel and the inner wall of the pressure vessel known as the downcomer, reverses direction inside the hemispherical lower head section, and flows upward through passages in the bottom of the core barrel and through the reactor core mounted inside the core barrel before being discharged through outlet nozzles. Heat energy generated by the fission reactions in the core is absorbed by the reactor coolant and is utilized to generate steam for use by a turbine-generator in producing electricity. As a safety precaution, energy absorbers are mounted under the suspended core barrel so that in the very unlikely event that there should be a complete failure of the core barrel suspension system, the impact of the entire core barrel assembly falling on the lower hemispherical head section is lessened to preserve pressure vessel integrity. In such an event, actual contact with the lower hemispherical head would be made by a horizontal secondary core support plate underneath the energy absorbers. In order to keep the kenetic forces in such an accident to a minimum, the gap between the secondary core support plate and the lower hemispherical head section of the pressure vessel is very small. However, due to the differences in the coefficients of thermal expansion of the stainless steel internals and the lower alloy pressure vessel, this gap varies in size over the temperture range to which these components are exposed. Typically, this gap can narrow from 1.06 inches cold to 0.5 inches when the reactor is operating at full power. Periodically, the reactor is shutdown for refueling. During this sequence, the hemispherical upper head is removed along with the upper core support and the components it supports so that fuel assemblies can be replaced and rearranged. The possibility exists during the refueling procedure for debris to fall down into the lower portion of the vessel. Such debris can include small parts, such as nuts and bolts. Debris can also be introduced into the pressure vessel by circulation of reactor coolant following failure of the hardware or maintenance on other parts of the nuclear steam supply system. For instance, during retubing of the steam generators, small pieces of tubing and pieces of weld material can be left behind despite attempts to clean them out. If such debris should lodge between the secondary lower core support and the lower hemispherical head section of the pressure vessel when the reactor coolant is cold, it could cause the core barrel to unseat from its support flange as the components heat up, due to the difference in the coefficients of thermal expansion of the internals and the pressure vessel. This in turn, could subject the internals to undesired vibration induced by turbulent reactor coolant flow. As taught by U.S. Pat. No. 4,096,032, it is known to insert filters in the bottom of the core barrel of a pressurized water reactor temporarily during cold hydrostatic and hot functional testing to collect debris from construction of the nuclear steam supply system. However, this is done before the fuel assemblies are installed and it collects the debris in the lower hemispherical head section of the pressure vessel from which it must be removed prior to operation of the reactor or it will cause the very problem which the present invention seeks to avoid. It is the primary object of the present invention to provide apparatus which prevents debris from lodging in the gap between the internals and the lower hemispherical head section of the reactor pressure vessel without restricting the flow of reactor coolant, and which does so as the gap varies in size over the full range of temperatures to which the reactor is subjected. SUMMARY OF THE INVENTION This and other objects are realized by strainer means which comprises an annular member with apertures therethrough secured to the lower end of the intrenals of a nuclear reactor and extending radially outward to the hemispherical lower head section of the pressure vessel above the gap formed by the bottom of the internals and the inner surface of the lower hemispherical head section. This annular member is a resilient planar device, preferably upwardly convex, which is fixed to the internals and bends to maintain its outer peripheral edge in contact with the lower hemispherical head section as the size of the gap changes with temperature. The means for mounting the annular planar member to the internals can include an annular rim around its inner edge which is welded to the secondary core support base plate. With this arrangement, debris is prevented from entering the gap between the internals and the lower hemispherical head section of the reactor pressure vessel over the full range in the size of the gap, yet there is always a flow of reactor coolant through the gap to provide cooling for the reactor components. |
062663860 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a method and apparatus to aid in the maintenance and repair of the lower internal assembly of a nuclear reactor vessel, and more specifically, to a method and apparatus which provides a frame capable of inverting the lower internal assembly so that repair and maintenance operations may be performed with greater speed, thus reducing the exposure of workers to radiation. 2. Background Information Because of the radiation hazard present while performing repair and maintenance operations on the components of a nuclear reactor, it is desirable to limit the exposure of workers to any radioactive components during such maintenance and repair procedures. A nuclear reactor includes a stationary reactor vessel which encloses a removable reactor core assembly. The reactor core assembly includes two main assemblies, the upper internal assembly and the lower internal assembly. For example, the upper internal assembly includes the control rod drive mechanisms, control rod drive shafts and the upper core plate. The lower internal assembly includes the core barrel, lower instrumentation guide tubes, tie plates and a baffle. The nuclear fuel assemblies or fuel cells are maintained within the core barrel between the upper core plate and the lower core plate. It is known in the prior art to perform maintenance operation on the lower internal assembly. Access to the lower internal assembly is gained by removing the reactor vessel upper head assembly and the upper internal assembly, including the upper core plate. This procedure exposes the fuel assemblies which are also removed. Once the upper internal assembly, which includes the upper core plate, and fuel assemblies have been removed, the lower internal assembly may be removed from the reactor vessel. When removed from the reactor vessel, the lower internal assembly is set on a storage stand which supports the lower internal assembly above the maintenance bay floor. While on the maintenance stand the lower end of the core barrel is approximately thirty feet under water. Typically, the only procedure performed on the lower internal assembly is the inspection of the weld between the lower core forging and the core barrel. Should maintenance be required on elements of the lower internal assembly below the lower core forging, it would be necessary to construct a tool capable of being submerged thirty feet and turned at a 90-degree angle in order to access the lower internal assembly. Maintenance on the lower face of the lower core forging would require the instrument to have an additional 90-degree turn to access the lower side of the lower core forging. Maintenance procedures using such a tool would be time consuming and would expose workers to radiation throughout the period in which the tool was used. Such a high level exposure to radiation is not desirable. Therefore, there is a need for a method and apparatus to allow workers to repair the lower internal assembly of a nuclear reactor core assembly which would reduce the amount of exposure to radiation. SUMMARY OF THE INVENTION These needs and others are satisfied by the invention which is directed to a method and apparatus for up-ending the lower internal assembly of a nuclear reactor so that the lower internal assembly and lower core forging may be directly accessed by maintenance and repair workers. The apparatus for up-ending the core barrel and lower internal assembly includes a support which is capable of supporting the core barrel and lower internal assembly in the upright orientation, the horizontal orientation, and the inverted orientation. The apparatus includes a frame assembly having support brackets and support saddles for the core barrel. The apparatus further includes a spider assembly which is fitted within the core barrel and is used to support the internal baffle while the barrel is in the horizontal orientation. The apparatus allows workers to invert the lower internal assembly so that the lower core forging is positioned above the core barrel. This invention further provides the method for inverting the lower internal assembly. The lower internal assembly is removed from the reactor vessel and placed on the storage stands as is known in the prior art. The spider assembly, which supports the baffle located within the core barrel, is then inserted in the lower internal assembly. The lower internal assembly may be lifted and inserted into the frame. Once the lower internal assembly is positioned within the frame, the frame is rotated ninety degrees before being inverted. Once the lower internal assembly is in the inverted orientation, the lower internal assembly is removed from the frame and positioned on the storage stand in the inverted position. Any maintenance on the lower face of the core forging or structures typically located below the lower core forging can now be performed with tools that directly access the lower internal assembly and lower core forging without the necessity of having such tools bent 90 degrees or more. Accordingly, maintenance and repair procedures can be performed more quickly with a reduced radiation exposure to the workers. |
claims | 1. Method of measurement by atomic interferometry, in which each session of measurements is executed with at least two sets of atoms (11, 12) each subjected to conditions of formation of atomic interference,the atoms of each set of atoms (11, 12) being of a species dedicated to said set of atoms and different from the species of atoms of each other set of atoms,method in which, for each session of measurements, said conditions are produced for each set of atoms (11, 12) throughout a volume that is associated with said set of atoms and that contains at least one point in common with the volume associated with each other set of atoms, and are produced between a start time point and an end time point respectively before and after an intermediate time point common to all the sets of atoms,and in which a measurement result (P11, P12) is obtained in each session of measurements independently for each set of atoms (11, 12), each measurement result varying according to a first function of a total phase shift that appeared for the corresponding set of atoms during formation of the atomic interference, said total phase shift comprising a sum of a second function of an external parameter (a) a value of which is sought and of a constant phase shift that is undergone by the corresponding set of atoms during said formation of the atomic interference,the method being comprising the following steps:/1/ during a session of measurements, applying a value for at least one operating parameter, called internal parameter and making it possible to control a difference between the constant phase shifts to which the two sets of atoms are respectively subjected, the value applied for said at least one internal parameter being such that a difference between the total phase shifts that the two sets of atoms (11, 12) undergo, respectively, is between Π/4 and 3Π/4 in absolute value and modulo Π;/2/ for each measurement result (P11, P12) obtained for one of the sets of atoms (11, 12) in said session of measurements, determining a derivative value of said measurement result with respect to the external parameter (a), said derivative being evaluated for said measurement result;/3/ selecting that one of the sets of atoms (11, 12) for which the derivative value determined in step /2/ is largest in absolute value; and/4/ calculating the value of the external parameter (a) from the measurement result (P11, P12) obtained in step /2/ for the set of atoms selected in step /3/. 2. Method according to claim 1, in which said at least one internal parameter comprises at least one amplitude of a phase jump introduced between two pulses of laser radiation that are used to form the atomic interference for one of the sets of atoms (11, 12), at least one rate of variation of a frequency of laser radiation that is used to form the atomic interference for one of the sets of atoms, or an intensity and a gradient of a magnetic field that is applied to the sets of atoms during the formation of the atomic interferences, or a combination of several among said at least one amplitude of phase jump, said at least one rate of variation of frequency of laser radiation and said magnetic field intensity and gradient. 3. Method according to claim 1, in which the value applied for said at least one internal parameter is such that the difference between the total phase shifts that the two sets of atoms (11, 12) undergo, respectively, is between 15Π/32 and 17Π/32, in absolute value and modulo Π,and in which, for that one of the sets of atoms (11, 12) that is selected in step /3/, the first function is replaced with an affine function of the total phase shift that appeared during the formation of the atomic interference for the selected set of atoms, in a whole interval of values having a length of interval greater than or equal to 3Π/8, and which contains said total phase shift that appeared during the formation of the atomic interference. 4. Method according to claim 1, in which the first function has the expression P=P0·[1−C×cos(ΔΦtot)] for each set of atoms (11, 12), where P denotes the measurement result, ΔΦtot is the total phase shift that appeared during the formation of the atomic interference for said set of atoms, and P0 and C are two non-zero numbers. 5. Method according to claim 1, in which said at least one internal parameter comprises an amplitude of a phase jump introduced between two pulses of laser radiation that are used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises a term proportional to the amplitude of the phase jump. 6. Method according to claim 1, in which said at least one internal parameter comprises a rate of variation of frequency of laser radiation that is used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises the term −2Π×α×T2 where T is a base time for a sequence of interactions between the atoms and photons that is implemented to form the atomic interference of said set of atoms, and α is the rate of variation of frequency of the laser radiation. 7. Method according to claim 1, in which said at least one internal parameter comprises an intensity and a gradient of a magnetic field that is applied to the sets of atoms (11, 12) during the formation of the atomic interferences, and the constant phase shift that is undergone by each set of atoms comprises the term (Aat/Mat)×B0×B1×ℏ×k ×T2, where B0 and B1 are the intensity and the gradient of the magnetic field respectively, T is a base time for a sequence of interactions between the atoms and photons that is implemented to form the atomic interference for said set of atoms, k is a modulus of a momentum received or transferred by one of the atoms during each interaction between the atoms and the photons, divided by ℏ=h/(2Π) where h is Planck's constant, and Aat/Mat is a coefficient that depends on the species of atoms. 8. Apparatus for measurement by atomic interferometry comprising:a source of atoms (100) suitable for producing at least two sets of atoms (11, 12), with the atoms of each set of atoms that are of a species dedicated to said set of atoms and different from the species of atoms of each other set of atoms;means (101-103) suitable for producing conditions of atomic interference for each set of atoms (11, 12), so that said conditions are produced for each set of atoms throughout a volume that is associated with said set of atoms and that contains at least one point in common with the volume associated with each other set of atoms, and produced between a start time point and an end time point respectively before and after an intermediate time point that is common to all the sets of atoms, so as to constitute a session of measurements;a detection device arranged for providing measurement results (P11, P12) respectively and independently for all the sets of atoms (11, 12) of each session of measurements; andan analysis unit suitable for calculating at least one value of an external parameter (a) from each measurement result (P11, P12),in which each measurement result (P11, P12) varies according to a first function of a total phase shift that appeared for the corresponding set of atoms during the formation of the atomic interference, said total phase shift comprising a sum of a second function of the external parameter (a) the value of which is sought and of a constant phase shift that is undergone by the corresponding set of atoms during said formation of the atomic interference,the apparatus being applying, during each session of measurements, a value for at least one operating parameter, called an internal parameter and making it possible to control a difference between the constant phase shifts that the two sets of atoms (11, 12) undergo, respectively, so that a difference between the total phase shifts to which the two sets of atoms are subjected respectively, is between Π/4 and 3Π/4, in absolute value and modulo Π;and the analysis unit is suitable for executing steps /2/ to /4/ of a method of measurement by atomic interferometry according to claim 1. 9. Apparatus according to claim 8, in which, for each session of measurements, the conditions of atomic interferences are produced for all the sets of atoms (11, 12) using a single laser source assembly (102, 103), common to said sets of atoms. 10. Apparatus according to claim 8, forming an accelerometer, a gravimeter or a gyrometer. 11. Method according to claim 2, in which the value applied for said at least one internal parameter is such that the difference between the total phase shifts that the two sets of atoms (11, 12) undergo, respectively, is between 15Π/32 and 17Π/32, in absolute value and modulo Π,and in which, for that one of the sets of atoms (11, 12) that is selected in step /3/, the first function is replaced with an affine function of the total phase shift that appeared during the formation of the atomic interference for the selected set of atoms, in a whole interval of values having a length of interval greater than or equal to 3Π/8, and which contains said total phase shift that appeared during the formation of the atomic interference. 12. Method according to claim 2, in which the first function has the expression P=P0·[1−C×cos(ΔΦtot)] for each set of atoms (11, 12), where P denotes the measurement result, ΔΦtot is the total phase shift that appeared during the formation of the atomic interference for said set of atoms, and P0 and C are two non-zero numbers. 13. Method according to claim 3, in which the first function has the expression P=P0·[1−C×cos(ΔΦtot)] for each set of atoms (11, 12), where P denotes the measurement result, ΔΦtot is the total phase shift that appeared during the formation of the atomic interference for said set of atoms, and P0 and C are two non-zero numbers. 14. Method according to claim 2, in which said at least one internal parameter comprises an amplitude of a phase jump introduced between two pulses of laser radiation that are used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises a term proportional to the amplitude of the phase jump. 15. Method according to claim 3, in which said at least one internal parameter comprises an amplitude of a phase jump introduced between two pulses of laser radiation that are used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises a term proportional to the amplitude of the phase jump. 16. Method according to claim 4, in which said at least one internal parameter comprises an amplitude of a phase jump introduced between two pulses of laser radiation that are used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises a term proportional to the amplitude of the phase jump. 17. Method according to claim 2, in which said at least one internal parameter comprises a rate of variation of frequency of laser radiation that is used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises the term −2Π×α×T2, where T is a base time for a sequence of interactions between the atoms and photons that is implemented to form the atomic interference of said set of atoms, and α is the rate of variation of frequency of the laser radiation. 18. Method according to claim 3, in which said at least one internal parameter comprises a rate of variation of frequency of laser radiation that is used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises the term −2Π×α×T2, where T is a base time for a sequence of interactions between the atoms and photons that is implemented to form the atomic interference of said set of atoms, and α is the rate of variation of frequency of the laser radiation. 19. Method according to claim 4, in which said at least one internal parameter comprises a rate of variation of frequency of laser radiation that is used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises the term −2Π×α×T2, where T is a base time for a sequence of interactions between the atoms and photons that is implemented to form the atomic interference of said set of atoms, and α is the rate of variation of frequency of the laser radiation. 20. Method according to claim 5, in which said at least one internal parameter comprises a rate of variation of frequency of laser radiation that is used to form the atomic interference for one of the sets of atoms (11, 12), and the constant phase shift that is undergone by said set of atoms comprises the term −2Π×α×T2, where T is a base time for a sequence of interactions between the atoms and photons that is implemented to form the atomic interference of said set of atoms, and α is the rate of variation of frequency of the laser radiation. |
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description | The present invention relates to the field of tokamak fusion reactors. In particular the invention relates to a combined neutron shield and solenoid for use in the central column of a tokamak, particularly a spherical tokamak. A tokamak features a combination of strong toroidal magnetic field, high plasma current and, usually, a large plasma volume and significant auxiliary heating, to provide hot, stable plasma. This allows tokamaks to generate conditions so that fusion can occur. The auxiliary heating (for example via tens of megawatts of neutral beam injection of high energy H, D or T) is necessary to increase the temperature to the sufficiently high values required for nuclear fusion to occur, and/or to maintain the plasma current. The problem is that, because of the large size, large magnetic fields, and high plasma currents generally required, build costs and running costs are high and the engineering has to be robust to cope with the large stored energies present, both in the magnet systems and in the plasma, which has a risk of ‘disrupting’—mega-ampere currents reducing to zero in a few thousandths of a second in a violent instability. The situation can be improved by contracting the donut-shaped torus of a conventional tokamak to its limit, having the appearance of a cored apple—the ‘spherical’ tokamak (ST). The first realisation of this concept in the START tokamak at Culham demonstrated a huge increase in efficiency—the magnetic field required to contain a hot plasma can be reduced by a factor of 10. In addition, plasma stability is improved, and building costs reduced. To obtain the fusion reactions required for economic power generation (i.e. much more power out than power in), a conventional tokamak would have to be huge so that the energy confinement time (which is roughly proportional to plasma volume) can be large enough so that the plasma can be hot enough for thermal fusion to occur. WO 2013/030554 describes an alternative approach, involving the use of a compact spherical tokamak for use as a neutron source or energy source. The low aspect ratio plasma shape in a spherical tokamak improves the particle confinement time and allows net power generation in a much smaller machine. However, a small diameter central column is a necessity, which presents challenges for design of the plasma confinement vessel and associated magnets. During the initial phase of starting up a tokamak, the neutral gas which fills the confinement vessel must be ionised to produce a plasma. The process, known as “breakdown”, “formation” or “initiation”, is achieved by passing a time varying current through toroidally wound poloidal field (PF) coils of the tokamak. This time varying current generates a “loop voltage” inside the vessel that, when sufficiently large, causes the gas to break down and form a plasma. The loop voltage produced is a function of the position of the toroidal field coils and the time variation of the current. As well as generating a loop voltage inside the vessel, a current will also be induced in any other toroidally wound conducting loops (e.g. the plasma or the confinement vessel wall). The most common plasma formation technique uses a solenoid wound in the central column of the tokamak to carry the time varying current and generate the loop voltage. This method is well known, reliable, and used in the majority of tokamaks. However, the compact geometry of spherical tokamaks makes implementation of this method problematic—there is limited space in the centre of the torus, and this space is needed for the toroidal field coils, cooling, and neutron shielding. As the size and efficiency of a spherical tokamak is related to the size of the central region, the space taken up by a solenoid has a direct impact on this efficiency. Current spherical tokamaks such as MAST and NSTX use a solenoid—but the increased neutron load expected in next generation fusion reactors would make the designs used for those tokamaks impractical due to the extra shielding required. According to a first aspect, there is provided neutron shielding for the central column of a tokamak nuclear fusion reactor. The neutron shielding comprises an electrically conductive neutron absorbing material. The neutron shielding is arranged such that the electrically conductive neutron absorbing material forms a solenoid for the initiation of plasma within the tokamak. Further embodiments are described in claim 2 et seq. The use of a conductive material for neutron shielding enables the construction of the neutron shielding and solenoid in a single unit. In other words, the neutron shielding may be constructed in the form of a solenoid, and a power supply provided so that the plasma initiation current may be driven through this solenoid. A suitable material must be both electrically conductive (e.g. with a conductivity of greater than 1 MS/m at 300K) and neutron absorbing. The neutron shield may be constructed so that there is a helical current path along the central column, forming the solenoid. If there are multiple layers of shielding, alternate layers may have helices with a different sense of rotation, such that the layers may be connected alternately at the top and bottom to form a single solenoid, similar to layers in a conventional wound wire solenoid. While the below description is applicable to shielding for the central column of any tokamak, it is particularly beneficial for a spherical tokamak due to the constraints imposed on the width of the central column by the spherical tokamak design. Electrically insulating material may be provided within the shielding to define the current path. This insulating material may be neutron shielding itself, or it may have limited or no neutron shielding effect. In the latter case, multiple layers of neutron shielding should be used, arranged such that there is no “line of sight” through the insulating material from the plasma chamber to the central column (i.e. no straight path leading from the plasma chamber to the central column along which a neutron could travel). Alternatively, the neutron shielding may be constructed such that there is no line of sight even where only a single layer is used. Otherwise, areas of the central column which have line of sight to the plasma chamber will experience much higher and potentially damaging radiation doses. The neutron shield may be constructed from several segments of electrically conductive neutron absorbing material, which are connected together to form a helical current path. As an example, this may be done by connecting several segments, each of which is a segment of a helix, or by stacking several horizontal annular arc segments, with each being connected vertically to the next to form an approximately helical path (shown in more detail in the example of FIG. 3, described below). The segments may be connected by interlocking cooperating features to provide both electrical connection and structural rigidity. Alternatively, there may be separate features providing each of structural connection and electrical connection (e.g. a non-conducting set of interlocking cooperating features which hold the segments in place such that there is contact between conducting faces). The features providing structural connection may have shear strength greater than that of the electrically conductive neutron absorbing material. The segments may have insulating layers to prevent unwanted electrical contact between the segments, or the insulating layers may be provided separately during construction of the neutron shielding. The segments may comprise an material with a lower resistivity than the electrically conductive neutron absorbing material placed to facilitate the electrical connection. An exemplary construction will now be described. However, it will be apparent to the skilled person that many other constructions are possible, particularly in light of the ability to cast cemented carbides into a variety of shapes. FIGS. 1 and 2 show a shield segment 1 according to the exemplary construction. The shield segment has top 2 and bottom 3 surfaces which form annular arcs, and the shield segment has sides which extend generally vertically between the top and bottom surfaces. An insulating layer 4 is provided on the top surface of the segment, except where it connects to the next segment in the helix, and on one of the end surfaces of the segment. The connection to the next segment is made using complementary interlocking features located at one end of the top surface and at the opposite end of the bottom surface, such as a protrusion 5 and corresponding recess 6. The protrusion 5 is made from a material with a higher shear strength than the electrically conductive neutron absorbing material, to provide additional rigidity to the segments. Alternatively, a recess may be provided in each segment on both the top and bottom surface, and a dowel or similar connection inserted into both recesses to connect the segments. Electrical connection is achieved by the use of an electrically conductive region 7 of the top surface, which extends to the same level as the insulating material (otherwise, there would be a gap between the electrically conductive neutron absorbing material the thickness of the insulating layer). This may be an extension of the electrically conductive neutron absorbing material, or it may be a patch of a different electrically conductive material, e.g. one with a higher conductivity thean the rest of the neutron shield, e.g. copper. FIG. 3 shows how the segments may be arranged to form a solenoid coil, with the protrusion or dowel 5 of each segment being locked into the bore of the next segment, and arranged to form an approximately helical shape. As shown in FIG. 4, where the arc angle is a little less than 180°, a complete shielding layer is provided by having two “series” 1a and 1b of segments, each defining a separate helix. FIG. 4 also demonstrates how the segments are arranged around the cryostat 8 of the central column. In order to prevent there being line-of-sight through the insulating layer, a second layer 12 of shielding segments may be overlaid on the first layer 11 as shown in FIG. 5. The segments of the second layer have an inner radius corresponding to (or slightly greater than) the outer radius of the segments of the first layer, and the complementary interlocking features of the segments of the second layer are provided on the opposite ends of the upper/lower surfaces compared to those of the first layer. This allows the second layer to wrap around the first layer, and ensures that the second layer is wound in the opposite sense of rotation. As such, the first and second layer may be connected at the top or bottom to form a single solenoid. Alternatively, the second layer may be wound in the same sense and connected in parallel. The second layer is offset in the axial and rotational directions from the first layer to avoid any line of sight through the insulating layers. There are several possibilities for electrically conductive, neutron absorbing materials. Previous work has shown the suitability of cemented carbides, borides, or borocarbides, e.g. tungsten carbide, as a neutron shielding material (see WO 2016/009176 A1). These materials are electrically conductive (due to the metal binder and often the carbide/boride aggregate being conductive). Cemented carbides are a metal matrix composite in which particles of a carbide act as the aggregate, and a metallic binder serves as the matrix. Cemented carbides are formed by a sintering process, in which the material is heated to a point where the binder is liquid, but the carbide particles remain solid. The carbide grains are thereby embedded into the liquid binder, which is then allowed to set. This results in a material with superior qualities to either the carbide or the binder taken alone. The ductile binder offsets the natural brittleness of the carbide ceramic, and the carbide particles make the resulting composite much harder than the binder alone. Due to the metal binder, cemented carbides typically have a high thermal conductivity, which reduces the thermal stress experienced by the material due to uneven heating. The coefficient of linear thermal expansion of cemented carbides or borides is typically in the range of 4 to 5×10−6. Cemented materials are also resistant to sputtering (ablation of the outer surface of the material by energetic particles). For example, cemented tungsten carbide typically has one quarter of the sputtering rate of pure tungsten. Cemented borides are equivalent, but using boride particles as the aggregate, rather than carbide. Borocarbide particles may also be used. The choice of carbide/boride and binder will be guided by the conditions in the reactor. The need to withstand high neutron flux prevents the use of many elements and isotopes, such as cobalt and nickel, which would become radioactive due to neutron exposure. High magnetic fields require structural considerations to be taken into account when using ferromagnetic material, as the resulting forces would cause large stresses within the reactor. Similar considerations occur for the choice of carbide. Also, the material must of course be able to reduce the flux of neutrons which reach components behind the shield. Carbon will naturally act as a moderator, slowing the fission neutrons down, which allows greater freedom of choice in the other elements that may be used (since many more elements are effective absorbers of slow neutrons than faster neutrons). Boron-10 is an effective neutron absorber. Promising candidates for the carbide are tungsten carbide, as the neutron absorption is favourable and the mechanical properties have been well studied, tungsten boride, and boron carbide, which combines the moderating properties of carbon with the neutron absorption of boron. Multiple carbides may be used in order to balance structural and neutronics properties of the material. In addition, other substances may be added to the cemented material in addition to the carbides, for example borides may be added to a predominantly carbide composite in order to introduce boron into the shielding, or vice versa. Addition of tungsten boride to a cemented tungsten carbide may improve the resistance to corrosion. Borocarbides which may be used include tungsten borocarbide, specifically a ternary tungsten borocarbide. Other substances that may be added to the material include oxides and nitrides, for example titanium nitride may be added to improve the structural properties of the material. Other alternatives to tungsten carbide or tungsten borocarbide include borides and/or carbides of elements corresponding to the sixth period of the periodic table (or beyond). The melting points of the elements increase across the sixth period, peaking at group six (tungsten). Therefore the main candidate elements are hafnium, tantalum, tungsten and rhenium. The platinum metals may be theoretically suitable for neutron shielding but are considered to be less useful because osmium compounds are highly toxic, and because of the prohibitively high cost of iridium and platinum. Rhenium is also very expensive and very rare. The three most likely candidates are therefore hafnium, tantalum and tungsten. Of these, tungsten (including its compounds) is the cheapest and most widely available, and easy to process by powder methods. Other suitable shielding materials include the pure metals of the sixth period of the periodic table, and alloys or compounds containing those metals, including composites containing an electrically conductive binder and an aggregate containing a non-conductive compound of such metals. |
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042959349 | description | The main components of the nuclear reactor are diagrammatically represented in FIG. 1. Within a concrete enclosure 2 are to be found the main vessel/4 of said nuclear reactor and a heat-exchanger 6. It is to be noted that the nuclear reactor might be provided with a plurality of cooling loops comprising, each, a heat-exchanger 6 associated to vessel 4. In the specific example described, exchanger 6 is adapted to provide the exchange of heat directly between the liquid metal issuing from the reactor vessel 4 and the steam/water. Quite obviously, it would not be going beyond the scope of the present invention to resort to an exchanger 6 adapted to exchange heat between the sodium fed into said vessel (viz. the "primary" sodium, in the present case) and secondary sodium, as is usual in reactors with loops. In particular, the present invention can be applied to a reactor of the latter type, the main vessel of which contains a primary vessel for separating the hot liquid metal from the liquid metal that has been cooled in the intermediate heat-exchangers. Pump 6' for driving the liquid metal is integrated in the exchanger with a view to rendering the whole installation less cumbersome by removing the duct connecting said devices. In the particular instance of a semi-integrated reactor, vessel 4 comprises, first, an upper portion 8 supported through supporting peripheral flanges such as 12 resting on supporting members 14 connected to concrete structure 2. The vessel is closed by a roof slab 4a resting on the upper peripheral flange of main vessel 8. Within main vessel 8 is contained a primary vessel 16 coaxial with said main vessel and provided with a bottom 18. Said primary vessel 16 is provided, along the periphery thereof, with supporting members 20 cooperating with supporting members 22 integral with the main vessel inner surface for forming a semi-tight partition. It is to be noted that the members supporting the main vessel and those supporting the primary vessel are situated substantially in the same horizontal plane. Primary vessel 16 contains core 24 resting on bottom 18 together with lateral neutronic protecting members 26. In addition, the main vessel can, optionally, be surrounded by a safety vessel 9 acting as a lining and adapted to collect sodium should said main vessel be leaking. The flow of liquid metal between vessel 4 and exchanger 6 is obtained through an outlet duct 28 connecting the inside of primary vessel 16 with the inlet of exchanger 6, and through a duct 30 for admitting cooled liquid metal, the latter duct connecting the oulet of the exchanger-pump assembly with the annular space 32 between the main vessel and the primary vessel. It can readily be seen that duct 28 along which flows a hot liquid metal that has passed through the core, passes through annular space 32 containing a less hot liquid metal. Referring now to FIG. 2, it can be seen that, according to the present embodiment, the lateral wall of primary vessel 16 is, in fact, constituted by an outer sleeve 16a and an inner sleeve 16b defines an interstice, or space, 16c. In addition lower perforated plate 18, that usually goes by the name of "core diagrid" is in fact constituted by an upper plate 18a, and a lower plate 18b connected by a perforated lateral plate 18c, thus forming a space 18 in communication with annular space 32. Into the sleeves connecting plates 18a and 18b are inserted bottom fittings 40a of fuel assemblies 40 that form the nuclear reactor core 24. Lateral wall 18c is provided with orifices, or ports, such as 18d, and sleeves 42 are provided with orifices 42a for feeding corresponding orifices in bottom fittings 40a. It is to be noted that the fuel-assembly bottom fittings have, each, a lower orifice 40b opening into a space 46 limited by sheet 48 integral with lower plate 18b, said space forming the leak manifold. The liquid metal main path is the following: cold liquid metal is driven along duct 30 by the forcing pressure of the pump (viz from 5 to 10 bars) and is fed into annular space 32. The liquid metal main flow penetrates into corediagrid 18 via orifices, or ports, 18d, then rises into fuel-assemblies 40 where is it heated. Then, the thus-heated liquid metal is evacuated towards the heat-exchanger, or exchangers, via conduit, or conduits, 28. As mentioned previously, one of the main features of the present invention lies in that the upper portion 8 of main vessel 4 is caused to cool. To that end, the annular space 32 between the main vessel and the primary vessel is divided into an upper zone 32d and a lower zone 32b, by means of a horizontal partition, the latter being constituted preferably, by the combination of supporting members 20 and 22, of the primary vessel and of the main vessel, respectively. These members therefore form a continuous horizontal ring separating space 32a from space 32b and provided, as will be specified later on, with calibrated orifices. The cold liquid metal fed through inlet duct 30 fills all the lower portion 32b of the annular space. The primary vessel upper portion is cooled by cold metal liquid introduced into upper portion 32a of annular space 32. To that end, it is first resorted to that portion of the fuel-assembly feed liquid leaking through the bottom fittings 40b and collected by leak-manifold 46. To that end, tubes 50 arranged along the periphery of diagrid 18 pass through the whole thickness of the latter. Moreover, the lower ends of sleeves 16a and 16b are welded to the core-diagrid upper sheet 18a, whereas tubes 50 open into interstice, or space 16c. In the latter interstice, there is a thermal-insulation layer 52, constituted by a plain sheet or by sticks of highly heat-conductive steel, fixed to inner sleeve 16b. Outer sleeve 16a is provided with calibrated orifices 54 opening into the upper portion 32a of annular space 32, viz. above the horizontal partition constituted by supporting members 20 and 22. In addition, a ring-shaped horizontal partition 56 mounted between sleeves 16a and 16b above orifices 54 forms a barrier adapted to restrain the liquid metal from refluxing into the space between 16a and 16b, above ring 56. Outer sleeve 16a is provided, at the upper extremity thereof, with a plurality of orifices, or ports, 16d adapted to cause annular space 32a to communicate with the inside of the primary vessel, at a slightly higher level than the liquid metal normal level in said primary vessel. This coolant circuit of the upper portion 8 of the main vessel operates as follows: the cold liquid metal at average pressure contained in lead-manifold 46 is taken up by tubes 50 and is thus fed into annular interstice 16c from which it is sent into upper zone 32a of annular space 32 via calibrated orifices 54. The liquid metal therefore fills up zone 32a from which it is evacuated through orifices 16d. The main vessel upper portion is thus cooled down to a temperature that is lower than the limit creep temperature of the steel used, viz. about 425.degree. C. On the other hand, the pouring operation at the top of the vessel permits to form a further layer N' of liquid metal over surface N of the hot liquid metal, said layer N' being "colder" than layer N and, therefore, having a lower vapour-pressure, which is of advantage since surface layer N' thus isolates the hot liquid metal from inert gas cover 16d (e.g. argon), thus restricting the metal vapour concentration in said inert gas and restraining top slab 4a from being overheated by the radiations of the liquid metal. It is to be noted that outer sleeve 16a alone provides the mechanical support of core-diagrid 18 and that sleeve 16b is used only for delimiting the hot sodium, settling thermal insulation layer 52 an channeling the leaks towards space 32a. The function of the insulation layer, however, is merely to separate the primary vessel hot liquid metal from "cold" liquid metal contained in space 32. Now, in the primary vessel, hot liquid metal is actually to be found above core 24 only, i.e. at the upper onlets of fuel-assemblies 40. That is why, according to a variant (not shown), sleeve 16b can be interrupted slightly under the core upper surface and interstice 16c is closed, at the lower portion thereof, by a ring-shaped partition similar to partition 56. In such a case, tubes 50 should be extended by means of ducts connecting said tubes with annular interstice 16c or directly with orifices 54. It is, however, to be feared that the leak-flow rate of the fuel-assembly bottom fittings which is under average pressure (e.g; 20 meter head of sodium) would not be sufficient for ensuring the cooling down of upper portion 8 of main vessel 4. This is the reason why, as is more clearly visible in FIG. 3, orifices 60 have been provided in the horizontal partition formed by supporting members 20 and 22, said orifices 60 being calibrated or, if required, adjustable by means of a device (not shown) of the needle valve type. An extra amount of cold liquid metal is thus obtained, since in zone 32b the liquid metal is substantially at the forcing pressure of the pumps. In addition, orifices 60 act as gas-vents when the main vessel is being filled with sodium. With reference to FIGS. 3 to 5, an embodiment of how hot liquid metal outlet duct 28 passes through the main vessel and the primary vessel will now be given. It can be easily understood that, at that level, two types of problems are raised, viz, first, the problem of the differential thermal expansion of both vessels and, secondly, the problem of how to isolate duct 28 carrying hot liquid metal from the cold liquid metal contained in space 32a, at least as regards that portion of said duct in annular space 32a. As far as the thermal expansion of the vessels is concerned, it is of prime importance to specify that this problem has already been almost fully solved, since, according to the invention, the supporting members 20 and 22 of the primary vessel and of the main vessel are at the same level, and, in addition, the spacing between supporting plate XX' and the duct YY'-axis is as small as possible. Therefore, the "expansion length" of both vessels between plate XX' and axis YY' is very small. However, the vessel temperature differential may be not negligible, since the hot liquid metal is at a temperature in the vicinity of 530.degree. C. (primary vessel), whereas the cold liquid metal is at a temperature of about 350.degree. C. (main vessel). As can be seen in FIG. 3, outlet piping 28 is actually welded to upper portion 8 of main vessel 4 and communication between said piping and the inside of primary vessel 16 is ensured by a small sleeve 70 passing through annular space 32. Said small sleeve 70 is attached to sleeve 16b by means of flanges 70a and 70b integral with sleeves 70 and 16b respectively. At the free end thereof, small sleeve 70 is provided with a semi-tight gasket 72 cooperating with the inner surface of tube 28. In addition, said gasket has a curved outer profile so as to form a "ball joint" in a way. Said gasket 72 is preferably constituted by a ring mounted, with some clearance, in a groove provided in the outer surface of the free end of small sleeve 70. It will be thus clearly understood that the effects resulting from radial and axial expansion of both vessels can be absorbed without generating stresses while, at the same time, minimizing the liquid metal leaks. As regards the thermal insulation of small sleeve 70, it is provided by two relatively movable annular insulating members, viz. An outer sheath 74 attached to outer sleeve 16a, on the one hand, and a cylindrical bushing 76 fixed to the outer surface of the small sleeve, on the other hand. The space 78 between these two insulating members contains stagnant liquid metal. Thus, these insulating members are not submitted to any mechanical stress resulting from the expansion of the vessels. Preferably, each of said insulating means 74 and 76 comprises two cylindrical sheets (74a, 74b and 76a, 76b, respectively) between which are mounted steel sticks 80. Annular sheets 74c, 76c, respectively, permit to immobilize said sticks. They are provided with orifices 82 adapted to prevent any pressure effect in the inner spaces defined by said sheets. Exhaust holes (not shown) are also provided in the lower portion of sheets 74a, 76a. It will thus be easily understood that a very good thermal insulation is obtained between the cold liquid metal in zone 32a and the hot liquid metal in tubing 28 and small sleeve 70, without submitting these insulating means to any thermal stress whatever, while allowing both vessels to expand freely. The combination of the above-described devices, in the present example adapted to provide the cooling of the main vessel, on the one hand, and the exhaust of hot sodium from the primary vessel, on the other hand, can be directly applied to a reactor with loops comprising a primary vessel for separating hot sodium from cold sodium. In FIGS. 1 to 3, cold liquid metal duct 30 is shown as welded to an orifice, or port, of main vessel 16 opening into annular space 32 under the semi-tight partition. A variant of interest (shown in FIG. 3a) consists in connecting duct 30 with main vessel 16 in the vicinity of the liquid metal free level in annular space 32a, and in extending same, in said annular space, by a small sleeve passing through the semi-tight partition that separates space 32a from space 32b, or else in providing flange 20 with a ring 20a adapted to define with vessel 8 an annular space 32c in communication (by means of large orifices 61 in flange 22) with annular space 32b and with annular space 32a by the clearance between collar 20a and vessel 8 (and, if required, by means of orifices 60 in either collar 20 or flange 22). That variant permits to have the orifices in ducts 28 and 30, situated above rest flange 12 of main vessel 8, which may facilitate the building of vessel 8 that can thus be stopped at a level lower than flange 12 so that ducts 28 and 30 no longer pass through said vessels. The main advantage of that variant is to prevent the suction of the liquid metal around the primary vessel, should a big leak occur in duct 30 between the main vessel and the heat-exchanger; in fact, it can be observed that such a risk is very limited, in view of the fact that said duct is provided with a protective sheath, e.g. of the type of that of the primary sodium ducts in French experimental reactor RAPSODIE. |
description | The United States Government has rights in this invention pursuant to Contract No. DE-AC03-76SF00098 between the United States Department of Energy and the University of California. This application claims benefit of earlier filing date of pending U.S. patent application Ser. No. 11/110,310 filed Apr. 19, 2005, entitled “CYLINDRICAL NEUTRON GENERATOR,” hereby incorporated by reference, which in turn claims benefit of earlier filing date of U.S. patent application Ser. No. 10/100,962 filed Mar. 18, 2002, also entitled “CYLINDRICAL NEUTRON GENERATOR,” hereby incorporated by reference, which in turn claims benefit of earlier filing date of Provisional Applications Ser. Nos. 60/276,669 filed Mar. 16, 2001 and 60/316,792 filed Aug. 31, 2001, also incorporated by reference. The invention relates to neutron tubes or sources, and more particularly to neutron tubes or sources based on plasma ion generators, including compact neutron tubes or sources which generate a relatively high neutron flux using the D-D reaction. Conventional neutron tubes employ a Penning ion source and a single gap extractor. The target is a deuterium or tritium chemical embedded in a molybdenum or tungsten substrate. Neutron yield is limited by the ion source performance and beam size. The production of neutrons is limited by the beam current and power deposition on the target. In the conventional neutron tube, the extraction aperture and the target are limited to small areas, and so is the neutron output flux. Commercial neutron tubes have used the impact of deuterium on tritium (D-T) for neutron production. The deuterium-on-deuterium (D-D) reaction, with a cross section for production a hundred times lower, has not been able to provide the necessary neutron flux. It would be highly desirable and advantageous to make high flux D-D neutron sources feasible. This will greatly increase the lifetime of the neutron generator, which is unsatisfactory at present. For field applications, it would greatly reduce transport and operational safety concerns. For applications such as mine detection, where thermal neutrons are presently used, the use of the lower energy D-D neutrons (2.45 MeV rather than 14.1 MeV) also would decrease the size of the neutron moderator. The present invention has three potential competitors for field or small-laboratory use: (1) isotopic sources based on a sample of a radioactive substance, e.g. californium-252, that emits neutrons; (2) accelerator sources, usually based on an ion source feeding a radiofrequency quadrupole (RFQ) linac and thence a neutron production target; and (3) conventional neutron tubes. Of these, the most direct and significant competitors are commercially available neutron tubes. As for the others, RFQ-based sources have never become a major commercial presence due to cost and complexity, and the safety concerns and lack of time structure that are inherent to isotopic sources limit their applications. Neutronics can identify possible explosives and nuclear materials by their composition, not just by their shape or density the way x-ray machines do. Since the September 11 terrorist attacks, detection of explosives and fissionable materials has become an urgent national need. Detecting such materials hidden in baggage or cargo is challenging under real-world conditions. Thermal neutron analysis (TNA) has been tried for inspection of checked baggage and cargo at airports. Low-energy neutrons cause nitrogen in explosives to emit gamma rays and cause fissile materials to give off neutrons of their own. The first-generation TNA screeners were too large, complex, and expensive; FAA-approved screening devices presently on the market use x-rays to look at shapes and densities, rather than using neutronics to detect actual composition. Besides the obvious considerations of cost-effectiveness and acceptable footprint, systems for inspecting baggage and cargo must offer trustworthiness (reliability combined with freedom from both false positives and false negatives), plus high throughput so that spot checks can be replaced by comprehensive inspection without bottlenecking an already heavily burdened process. Systems are also needed for relatively nonintrusive inspection of larger objects, e.g. an intermodal cargo container, or a vehicle. Detection of land mines or unexploded ordnance is another related application of great worldwide importance. A compact neutron generator design with a high neutron flux and adapted for these uses would be highly advantageous. Neutron logging instruments consist of a neutron generator and gamma-ray detector packaged so as to fit into a small (e.g. 2-inch-diameter) borehole. Analyzing the gamma ray spectrum due to neutron capture and inelastic scattering in the subsurface allows elements in the medium to be identified. Applications include oil and mineral exploration, and basic geological studies. A neutron generator design with a high neutron flux and adapted for use in a borehole would be highly advantageous. The invention is a cylindrical neutron generator formed with a coaxial RF-driven plasma ion source. A deuterium plasma (or a deuterium and tritium plasma) is produced by RF excitation in a plasma ion generator using an RF antenna. A cylindrical neutron generating target is coaxial with the ion generator and is separated therefrom by plasma and extraction electrodes which contain many slots. The plasma generator emanates ions radially over 360° and the cylindrical target is thus irradiated by ions over its entire circumference. The plasma generator and target may be as long as desired. There are two alternate basic embodiments of the neutron generator, in which the position of the plasma generator and neutron target are reversed. In one embodiment the plasma generator is in the center and the neutron target is on the outside, and in the second embodiment, the plasma generator is on the outside and the target is on the inside. The plasma generator may be either cylindrical or annular shaped, and the target is a cylinder. The neutron target surrounds the cylindrical plasma ion generator or is positioned inside the annular shaped plasma ion generator. In both cases the plasma generator and target are coaxial or concentric. The embodiment with the target on the outside is preferred since the target area is larger. A more complex embodiment of the neutron generator, which combines the two basic embodiments, has a nested configuration that is formed by nesting concentric targets and plasma regions. The nested configuration places a coaxial target both inside and outside the plasma generating region, and nests several targets and plasma generating regions to increase the neutron flux. This invention enables the generator to operate with high current density, high atomic species and practically unlimited beam size in the axial or longitudinal direction. The structure is compact and rugged, e.g. the RF antenna can form part of the plasma electrode and chamber wall. Thus the source's lifetime should be greatly increased because no weak components exist. The geometry is ideal for borehole applications. The source is ideal for many neutronic applications. Because of the increased target area, the much safer D-D reaction can be used, eliminating any tritium from the source. FIGS. 1A, 2A, 3A show the neutron source geometry of a first embodiment 10 of the invention, which has a cylindrical neutron generating target outside a cylindrical plasma ion source. Neutron generator 10 has a cylindrical plasma ion source 12 at its center. The principles of plasma ion sources are well known in the art. Conventional multicusp ion sources are illustrated by U.S. Pat. Nos. 4,793,961; 4,447,732; 5,198,677; 6,094,012, which are herein incorporated by reference. Ion source 12 includes an RF antenna (induction coil) 14 for producing an ion plasma 20 from a gas which is introduced into ion source 12. Antenna 14 is typically made of titanium tubing, which may be water cooled. For neutron generation the plasma is preferably a deuterium ion plasma but may also be a deuterium and tritium plasma. A deuterium plasma with current density of about 100 mA/cm2 can be produced. Ion source 12 also includes a pair of spaced electrodes, plasma electrode 16 and extraction electrode 18, along its outer circumference. Electrodes 16, 18 electrostatically control the passage of ions from plasma 20 out of ion source 12. Electrodes 16, 18 contain many longitudinal slots 19 along their circumferences so that ions radiate out in a full 360° radial pattern. Coaxially or concentrically surrounding ion source 12 and spaced therefrom is cylindrical target 22. Target 22 is the neutron generating element. Ions from plasma source 12 pass through slots 19 in electrodes 16, 18 and impinge on target 22, typically with energy of 120 keV to 150 keV, producing neutrons as the result of ion induced reactions. The target 22 is loaded with D (or D/T) atoms by the beam. Titanium is not required, but is preferred for target 22 since it improves the absorption of these atoms. Target 22 may be a titanium shell or a titanium coating on another chamber wall 24, e.g. a quartz tube. Flange 26 extends from the ends of chamber wall 24. The extraction apertures in electrodes 16, 18 are in the form of slots 19 whose length can be extended to any desired value. The beam impinges on the target 22 in 360.degree. and therefore the target area can be enhanced by at least 2 orders of magnitude over conventional neutron sources. Thus the same or greater neutron flux can be generated from D-D reactions in this neutron generator as can be generated by D-T reactions in a conventional neutron tube, eliminating the need for radioactive tritium. The neutrons produced, 2.45 MeV for D-D or 14.1 MeV for D-T, will also go out radially in 360°. By making the neutron generator as long as practical in the axial or longitudinal direction, a high neutron current can be obtained. FIG. 2A shows further details of neutron generator embodiment 10 from FIG. 1A. Induction coil (RF antenna) 14 is positioned inside concentric cylindrical electrodes 16, 18. Ions passing through the slots 19 in electrodes 16, 18 strike target (surface) 22. FIG. 3A shows some further details and minor variations of the design. The entire generator 10 is contained within a vacuum chamber 27 which is spaced apart from target chamber wall 24. Only a single extraction grid 18 is shown; plasma grid 16 is not needed since the ions can be extracted with a single grid. Chamber wall 24, on which target coating 22 is formed, is surrounded by target cooling coils 28. Permanent magnets 30 are arranged in a spaced apart relationship, running longitudinally along plasma ion generator 12, to from a magnetic cusp plasma ion source. The principles of magnetic cusp plasma ion sources are well known in the art, as cited above. FIGS. 1B, 2B, 3B, 3C show the neutron source geometry of a second embodiment 40 of the invention, which is similar to neutron generator 10, except that the cylindrical ion source and neutron generating target are in a reversed position, i.e. the cylindrical neutron generating target is inside the cylindrical plasma ion source. Neutron generator 40 has a cylindrical plasma ion source 42 at its outside. Ion source 42 includes an RF antenna (induction coil) 44 for producing an ion plasma 50. Ion source 42 also includes a pair of spaced electrodes, plasma electrode 46 and extraction electrode 48, along its inner circumference. Electrodes 46, 48 electrostatically control the passage of ions from plasma 50 out of ion source 42 into the interior of neutron generator 40. Electrodes 46, 48 contain many longitudinal slots 49 along their circumferences so that ions radiate in a full 360° radial pattern. Extraction electrode 48 is inside plasma electrode 46, the reverse of neutron generator 10, since the direction of plasma flow from the plasma ion source 42 is radially inward rather than radially outward, as in neutron generator 10. Ion source 42 coaxially or concentrically surrounds and is spaced from an inner cylindrical target 52. Target 52 is the neutron generating element. Ions from plasma source 42 pass through slots 49 in electrodes 46, 48 and impinge on target 52, typically with energy of 120 keV to 150 keV, producing neutrons as the result of ion induced reactions. The target 52 is loaded with D (or D/T) atoms by the beam. Titanium is not required, but is preferred for target 52 since it improves the absorption of these atoms. Neutron generator 40 is enclosed in an external chamber 54. The extraction apertures in electrodes 46, 48 are in the form of slots 49 whose length can be extended to any desired value. The beam impinges on the target 52 in 360° and therefore the target area can be enhanced. However, between neutron sources 10 and 40, for the same outside source diameter, the target in source 10 will be larger because of its larger diameter. The neutrons produced, 2.45 MeV for D-D or 14.1 MeV for D-T, will also go out radially in 360°. By making the neutron generator as long as practical in the axial or longitudinal direction, a high neutron current can be obtained. FIG. 2B shows the internal structure of neutron source 40 without chamber 54, and FIG. 3B shows the structure of FIG. 2B inside chamber 54, with flanges 56 extending from the ends of chamber 54. FIG. 3C shows a minor design change in which the RF antenna 44 is incorporated into the plasma electrode 46 and the chamber wall 54. This arrangement makes the source more compact and efficient. FIGS. 4A-B show the neutron source geometry of a third embodiment 60 of the invention, which has a nested configuration that is formed by nesting concentric neutron generating targets and plasma generating regions. The nested configuration of source 60 is a combination of sources 10, 40, placing a coaxial target both inside and outside a plasma generating region, and nesting several targets and plasma generating regions to increase the neutron flux. Except for the additional number of each component, each one is structured and functions essentially the same as in the basic embodiments. Inside a cylindrical chamber 62, a plurality of concentric or coaxial alternating targets 64 and plasma generating regions 66 are arranged. Each target 64 is a cylinder. Each plasma generating region 66 is annular and has an RF antenna (induction coil) 68 positioned therein. While four targets 64 alternating with three plasma generating regions 66 are shown, at least one plasma generating region 66 with two targets 64 are needed and any number of additional nested layers may be added depending on the desired neutron yield. Each plasma generating region 66 is bounded by extraction electrodes 70 on both its inner and outer surfaces. Extraction electrodes 70 contain longitudinal slots, as previously shown, through which ions are extracted from plasma generating regions 66 and directed onto targets 64. A deuterium plasma with current density of about 100 mA/cm2 can be produced. Chamber 60 has a water inlet 72 and water outlet 74 for circulating water or other coolant to remove heat from the targets 64, as described further herein. The targets are typically made of copper with a thin coating of titanium on the surface. The power density generated by the beam is about 600 W/cm2 which can be removed by water cooling. As shown in FIGS. 4C-D, plasma generating regions 66 and targets 64 are nested coaxially. Extraction electrodes 70 with slots 76 bound the regions 66. The width of regions 66 is d1 and the gap from electrode 70 to target 64 is d2; d1 and d2 are typically 2.5 cm. Water inlet 72 and outlet 74 in chamber 62 connect to coolant channels in the chamber wall and targets 64 so that the targets can be cooled by flowing coolant during operation. A rod 78 may extend into the center of the source inside the inner target. Samples may be placed there for irradiation. A high voltage source 80 is connected between the extraction electrodes 70 and targets 64 to extract the ions from regions 66 where they are formed and accelerate them onto the targets 64 where they are collected and react. With a gap of about 2.5 cm, and adequate pumping in the region outside the ion sources, an extraction voltage of 80 kV or higher may be used. As shown in FIGS. 4E-F, RF antennas 68 are disposed within plasma generating regions 66. As shown in FIG. 4E, a coil of antenna 68 loops around the annular region 66 in a radial plane. As shown in FIG. 4F, a plurality, e.g. three, coils of antenna 68 are spaced apart axially in different radial planes along the length of the source. The separate coils of antenna 68 are all connected together in parallel by linear elements 88 which extend axially along the source. Thus the antenna generally comprises three or more antenna loops with a bi-filar arrangement, normally connected in parallel. The antenna is typically made of titanium tubing, which may be water cooled. FIG. 5 is a graph of the neutron production cross sections of the D-D and D-T reactions. Although the D-D cross section is much lower than the D-T cross section, the large target surface area provided by the cylindrical geometry, makes a source based on the D-D reaction practical since high neutron flux can be obtained. Thus the hazards of dealing with tritium in the source can be eliminated. FIG. 6 shows an application of the compact high flux neutron source for borehole instrumentation, e.g. oil well logging. A neutron generator 90 according to the invention, in combination with a gamma ray detector 92, is lowered into a borehole 94, with necessary electrical connections made through cable 96. Neutrons emitted by generator 90 interact with features in the ground, e.g. ground contamination region 98, and produce gamma rays that are detected by detector 92. The signals are analyzed by techniques known in the art to identify subterranean features or the presence of resources. Because of the cylindrical geometry of generator 90, it can be made of a diameter suitable for a typical borehole, while its length can be selected to give a sufficiently high neutron flux to improve detection capability. The combination of simplicity, compactness, and high flux offered by the compact neutron source of the invention is also advantageous for security applications such as thermal neutron analysis (TNA) to inspect baggage at airports. Compared to other technologies for performing TNA, it is much cheaper and simpler than RFQ-based systems, substantially smaller than previous neutron tubes capable of the same flux (which is important because high neutron flux is allows high throughput), and if D-D reactions are used, not only intrinsically safer than radioactive sources, but also, in contrast to them, capable of being gated rapidly on and off to allow finer discrimination by the detector and accompanying software. FIG. 7 illustrates a system 100 for detecting explosives 102 or fissile nuclear material 104 in a suitcase 106. System 100 includes a compact neutron source 108 of the type as generally described above but which may be slightly modified to also provide neutrons in an axial direction by extending the target to cover an end of the source and extracting ions axially as well as radially from the plasma generating region. Source 108 is connected by a power cable 110 to an external power supply 112. Source 108 is surrounded by a moderator 114 which slows the neutrons generated by source 108 to thermal energy to perform TNA. The moderator is further surrounded on the sides and back end by shielding 116 which prevents neutrons from escaping, so that all emitted neutrons are emitted from the front end and can be directed to the desired inspection point. Thermal neutrons may be reflected from the shield toward the open end. Also neutron source 108 could be placed transverse to the direction shown. External gamma ray detector 118 is used to detect gamma rays emitted by explosives and external neutron detector 120 is used to detect neutrons from fissile nuclear material. As described above, the neutron tube with coaxial geometry increases the available target area, thus giving up to 1000× the neutron output of existing tubes of comparable size. Such high output from so small a device opens up new horizons for explosives detection and also for R&D applications of neutronics. In a representative small version of this highly scalable design (length 26 cm, diameter 28 cm, and weight about 18 kg) with a single target, an expected output has been calculated of ˜1.2×1012 n/s for D-D neutrons and ˜3.5×1014 n/s for D-T. The neutron flux of course depends on several parameters, including voltage, current, and the size of the tube. The ultimate limiting factor in output is the power density on the target, which is conservatively limited at ˜650 W/cm2. Tubes with nested concentric targets and plasma regions can have as much as 10× higher average neutron flux than a tube of similar size (e.g. length 35 cm, diameter 48 cm) with a single target, e.g. ˜1.6×1013 n/s for D-D neutrons and ˜4.5×1015 n/s for D-T. As a broad general principle, “comparable” conventional neutron tubes (e.g., with diameters of several tens of cm and lengths of up to a few hundred cm) produce 106 to 108 neutrons per second in D-D operation. Many neutronics applications could be improved by a small high-flux neutron source according to the invention. These include: Condensed matter physics. Scattering of slow neutrons in condensed matter (solids or liquids) can determine structure on the atomic or molecular level. Neutrons penetrate deeply into matter, enabling study of new materials in realistic temperatures, pressures, and other ambient conditions. Material science. Point defects, dislocations, interphase boundaries, intrinsic junctions with microcracks, pores, etc. can be studied. Studies of molecular compounds. Small-angle neutron scattering (SANS) is a powerful method to investigate polymer systems and surface-active substances. Specular reflection provides information about the structure along the surface. Biology. Neutrons can “see” hydrogen better than photons can, so details of the structure and function of some biological systems can be better studied. Medical applications. Boron neutron-capture therapy trials; brachytherapy. Engineering analysis. Neutron diffraction probes internal stresses in multiphase materials. Both R&D and nondestructive evaluation could benefit. Earth sciences. Neutrons can probe the texture of rocks and minerals and the effects of external pressure on the structure of samples. Changes and modifications in the specifically described embodiments can be carried out without departing from the scope of the invention which is intended to be limited only by the scope of the appended claims. |
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054066026 | abstract | A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. |
claims | 1. A method for imaging a sample by use of a charged particle beam apparatus, the method comprising the steps of:imaging a test sample with a first magnification ratio sufficient for including polyhedral patterns formed the test sample within a viewing field by applying a charged particle beam to the test sample and by detecting secondary electrons generated from the test sample, the polyhedral patterns having a known shape;imaging each polyhedral pattern with a second magnification ratio which is higher than the first magnification ratio by sequentially moving and applying the charged particle beam;obtaining first relationships between a position irradiated with the charged particle beam within the viewing field and the beam landing angle for each of the polyhedral patterns using images obtained during the step of imaging each polyhedral pattern;obtaining second relationships between other positions within the viewing field and beam landing angles through interpolation; andimaging the sample by applying the charged particle beam to the sample with a pattern formed on the surface thereof to scan the sample while controlling beam incidence conditions of the charged particle beam to allow a beam landing angle to be maintained within a viewing field based on the first and second relationships. 2. The method for imaging a sample by use of a charged particle beam apparatus as defined in claim 1, wherein a polyhedral pattern with a known shape on the test sample has a shape of a quadrangular pyramid, a quadrangular frustum, or an almost quadrangular pyramid with a round top surface. 3. The method for imaging a sample by use of a charged particle beam apparatus as defined in claim 1,wherein a plurality of polyhedral patterns with a known shape are formed on the test sample; andwherein a control quantity for controlling a beam landing angle of a charged particle beam is obtained from images of a plurality of polyhedral patterns of the test sample. 4. The method for imaging a sample by use of a charged particle beam apparatus as defined in claim 1,wherein the charged particle beam is applied to the sample with the pattern formed on the surface thereof to scan the sample while controlling the beam incidence condition of the charged particle beam; andwherein dimensions of the pattern formed on the surface thereof are obtained by use of images obtained by imaging the sample irradiated and scanned by the charged particle beam. 5. A charged particle beam apparatus comprising:a charged particle beam irradiation means for applying a focused charged particle beam to a sample to scan the sample;an image capturing means for capturing images of the sample by detecting secondary charged particles generated from the sample irradiated with a charged particle beam emitted from the charged particle beam irradiation means with a first magnification ratio and a second magnification ratio, the second magnification ratio being higher than the first magnification ratio;an image processing means for processing images captured by the image capturing means, and for obtaining first relationships between positions irradiated with the charged particle beam within a viewing field and a beam landing angle in a plurality of directions for polyhedral patterns detected with the first magnification ratio, and obtaining second relationships between other positions within the viewing field and beam landing angles through interpolation;a display means for displaying images of the sample captured by the image capturing means; anda control means for controlling the charged particle beam irradiation means, the image capturing means, the image processing means, and the display means,wherein the image processing means causes beam incidence conditions of the charged particle beam allowing a beam landing angle to be maintained within a viewing field to apply the charged particle beam to the test sample, and obtains the first relationships between a positions irradiated by the charged particle beam and the beam landing angle in a plurality of directions; andwherein the control means controls the charged particle beam irradiation means based on the first and second relationships. 6. The charged particle beam apparatus as defined in claim 5, wherein a plurality of polyhedral patterns with a known shape of a quadrangular pyramid, a quadrangular frustum, or an almost quadrangular pyramid with a round top surface are formed on the test sample. 7. The charged particle beam apparatus as defined in claim 5, wherein the image processing means processes images captured by the image capturing means, by use of the sample with the pattern formed on the surface thereof, thereby obtaining dimensions of the pattern of the sample. |
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abstract | A method is disclosed for the use of an organic admixture composed of a polysaccharide such as hydroxypropylcellulose and a monosaccharide such as ethoxylated methylglucoside and de-ionized water and minerals such as zeolites for electromagnetic; radio and microwave frequency and radioisotope shielding of building materials such as wall liners, gypsum wallboard and high performance, high strength concrete. |
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claims | 1. An X-ray scatter reducing grid comprising: a plurality of radiation-absorbing plates disposed at predetermined intervals over an entire area exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; and at least two support members for supporting opposite end portions of each of said radiation-absorbing plates, said support members being provided with plate-receiving means which receive said plurality of radiation-absorbing plates, said radiation-absorbing plates being inserted in said plate-receiving means and being supported by said support members, wherein at least some of said radiation-absorbing plates are pulled so as to be stretched in a longitudinal direction of said radiation-absorbing plates, and said radiation-absorbing plates are fixed to said support members. 2. The X-ray scatter reducing grid as set forth in claim 1 , wherein said support members are constructed by two first support members which support the opposite end portions of each of said radiation-absorbing plates and two second support members which connect to said two first support members so that said four support members constitute a rectangular frame. claim 1 3. An X-ray scatter reducing grid comprising: a plurality of radiation-absorbing plates disposed in parallel at predetermined intervals over an entire area exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; and at least two support members for supporting opposite lengthwise end portions of each of said radiation-absorbing plates, said support members being provided with plate-receiving means which receive said plurality of radiation-absorbing plates, said radiation-absorbing plates being inserted in said plate-receiving means and being supported by said support members, wherein said plate-receiving means are constructed by a plurality of slots which receive and support the opposite lengthwise end portions of each of said radiation-absorbing plates, and wherein at least some of said radiation-absorbing plates are pulled so as to be stretched in a longitudinal direction of said radiation-absorbing plates, and said radiation-absorbing plates are fixed to said support members. 4. An X-ray scatter reducing grid comprising: a plurality of radiation-absorbing plates disposed in parallel at predetermined intervals over an entire area exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; and at least two support members for supporting opposite lengthwise end portions of each of said radiation-absorbing plates, said support members being provided with plate-receiving means which receive said plurality of radiation-absorbing plates, said radiation-absorbing plates being inserted in said plate-receiving means and being supported by said support members, wherein said plate-receiving means are constructed by a plurality of elongated holes which receive and support the opposite lengthwise end portions of each of said radiation-absorbing plates, and wherein at least some of said radiation-absorbing plates are pulled so as to be stretched in a longitudinal direction of said radiation-absorbing plates, and said radiation-absorbing plates are fixed to said support members. 5. An X-ray scatter reducing grid comprising: a plurality of radiation-absorbing plates disposed at predetermined intervals over an entire area exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; and at least two support members for supporting opposite lengthwise end portions of each of said radiation-absorbing plates, said support members being provided with plate-receiving means which receive said plurality of radiation-absorbing plates, said radiation-absorbing plates being inserted in said plate-receiving means and being supported by said support members, wherein said plate-receiving means is constructed by a plurality of elongated holes which receive and support the opposite end portions of each of said radiation-absorbing plates, and wherein at least some of said radiation-absorbing plates are pulled so as to be stretched in a longitudinal direction of said radiation-absorbing plates, and said radiation-absorbing plates are fixed to said support members. 6. A method of fabricating an X-ray scatter reducing grid, comprising: inserting lengthwise opposite end portions of each of a plurality of radiation-absorbing plates into plate-receiving means formed in at least two support members, said radiation-absorbing plates being disposed in parallel at predetermined intervals over an entire area to be exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; supporting the lengthwise opposite end portions of each of said radiation-absorbing plates by said plate-receiving means; fixing said radiation-absorbing plates to said plate-receiving means by at least one of adhering, fusing, and press-fitting, pulling at least some of said radiation-absorbing plates so as to stretch pulled ones of said radiation-absorbing plates in a longitudinal direction of said radiation-absorbing plates; and fixing said pulled ones of said radiation-absorbing plates to said support members. 7. A method of fabrication an X-ray scatter reducing grid, comprising: inserting lengthwise opposite end portions of each of a plurality of radiation-absorbing plates into plate-receiving means formed in at least two support members, said radiation-absorbing plates being disposed in parallel at predetermined intervals over an entire area to be exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; supporting the lengthwise opposite end portions of each of said radiation-absorbing plates by said plate-receiving means; wherein said X-ray scatter reducing grid includes at least one of a ceiling plate and a bottom plate, the method further comprising fixing said radiation-absorbing plates to at least one of said ceiling plate and said bottom plate, pulling at least some of said radiation-absorbing plates so as to stretch pulled ones of said radiation-absorbing plates in a longitudinal direction of said radiation-absorbing plates; and fixing said pulled ones of said radiation-absorbing plates to said support members. 8. An X-ray scatter reducing grid comprising: a plurality of radiation-absorbing plates disposed in parallel at predetermined intervals over an entire area exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; and two support members for supporting opposite end portions of each of said radiation-absorbing plates, said support members being provided with plate-receiving means which receive said plurality of radiation-absorbing plates, said radiation-absorbing plates being inserted in said plate-receiving means and being supported by said support members, wherein said radiation-absorbing plates are fixed to said two support member, elastic bodies being interposed between said two support members so that said two support members are urged in a direction in which said radiation-absorbing plates are stretched. 9. The X-ray scatter reducing grid as set forth in claim 8 , wherein said support members are constructed by two first support members which support the opposite end portions of each of said radiation-absorbing plates and two second support members which connect to said two first support member so that said four support members constitute a rectangular frame. claim 8 10. An X-ray scatter reducing grid comprising: a plurality of radiation-absorbing plates disposed at predetermined intervals over an entire area exposed to radiation, each radiation-absorbing plate comprising a radiation-absorbing substance and having width in a direction in which said radiation travels; and at least two support members for supporting opposite lengthwise end portions of each of said radiation-absorbing plates, said support members being provided with plate-receiving means which receive said plurality of radiation-absorbing plates, said radiation-absorbing plates being inserted in said plate-receiving means and being supported by said support members, wherein said plate-receiving means is constructed by a plurality of elongated holes which receive and support the opposite end portions of each of said radiation-absorbing plates, and wherein said radiation-absorbing plates are fixed to said two support members, elastic bodies being interposed between said two support members so that said two support members are urged in a direction in which said radiation-absorbing plates are stretched. |
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047822317 | summary | This invention relates to a standard component .sup.99m Tc elution generator useful for medical purposes and consisting of prefabricated component parts. The fundamental part of the generator is formed by the main generator-ampoule-column made from materials with low radiactivation ability by neutrons. It serves first as the reactor irradiation ampoule and after suitable adjustment directly as proper generator column. The ampoule-column is filled with the target material--elution matrix, that contains high amounts of molybdenum (up to 40% by weight) consisting of water insoluable molybdates or polymolybdates releasing easily .sup.99m Tc technetium generated in the column by radioactive decay of mother .sup.99 Mo, formed in the matrix structure by neutron activation. The filling of the ampoule-column serves originally as target material for reactor irradiations and afterwards it is directly used as the proper elution matrix of the generator. The high content of molybdenum in the target material-elution matrix makes possible the utilization even of low to medium power reactors for neutron activation. The production possibility of the main generator column as completely inactive material already before the reactor irradiation facilitates substantially the post-irradiation assembly procedure. In such a way the manufacturer can supply not only the complementary components but even the main generator ampoule-column as inactive parts, that can be produced highly effective in large batches and kept for a long time on stock. BACKGROUND OF THE INVENTION .sup.99m Tc is a widely used radionuclide in radiopharmaceutical and nuclear medicine applications. The particular medical advantage of this radionuclide is its very short half-life of about 6 hours. However, the short half-life creates manufacture and delivery problems, because the radionuclide must be used very soon after it is produced. For this reason, .sup.99m Tc is preferably supplied to hospitals on demand by an on-site generator, through disintegration of isotopic molybdenum (.sup.99 Mo) and chemical separation of the .sup.99m Tc product. High purity and high activity are important, so that the .sup.99m Tc product may be used immediately as a pertechnate, in the preparation of radionuclide tracer compounds, etc. Current medical technology requires the use of radionuclide generators which can supply radioactive levels of at least 4 GBq, most often 8-12 GBq, and in some cases as high as 40 GBq per generator. Of radionuclide the generators now in use, elution generators are the most advantageous because they provide for rapid, efficient, and simple production of the desired radionuclides. However, in practice, most elution generators rely on aluminum oxide as a sorption material, which has a sorption capacity of only several percent by weight of molybdenum. This limits the activity of the generator to only several hundred MBq when natural isotopes of molybdenum are irradiated by a medium neutron flux ranging approximately between 10.sup.17 and 10.sup.18 n/m.sup.2 s. This level of activity is insufficient for medical applications. Known elution generators must therefore rely on either (a) the irradiation of enriched .sup.98 Mo as a target material for a high-intensity neutron flux; or (b) a carrier-free .sup.99 Mo isotope obtained by fission of uranium. These known devices and processes require a large capital investment; high energy and labor costs; a complex series of processing and purification steps involving highly radioactive components; the separation of .sup.99 Mo from uranium fission products, which are about twenty times more active than the useable radionuclide. Thus, for many applications the advantages of known elution generators have been outweighed by the practical and economic disadvantages, and other types of generators have been sought. Methods of .sup.99m Tc production allowing its separation from low specific activity parent .sup.99 Mo are to be used. For example, a sublimation of .sup.99m Tc may be obtained from a suitable .sup.99 Mo compound. Or, .sup.99m Tc products may be extracted from a strongly alkaline aqueous solution of a molybdate of .sup.99 Mo by methylethylketone. These processes permit the production of .sup.99m Tc products of medically sufficient activity from low to medium neutron flux irradiation within the range of about 5.times.10.sup.16 to 5.times.10.sup.17 n/m.sup.2 s. However, these chemical methods are substantially more complex, time-consuming, and labor intensive than a common elution generator. They may not be economically and conveniently practiced within a self-contained on-site apparatus. Instead, a centralized manufacturing and processing center will generally supply technetium 99m produced by sublimation or extraction to local hospitals and clinics. Although miniaturized extraction and sublimation generators are available for on-site hospital use, they remain complex and expensive. The technology of known .sup.99m Tc generators is discussed in R. E. Boyd, Recent Developments in Generators of .sup.99m Tc Radiopharmaceutical and Labeled Compounds, IAEA (Vienna: 1973), p. 1-26; R. E. Boyd, Technicium 99m Generators--The Available Options, Int. J. Appl. Rad. & Isot. (New York: 1982), Vol. 33, p. 801-2; and V. J. Molinsku, A Review of .sup.99m Tc Generator Technology, Int. J. Appl. Rad. & Isot. (New York: 1982), Vol. 33, p. 811-19. Other practitioners have sought to improve elution generators by replacing the aluminum oxide sorption material with a sorption matrix. The matrix is intended to recover greater amounts of molybdenum from which .sup.99m Tc can be eluated, thus improving efficiency and yield. J. V. Evans, P. W. Moore, M. S. Shying, & J. M. Soddeau, "A New Generator For .sup.99m Tc," Third World Congress on Nuclear Medicine and Biology, pp. 1592-5 (Paris: 1982). The Evans device uses a sorption matrix of zirconium molybdate obtained from irradiated molybdenum oxide that is dissolved in a lye solution, precipitated by zirconium nitrate, and dried at 105.degree. C. The approximate chemical composition of this sorption material is ZrO.sub.2.MoO.sub.3.xH.sub.2 O, having a molybdenum concentration of approximately 25% by weight. Through a hydration and shaping process, the material achieves elution of .sup.99m Tc from .sup.99 Mo with an efficiency of 70-90%. In experiments performed by the inventors in addition to zirconium molybdate also with titanium molybdate and polymolybdates of both elements containing 10-40% by weight molybdenum (with preferred content 20-30%) elution efficiencies 40-80% have been achieved. In contrary to previous authors (Evans et al) the elution matric was not made from already previously irradiated radioactive material dried at 105.degree. followed by hydration but in our case the elution matrix has been made from completely inactive material and dried at lower temperatures, prior to its activation by neutrons in the reactor. The drying has been performed at 40.degree.-50.degree. C., lasting many hours, in some cases even drying at room temperature (approximately. 20.degree. C.), lasting many days, has been used. The grain size of the matrix material usually has been in the range 50-140 /um. The inactive matrix prepared in such a way has been directly used as target material for the exposure to neutron activation in the reactor prior to elution. SUMMARY OF THE INVENTION It is an object of the invention to overcome or minimize a number of disadvantages of known .sup.99m Tc devices. These disadvantages include: (a) the limited availability of enriched .sup.98 Mo and of reactors with a high neutron flow; (b) the practical and economic difficulties inherent in elution devices based on .sup.99 Mo obtained from fission products; (c) MAAE control and disposal of used fission material and fission byproducts; (d) the risk of radioactive contamination; (e) the complexity and cost of sublimation and extraction processes and devices; (f) transport and delivery problems involving .sup.99m Tc products, activated generators, and fully assembled generators; (g) reliance on a neutron flow exceeding 10.sup.18 n/m.sup.2 s to achieve elution generator activity above 2 GBq from n, gamma reaction with .sup.99 Mo; and (h) the need for special sterilization procedures. According to the invention, these problems are alleviated or eliminated by an elution generator made of independent component parts. The major component is a main column which serves first as an irradiation ampoule for activation of the target material, and then is adjusted for immediate use as an elution column containing the activated material serving directly as elution matrix. The main column is preferably cylindrical and is provided with a supply means and a discharge means, such as tubes or hoses. It is composed of a corrosion-resistant material that is relatively inert with respect to neutrons, such as aluminum, zirconium, or quartz. The column is filled with target material containing at least 10% molybdenum by weight, so that efficent selective elution of .sup.99m Tc from .sup.99 Mo will result. Powdered or granular molybdates are generally used, such as polymolybdates of zirconium and/or titanium. The particles are held within the column by a porous material that is relatively unaffected by neutrons, such as a porous sinter of silicon or zirconium oxide, graphite, felt, quartz, or an aluminum fiber composite. The porous material allows the elution solution to pass freely. The main column can be hermetically sealed for radiation sterilization during activation by the neutron flow. Before exposure in the reactor, the ends of the column are hermetically sealed, such as by fusing, aluminum packing, or screw-type closures with aluminum packing. In addition, the entire column can be wrapped with aluminum foil to prevent bacterial contamination after removal from the reactor and to provide an aseptic connection with other components of the apparatus. After irradiation, the sealed ends of the column are breached and the supply and discharge tubes are attached in a sterile manner. The opposing ends of the supply and discharge tubes are sealed or plugged, so that the interior of the column is maintained in a sterile condition. The column, together with connecting means, is placed in a container for transportation to the user, preferably of lead or uranium deprived of .sup.235 U. The primary container is placed in another secondary protective container which is hermetically sealed. The other components of the assembled device, also sterilized and protected against contamination, comprise a vessel containing an apyrogenous elution solution, preferably 0.9% NaCl by weight, in sterile connection with the supply tube; a protective column filled with a sorbent, preferably zirconium oxide or aluminum oxide, and connected to the discharge tube; a piercing head with a connecting hose from the discharge tube; and evacuated bottles for the eluate connected to the piercing head. The apparatus also comprises a laboratory screening container and a support base for the assembled components. The apparatus can be delivered assembled or unassembled. Unassembled delivery is preferable, because the main column can then be separately manufactured, processed, and delivered. Table I shows the activity of the new elution generator at different column volumes and different neutron flows, when used with a target material containing 25% molybdenum by weight in a sorption matrix at a bulk weight of 1 g/ml. The activities are related to .sup.99 Mo and to a reference date 72 hours after a prior continuous irradiation for 90 hours. TABLE I ______________________________________ .sup.99 Mo ACTIVITY OF .sup.99m Tc GENERATOR IN GBq neutron flow is in n/m.sup.2 s; volume is in ml; dimensions are in cm. MAIN COLUMN .sup.99 Mo ACTIVITY IN GBq VOL- HEIGHT W AT GIVEN UME GIVEN DIAMETER NEUTRON FLUX .times. 10.sup.17 ml 1.0 cm 1.5 cm 2.0 cm 5 .times. 10.sup.-1 1 2 5 ______________________________________ 3 3.8 1.7 -- 0.5 1 2.5 6.5 5 6.4 2.8 -- 1 2 4 11 10 12.7 5.6 3.2 2 4 8 22 20 -- 11.3 6.4 3.5 8 17 43 30 -- -- 9.5 5 12 25 65 ______________________________________ When the main column is made of zirconium, both it and the elution matrix are activated. Due to the 10:1 ratio of Zr to Mo by weight, and due to the activating cross sections, the irradiation time, and the lethal time, the main column exhibits .sup.97 Zr activity with a half-life of 17 hours, and which is not more than double the .sup.99 Mo activity at the time of irradiation. At the reference time of 72 hours, .sup.97 Zr activity is roughly 20% of the .sup.99 Mo activity. The column also exhibits a .sup.95 Zr activity having a half-life of 64 days in equilibrium with .sup.95 Nb having a half-life of 35 days, representing roughly 10% the activity of .sup.99 Mo after irradiation and 20% at the reference time. These conditions create no significant or substantial hazards or problems with respect to construction and screening of the generator. When the main column is made of aluminum, the total radioactivity is lower, as only the target material is activated. The half-life of .sup.28 Al is 2.2 minutes, and therefore nuclear-clean aluminum creates no lingering radioactive byproducts from the neutron irradiation, and no column activity remains at the reference time. Column contaminants can be activated by neutron irradiation, and in addition aluminum itself may become activated by fast neutrons via .sup.27 Al/n,alfa/.sup.24 Na reaction to natrium (.sup.24 Na). Natrium has a half-life of 15 hours. From an irradiation point of view the most preferable column material is quartz, where only very small amounts of .sup.31 Si (half-life=2.6 hours) are activated. Quartz is also chemically inert and is low in contaminants; but quartz columns tend to be fragile. A neutron flux of 10.sup.16 n/m.sup.2 s corresponds to an irradiation dose of 360 kGy during one hour. All microorganisms and their latent forms are destroyed by irradiation at all doses exceeding 360 kGy. Thus, irradiation for more than one hour in a moderate neutron flow, as disclosed herein, serves the dual even purpose of target sorption material activation and reliable column sterilization. |
summary | ||
043292488 | claims | 1. A process for immobilizing high level nuclear waste containing a major proportion of aluminium and/or iron compounds which comprises the steps of (1) mixing the waste with a minor proportion of a mixture of oxides selected from the group consisting of TiO.sub.2, ZrO, SiO.sub.2, Al.sub.2 O.sub.3, CaO, SrO and BaO, at least one of the selected oxides being from the group consisting of TiO.sub.2, ZrO.sub.2 and SiO.sub.2, the oxides in said mixture and the relative proportions thereof being selected so as to form a mixture which when heated at temperatures between 800.degree. and 1400.degree. C. crystallizes to produce a mineral assemblage containing (i) crystals belonging to or possessing structures closely related to the titanate mineral classes capable of providing lattice sites in which the fission product and actinide elements of said waste are securely bound, and (ii) crystals thermodynamically compatible with said crystals (i) comprising at least one non-radioactive phase containing aluminium and/or iron, said crystals (i) and (ii) belonging to or possessing crystal structures closely related to crystals belonging to mineral classes which are resistant to leaching and alteration in geologic environments; and (2) heating at a temperature within said range and then cooling said mixture under reducing conditions so as to cause crystallization of the mixture to a mineral assemblage having the fission product and actinide elements of said waste incorporated as solid solutions within the crystals (i) thereof, and aluminium and/or iron crystallized in said at least one non-radioactive crystal phase (ii). 2. A process according to claim 1, wherein said waste is mixed with from about 20 to 40% by weight of said mixture of oxides. 3. A process according to claim 1, wherein said heating and cooling is carried out under reducing conditions such that said iron is maintained dominantly in a divalent state. 4. A process according to claim 1, wherein said mineral assemblage contains crystals belonging to or possessing structures closely related to the mineral classes selected from the group consisting of perovskite (CaTiO.sub.3), zirconolite (CaZrTi.sub.2 O.sub.7), and a hollandite-type mineral (BaAl.sub.2 Ti.sub.6 O.sub.16). 5. A process according to claim 1, wherein said mineral assemblage comprises crystals belonging to or possessing structures closely related to at least one of the mineral classes selected from the group consisting of perovskite (CaTiO.sub.3) and zirconolite (CaZrTi.sub.2 O.sub.7 -CaUTi.sub.2 O.sub.7 solid solution). 6. A process according to claim 1, wherein said crystals (ii) include at least one phase selected from the group consisting of hercynite (FeAl.sub.2 O.sub.4), ferrite ((NiFeMn)Fe.sub.2 O.sub.4) and ulvospinel (Fe.sub.2 TiO.sub.4) and their solid solutions, ilmenite (FeTiO.sub.3), pseudobrookite solid solutions (Al.sub.2 TiO.sub.5 --Fe.sub.2 TiO.sub.5), hollandite solid solutions (BaAl.sub.2 Ti.sub.6 O.sub.16 --Ba(FeTi)Ti.sub.6 O.sub.26), a davidite-type mineral (BaAl.sub.2 Fe.sub.8 Ti.sub.13 O.sub.38) and corundum (Al.sub.2 O.sub.3). 7. A process according to claim 1, wherein said at least one non-radioactive phase includes hercynite-rich spinel or ferrite spinel. 8. A process according to claim 1, wherein said mixture of oxides comprises at least three members selected from the group consisting of TiO.sub.2, ZrO.sub.2, Al.sub.2 O.sub.3, CaO, SrO and BaO, at least one of said members being selected from the subgroup consisting of TiO.sub.2 and ZrO.sub.2. 9. A process according to claim 8, wherein said mixture of oxides comprises at least two members selected from the group consisting of TiO.sub.2, ZrO.sub.2, Al.sub.2 O.sub.3 and CaO, at least one of said members being selected from the subgroup consisting of TiO.sub.2 and ZrO.sub.2. 10. A process according to claim 1 wherein the waste contains Al.sub.2 O.sub.3 in excess of Fe.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2 and CaO in proportions chosen so that the mineral assemblage comprises hercynite-rich spinel, perovskite and zirconolite. 11. A process according to claim 1 wherein the waste contains Al.sub.2 O.sub.3 in excess of Fe.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2 and CaO in proportions chosen so that the mineral assemblage comprises hercynite-rich spinel and zirconolite. 12. A process according to claim 1 wherein the waste contains Fe.sub.2 O.sub.3 in excess of Al.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2 and CaO in proportions chosen so that the mineral assemblage comprises ferrite spinel, perovskite and zirconolite. 13. A process according to claim 1 wherein the waste contains Fe.sub.2 O.sub.3 in excess of Al.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2 and CaO in proportions chosen so that the mineral assemblage comprises ferrite spinel and zirconolite. 14. A mineral assemblage containing immobilized high level nuclear waste containing a major proportion of aluminium and/or iron compounds, said assemblage comprising crystals (i) belonging to mineral classes which are resistant to leaching and alteration in geologic environments having a fission product and actinide elments of said nuclear waste incorporated as solid solutions within the crystals thereof, said crystals (i) comprising crystals belonging to or possessing structures closely related to at least one of the mineral classes selected from the group consisting of perovskite (CaTiO.sub.3) and zirconolite (CaZrTi.sub.2 O.sub.7 --CaUTi.sub.2 O.sub.7 solid solution), and crystals (ii) thermodynamically compatible with said crystals (i) containing aluminum and/or iron crystallized in at least one non-radioactive phase. 15. A process for immobilizing high level nuclear waste containing high concentrations of Al, Fe, Mn, Ni and Na compounds which compounds constitute a major proportion of the waste which comprises the steps of (1) mixing the waste with a minor proportion of a mixture of oxides selected from the group consisting of TiO.sub.2, ZrO, SiO.sub.2, Al.sub.2 O.sub.3, CaO, SrO and BaO, at least one of the selected oxides being from the group consisting of TiO.sub.2, ZrO and SiO.sub.2, the oxides in said mixture and the relative proportions thereof being selected so as to form a mixture which when heated at temperatures between 800.degree. and 1400.degree. C. crystallizes to produce a mineral assemblage containing (i) crystals belonging to or possessing structures closely related to the titanate mineral classes capable of providing lattice sites in which the fission product and actinide elements of said waste are securely bound, and (ii) crystals of at least one non-radioactive phase containing aluminium, iron, manganese, nickel and sodium, said crystals (ii) including crystals belonging to or possessing structure closely related to the nepheline (NaAlSiO.sub.4) mineral class, said crystals (i) and (ii) belonging to or possessing crystal structures closely related to crystals belonging to mineral classes which are resistant to leaching and alteration in geologic environments, and (2) heating at a temperature within said range and then cooling said mixture so as to cause crystallization of the mixture to a mineral assemblage having the fission product and actinide elements of said waste incorporated as solid solutions within the crystals (i) thereof, and the aluminium, iron, manganese, nickel and sodium crystallized in the crystals (ii), said heating and cooling being conducted under redox conditions such that the manganese and nickel are dominantly present in the divalent state. 16. A process according to claim 15, wherein said waste is mixed with from 20 to 40% by weight of said mixture of oxides. 17. A process according to claim 15, wherein said heating and said cooling are carried out at reducing conditions such that said manganese and/or nickel are maintained dominantly in a divalent state and said iron is maintained dominantly in a divalent or trivalent state. 18. A process according to claim 17, wherein said reducing conditions are such that the oxygen fugacity lies near the nickel-nickel oxide buffer. 19. A process according to claim 15, wherein said crystals (i) comprise crystals belonging to or possessing structures closely related to the mineral classes selected from the group consisting of perovskite (CaTiO.sub.3), zirconolite (CaZrTi.sub.2 O.sub.7), and a hollandite-type mineral (BaAl.sub.2 Ti.sub.6 O.sub.16). 20. A process according to claim 15, wherein said crystals (i) comprise crystals belonging to or possessing structures closely related to at least one of the mineral classes selected from the group consisting of perovskite (CaTiO.sub.3) and zirconolite (CaZrTi.sub.2 O.sub.7 --CaUTi.sub.2 O.sub.7 solid solution). 21. A process according to claim 15, wherein said crystals (ii) comprise at least one phase selected from the group consisting of hercynite-rich spinel (Fe.sup.II Al.sub.2 O.sub.4), corundum (Al.sub.2 O.sub.2), pseudobrookite solid solutions (Al.sub.2 TiO.sub.5 --FeTi.sub.2 O.sub.5), and hollandite solid solutions (BaAl.sub.2 Ti.sub.6 O.sub.16 --Ba(FeTi) Ti.sub.6 O.sub.16). 22. A process according to claim 15, wherein said crystals (ii) comprise at least one phase selected from the group consisting of ferrite-spinel (composed principally of the end members Ni, Fe.sub.2.sup.II O.sub.4 --MnFe.sub.2.sup.III O.sub.4 Fe.sup.II Fe.sub.2.sup.III O.sub.4 --Fe.sub.2.sup.II TiO.sub.4 --Fe.sup.II Al.sub.2 O.sub.4), ilmenite (FeTiO.sub.3), ulvospinel (Fe.sub.2 Ti.sub.3 O.sub.4), ferropseudobrookite (FeTi.sub.2 O.sub.5), hollandite (Ba(Al,Fe.sup.III,Fe.sup.II,Ni,Ti).sub.2 --Ti.sub.6 O.sub.16) and a davidite-type mineral (Ba(Fe.sup.III,Al).sub.2 --Fe.sub.8.sup.II Ti.sub.13 O.sub.38). 23. A process according to claim 15, wherein said crystals (ii) include phercynite-rich spinel or ferrite spinel. 24. A process according to claim 15, wherein said mixture of oxides comprises at least four members selected from the group consisting of TiO.sub.2, ZrO.sub.2, SiO.sub.2, Al.sub.2 O.sub.3, CaO, SrO, BaO, at least one of said members being selected from the subgroup consisting of TiO.sub.2, ZrO.sub.2 and SiO.sub.2. 25. A process according to claim 24, wherein said mixture of oxides comprises at least three members selected from the group consisting of TiO.sub.2, ZrO.sub.2, SiO.sub.2, Al.sub.2 O.sub.3, CaO, at least two of said members being selected from the subgroup consisting of TiO.sub.2, ZrO.sub.2 and SiO.sub.2. 26. A process according to claim 15 wherein the waste contains Al.sub.2 O.sub.3 in excess of Fe.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2, CaO and SiO.sub.2 in proportions chosen so that the mineral assemblage comprises hercynite-rich spinel, perovskite, zirconolite and nepheline. 27. A process according to claim 15 wherein the waste contains Al.sub.2 O.sub.3 in excess of Fe.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2, CaO and SiO.sub.2 in proportions chosen so that the mineral assemblage comprises hercynite-rich spinel, zirconolite and nepheline. 28. A process according to claim 15 wherein the waste contains Fe.sub.2 O.sub.3 in excess of Al.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2, Al.sub.2 O.sub.3, CaO and SiO.sub.2 in proportions chosen so that the mineral assemblage comprises ferrite spinel (Mn,Ni,Fe).sup.II Fe.sub.2.sup.III O.sub.4, perovskite, zirconolite and nepheline. 29. A process according to claim 15 wherein the waste contains Fe.sub.2 O.sub.3 in excess of Al.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2, Al.sub.2 O.sub.3, CaO and SiO.sub.2 in proportions chosen so that the mineral assemblage comprises ferrite spinel, zirconolite and nepheline. 30. A process according to claim 15 wherein the waste contains Al.sub.2 O.sub.3 in excess of Fe.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2, CaO and SiO.sub.2 in proportions chosen so that the mineral assemblage comprises hercynite-rich spinel, perovskite, zirconolite, nepheline and a pseudobrookite-type solid solution (Al.sub.2 TiO.sub.5 -FeTiO.sub.5). 31. A process according to claim 15 wherein the waste contains Al.sub.2 O.sub.3 in excess of Fe.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2, CaO, BaO and SiO.sub.2 in proportions chosen so that the mineral assemblage comprises hercynite-rich spinel, perovskite, zirconolite, nepheline and a hollandite type solid solution (BaAl.sub.2 Ti.sub.6 O.sub.16 --Ba(Fe,Ni,Mn,Ti).sub.2 --Ti.sub.6 O.sub.16). 32. A process according to claim 15 wherein the waste contains Fe.sub.2 O.sub.3 in excess of the Al.sub.2 O.sub.3 on a weight basis and the mixture of added oxides comprises TiO.sub.2, ZrO.sub.2, Al.sub.2 O.sub.3, CaO and SiO.sub.2 in proportions chosen so that the mineral assemblage comprises ferrite spinel (Mn,Ni,Fe).sup.II Fe.sub.2.sup.III O.sub.4, perovskite, zirconolite, nepheline, ilmenite (FeTiO.sub.3) and pseudo-brookite solid solution (FeTi.sub.2 O.sub.5 --Al.sub.2 TiO.sub.5). 33. A process according to claim 32 wherein the mixture of added oxides also comprises BaO and the mineral assemblage also comprises a complex davidite-type mineral Ba(Al,Fe.sup.III).sub.2 --Fe.sub.8.sup.II Ti.sub.13 O.sub.38. 34. A process according to claims 1 or 15 wherein the selected mixture of oxides is mixed directly with a high level nuclear waste sludge without preliminary drying or calcining of the sludge. 35. A mineral assemblage containing immobilized high level nuclear waste containing Al, Fe, Mn, Ni and Na compounds, said compounds constituting a major proportion of said waste, said assemblage comprising crystals (i) belonging to mineral classes which are resistant to leaching and alteration in geologic environments and having fission product and actinide elements of said waste incorporated as solid solutions within the crystals thereof, said crystals (i) belonging to or possessing crystal structures closely related to at least one of the mineral classes selected from the group consisting of perovskite (CaTiO.sub.3) and zirconolite (CaZrTi.sub.2 O.sub.7 --CaUTi.sub.2 O.sub.7 solid solution), and crystals (ii) containing Al, Fe, Mn, Ni and Na, said crystals (ii) including crystals possessing crystal structures belonging to or closely related to the nepheline (NaAlSiO.sub.4) mineral class. 36. A mineral assemblage according to claim 35, wherein said crystals (ii) include hercynite-rich spinel or ferrite spinel. |
abstract | To provide an apparatus for treating a radioactive nitrate waste liquid that includes a denitrification tank (12A) which accommodates active sludge that adsorbs or takes in the radioactive substance in a nitrate waste liquid (11) and in which an anaerobic microorganism that reduces the nitrate to nitrogen gas grows, and a reaeration tank (14) in which a denitrification-treated liquid (24) treated in the denitrification tank (12A) is aerated and mixed with active sludge. A pH adjuster (21), a carbon source (22), and nitrogen gas are supplied to the denitrification tank (12A) so as to separate a denitrified liquid into a solid content and the denitrification-treated liquid (24) by using a first solid-liquid separating film (25), and the denitrification-treated liquid (24) treated with the active sludge in the reaeration tank (14) is reaerated and separated into a solid content and a reaeration-treated liquid (27) by using a second solid-liquid separating film (28). |
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