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abstract | A power cycle test method for testing an electronic equipment (30) includes: configuring a total test count and a current test count; updating the current test count by incrementing the current test count by a value; utilizing a corresponding AC control signal, a corresponding DC control signal, and a reboot control signal to control the electronic equipment in sequence; checking whether the electronic equipment is in a workable condition when the electronic equipment is respectively controlled under the control signals; repeating the updating step, the utilizing step and the checking step until the current test count is equal to the total test count; and generating a result message if the current test count is equal to the total test count. A related system is also disclosed. |
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description | This application claims the benefit of DE 10 2010 014 002.3, filed Apr. 7, 2010. The present embodiments relate to a method for operating a particle therapy system and the particle therapy system. Particle therapy systems are used for treating tumors using heavy particles such as, for example, protons or carbon ions. This includes irradiating a tumor with heavy particles using a raster scan method, for example. As well as tumors of patients, phantoms may also be irradiated (e.g., for research or maintenance purposes). The basis of the raster scan method is that the intensity of the radiation delivered by a particle generation and acceleration device is set for each ISO energy layer, and the dose applied to each raster scanning point is acquired in realtime and maintained for each scanning point until the planned target dose is reached. The radiation may be switched on and off with the aid of fast magnets, for example. Since a minimum time per scanning point is used while the radiation is applied to a scanning point, the minimum time, during which the radiation is applied to a scanning point at a minimum dose, determines the maximum beam intensity that may be used. The technological schemes for extracting the particle beam from the particle generator and accelerator device may lead to particle stream profiles that are subject to great fluctuations. Typical schemes for extracting the particle beams are, for example, resonance extraction or extraction using a knock-out exciter (e.g., KO exciter). The generated particle stream profiles are dominated by rapid fluctuations in the range of several microseconds and change (e.g., at the start of the extraction in addition to a time domain of several tens of milliseconds up to several seconds). FIG. 2 shows a variation in intensity over time of a particle beam according to the prior art. The intensity was measured at a resolution of 50 μs. Both the long-term fluctuations and the very short-lived fluctuations are shown. A method for emitting radiation of a charged particle beam and an acceleration device are known from JP-11329800. The acceleration device includes a device for impressing a high frequency, a deflecting device for an emission, a current measuring device and a calculating machine. The device for impressing the high frequency generates a high-frequency electric field, magnetic field or electromagnetic field on the basis of a high-frequency signal in order to impress the high-frequency field onto a charged particle stream. The deflecting device emits the charged particle beam that is moved outside of a resonance stability limit by the device for impressing the high frequency. The current measuring device measures a value of a current of the charged particle beam that is emitted by the deflecting device for an emission. The calculating machine determines an intensity of the high-frequency signal in accordance with the value of the current measured by the current measuring device. Due to the strong fluctuations in particle beam intensity, which is evident in, for example, resonance extraction, radiation technology is currently implemented such that irradiation intensities exhibiting very high variability are accepted. Relatively long radiation exposure times are accepted that are dominated by the dose of scanning points having the lowest dose. In order to obtain acceptable radiation exposure times, the minimum doses per scanning point are set to a relatively high level at the planning stage. In this way, extreme cases, in which the duration of the beam application becomes very long due to the restriction to the small intensity, may be avoided. In certain situations, disproportionately high doses may be accepted at scanning points with a low dose requirement. It would be desirable if a particle beam intensity (e.g., a number of projectiles or particles per time unit) runs in approximately the same shape as the rectangular function shown in FIG. 1. In other words, it would be desirable if the particle beam intensity rises very steeply to the desired intensity at a predetermined time instant and declines, again very steeply, after a predetermined time of, for example, 5 seconds. The present embodiments may obviate one or more of the drawbacks or the limitations in the related art. For example, the operation of a particle therapy system may be improved such that radiation exposure times are shortened and both small and large radiation doses may be applied at different scanning points. In one embodiment, a method for operating a particle therapy system is provided. The method includes generating and accelerating irradiation particles and generating a particle beam from at least one portion of the generated irradiation particles. A particle beam intensity of the particle beam is measured automatically. A plurality of scanning points is automatically irradiated sequentially with the particle beam in accordance with a predefined irradiation plan. In the irradiation plan, each scanning point of the plurality of scanning points is assigned a setpoint value for the particle beam intensity at the scanning point. The particle beam intensity is influenced automatically as a function of the measured particle beam intensity and the individually specified setpoint value for the particle beam intensity of the scanning point that is to be irradiated. Because the particle beam intensity is measured and influenced or adjusted automatically in accordance with the predefined setpoint value, variations in radiation intensity as a function of the time may be realized. The variations in radiation intensity, for example, come very close to the desired waveform shown in FIG. 1, having a very steep rise, followed by a constant progression and a fast decline. Detectors used in particle therapy systems for measuring the dose applied per scanning point may be used, for example, for automatic sensing of the particle beam intensity. There is no need to introduce additional detectors that adversely affect the beam quality and consequently the result of the medical treatment into the particle beam. The beam intensity measured by the detectors may be passed on in digital form, for example, in realtime to a digital controller. Standard industrial bus signals may be used, for example, for transmitting the digitized measurement signals. The controller may be, for example, a PID controller that uses a proportional regulating function, an integrating regulating function and a differentiating regulating function. The PID controller may be suitable for correcting both the fast and the slow fluctuations shown in FIG. 2. Alternatively, other known controllers may also be used. An analog form of transmission may also be provided in addition to digital transmission of the particle beam intensity measured by the detectors. The beam intensity is no longer limited by the scanning point having the lowest dose, and the radiation exposure time may be considerably reduced. The minimum doses per scanning point may be set to much lower levels in the irradiation plan. This enables an improved homogeneity, conformity of the irradiation to be achieved and the radiation exposure time to be reduced. This leads to a significant increase in patient throughput and consequently to a more effective use of the particle therapy system. A change to the setpoint value during the sequential irradiation of the plurality of scanning points may be provided at certain scanning points. A change in the setpoint value may take place during the irradiation of the scanning points associated with an ISO energy layer. The beam intensity is no longer limited by the lowest-dose scanning point in the ISO energy layer, and the radiation exposure time may be considerably reduced. This gives greater flexibility in the irradiation planning, since the minimum doses per scanning point may be set to significantly lower values. In one embodiment, the particle beam is directed to a particle beam output of a treatment station, and the particle beam intensity is measured at the particle beam output of the treatment station. This enables the automatic influencing of the particle beam intensity to be set such that the desired particle beam intensity is provided at the particle beam output of the treatment station. As a result, an influencing of the particle beam intensity may be taken into account when the particle beam is directed to the particle beam output. According to another embodiment, the irradiation particles are generated and accelerated by a synchrotron and/or a linear accelerator. Alternatively, the irradiation particles may also be generated and accelerated by a cyclotron. When the synchrotron and the linear accelerator are used, a knock-out exciter may be used to decouple a fraction of the irradiation particles in order to influence the particle beam intensity. The knock-out exciter is controlled, for example, with the aid of a control unit and applies a radio frequency of adjustable strength to the particle beam in the accelerator. The higher the power, the more particles are decoupled per time unit. In order to obtain a desired variation in particle beam intensity, the output power is adjusted automatically in the control unit of the knock-out exciter, for example, using the PID controller. This eliminates the usual need to set parameters of functions that adjust the time characteristic of the output power of the knock-out exciter for a pilot control operation. The high constancy of the particle beam intensity over time makes the application of the particle beam more precise and more homogeneous. This also leads to simplification of control and monitoring systems for application of the particle beam, as a result of which operation becomes safer and more reliable. With the optimal and constant radiation intensity thus achieved, the frequency of beam interruptions (e.g., interlocks) may be reduced. As a result of the reduced frequency of the beam interruptions, the treatment may be carried out more quickly, and the particle therapy system may be used more efficiently. In one embodiment, the particle beam may be directed to one treatment station of a plurality of treatment stations. The particle beam intensity (e.g., the applied particle beam intensity) is measured at each treatment station of the plurality of treatment stations, and the measured particle beam intensity of the treatment station, to which the particle beam is currently being directed, is automatically used for influencing the particle beam intensity. By measuring the particle beam intensity at each treatment station of the plurality of treatment stations and using the particle beam intensity to influence the particle beam intensity, the desired radiation intensity may be provided at the treatment station at which an irradiation session is currently being performed. According to another embodiment, quality information relating to a particle generation device (e.g., the synchrotron, the linear accelerator or the cyclotron) is determined. The quality information is determined automatically as a function of the measured particle beam intensity and the current influencing of the particle beam intensity. For example, if the number of particles stored in the accelerator changes over time due to technical problems, then in the case of the synchrotron with the linear accelerator, the setting of the fraction of decoupled irradiation particles is correctively adjusted in order to maintain the predefined setpoint value for the particle beam intensity. By suitable logging and evaluation of the setting of the fraction of decoupled irradiation particles, the information relating to the quality of the particle generation device may be derived. Measures for preventive maintenance of the particle generation device may be derived on the basis of the quality information, for example. According to yet another embodiment, a particle therapy system is provided. The particle therapy system includes a particle generation device for generating and accelerating irradiation particles, a beam generating device, a measuring device, a raster scan controller and a particle beam influencing device. The beam generating device (e.g., a knock-out exciter) is configured such that the beam generating device generates a particle beam from at least one portion of the irradiation particles of the particle generation device. The measuring device is operable to measure a particle beam intensity of the particle beam automatically. The particle beam influencing device is configured such that the particle beam influencing device influences and consequently adjusts the particle beam intensity as a function of the measured particle beam intensity and a predefined setpoint value for the particle beam intensity. The raster scan controller is configured for irradiating a plurality of scanning points sequentially with the particle beam in accordance with a predefined irradiation plan. The irradiation plan assigns a setpoint value for the particle beam intensity to each scanning point of the plurality of scanning points. The raster scan controller is coupled to the particle beam influencing device and transmits the setpoint value of a scanning point that is to be irradiated to the particle beam influencing device. The particle beam influencing device uses the setpoint value to adjust the desired particle beam intensity for the scanning point that is to be irradiated. Both small and large irradiation doses may be implemented for different scanning points and provided in short radiation exposure times. As a result, the use of the particle therapy system may be improved and the treatment time for the patient shortened. The particle therapy system may also include a particle beam feeder unit that directs the particle beam to a particle beam output of a treatment station. In one embodiment, the measuring device may be disposed at the particle beam output to automatically measure the particle beam intensity. Using the measuring device, a precise and desired beam intensity may be provided at the particle beam output. The particle therapy system may include a particle beam feeder unit that may direct the particle beam to one treatment station of a plurality of treatment stations. A measuring device for automatically measuring the particle beam intensity of the emitted particle beam (e.g., for detecting the applied particle beam intensity) is provided at each treatment station of the plurality of treatment stations. The particle therapy system may also include a switchover unit that directs the measured particle beam intensity of the one treatment station, to which the particle beam is currently being directed to the particle beam influencing device. This enables the plurality of treatment stations to be supplied with the particle beam, as a result of which the particle therapy system may be used more efficiently. With the aid of the switchover unit, the desired particle beam intensity is provided at each treatment station of the plurality of treatment stations. According to another embodiment, the particle therapy system includes a quality determining device for determining information relating to the quality of the particle generation device. The quality information is determined by the quality determining device as a function of the measured particle beam intensity and the influencing of the particle beam intensity by the particle beam influencing device. This enables malfunctions to be detected and rectified at an early stage or assists in the scheduling of maintenance activities on the particle therapy system in the future. The above-described particle therapy system is suitable for performing the above-described method and therefore includes the advantages described in connection with the method. The present embodiments include a computer program product (e.g., software that may be loaded into a memory (e.g., a non-transitory memory) of a programmable controller) of the particle beam influencing device, for example, of the particle therapy system. All of the above-described embodiments of the method may be implemented using instructions of the computer program product when the computer program product is executed in the particle therapy system. The present embodiments provide an electronically readable data medium (e.g., a CD or DVD), on which electronically readable control information (e.g., software) is stored. When the control information is read from the data medium and stored in a control device (e.g., the particle beam influencing device of the particle therapy system), the embodiments of the above-described method may be performed with the particle therapy system. The present embodiments are explained below with reference to the drawings. The foregoing and following description of the individual features, their advantages and their effects relate both to the device and to the method, without this being explicitly mentioned in detail in each case. The individual features disclosed in the method may also be used in combinations other than those shown. FIG. 3 shows a particle therapy system 100. The particle therapy system 100 includes a particle generation device 101, a beam generating device 102 and a particle beam feeder unit (not shown). Irradiation particles, protons or carbon ions, for example, are generated and accelerated in the particle generation device 101 (e.g., a linear accelerator and a synchrotron or a cyclotron). In the case of the synchrotron and the linear accelerator, a fraction of the generated and accelerated particles is decoupled from the particle generation device 101 with the aid of the beam generating device 102 and directed as a particle beam into the particle beam feeder unit (not shown). The beam generating device 102 may be, for example, a knock-out exciter that applies a radio frequency of adjustable power to the particle beam in the particle generation device 101. The higher the power, the more particles are decoupled per time unit and supplied to the decoupled particle beam. The particle beam feeder unit directs the generated particle beam to one treatment station of a plurality of treatment stations that may be arranged, for example, in different treatment rooms. The particle therapy system also includes a measuring device 103-105 at each treatment station for the plurality of treatment stations for the purpose of measuring a particle beam intensity of the particle beam that is supplied to the one treatment station by the particle beam feeder unit. Three treatment stations are shown by way of example in FIG. 3, the measuring device 103 being disposed at a first treatment station, the measuring device 104 being disposed at a second treatment station, and the measuring device 105 being disposed at a third treatment station. The measuring devices 103-105 measure the beam intensity, for example, with the aid of gas-filled ionization chambers and parallel-plate capacitors arranged in the gas-filled ionization chambers. The acquired particle beam intensity indicates a particle beam dose per time unit. Each measuring device 103-105 is assigned a dose determining device 106-108 that integrates the particle beam intensities of the assigned measuring devices over time in order to determine a particle beam dose at the treatment station. The particle therapy system 100 also includes a switchover unit 109 that is coupled to the measuring devices 103-105. The switchover unit 109 is also coupled to a particle beam influencing device 110. The particle beam influencing device 110 may include, for example, an electronic controller such as a microprocessor. A higher-order controller (not shown) of the particle therapy system 100 that also controls the particle beam feeder unit provides the switchover unit 109 with information indicating which treatment station is currently being supplied with the particle beam by the particle beam feeder unit. The switchover unit 109 selects the measured particle beam intensity of one of the measuring devices 103-105 in order to pass the measured particle beam intensity on to the particle beam influencing device 110. The connection between the measuring devices 100-105 and the switchover unit 109 may transmit, for example, digital signals over separate lines or via a bus system. The higher-order controller provides the particle beam influencing device 110 with a setpoint value (e.g., a predefined setpoint value) for the particle beam intensity, and the particle beam influencing device 110 provides a control value for a drive circuit 111 for the beam generating device 102. The control value for the drive circuit 111 is determined by the particle beam influencing device 110 as a function of the measured particle beam intensity by one of the measuring devices 103-105 and the predefined setpoint value for the particle beam intensity. For this purpose, the particle beam influencing device 110 includes a PID controller, for example. The PID controller provides the control value for the drive circuit 111, using which the radio frequency power of the knock-out exciter is set, for example, in order to regulate the decoupling rate or extraction rate of the particles in the synchrotron to a temporally constant rate that is predefined by the setpoint value. The particle beam influencing device 110 may be implemented, for example, in the form of a standard industrial controller (e.g., a programmable logic controller) and be integrated as a standard module into the particle therapy system. The control value that is transmitted by the particle beam influencing device 110 to the drive circuit 111 of the beam generating device 102 may be transmitted, for example, in analog or digital form. Alternatively, the switchover unit 109, acting independently of the higher-order controller, may also pass on the measured particle beam intensity from the one measuring device 103-105 to the particle beam influencing device 110. The particle beam influencing device 110 is disposed at a treatment station that reports a reference or setpoint intensity different from zero, since for reasons of patient safety, a higher-order system provides in any case that only one treatment station is supplied with a particle beam at any given time. The particle therapy system 100 may also include a quality determining device (not shown) that is integrated, for example, in the higher-order control device of the particle therapy system 100. The quality determining device records the control value that is output by the particle beam influencing device 110 to the drive circuit 111 and logs the control values over the course of time together with the setpoint values for the particle beam intensity. If, for example, the number of particles generated and stored by the particle generation device 101 changes over time due, for example, to technical problems, the control value of the particle beam influencing device 110 will assume anomalous values (e.g., upper or lower limit values) before an end of the planned radiation exposure time is reached. Using suitable logging and evaluation of the time instant at which the limit value is reached, a quality parameter may be derived for the operation of the particle generation device 101. Measures for preventive maintenance of the particle generation device 101 may be derived on the basis of the quality parameter. If the control value that is output by the particle beam influencing device 110 to the drive circuit 111 reaches a limit value, the control value may also signal that there are no more particles in the accelerator of the particle generation device 101. This information may be used in order to perform the regulated application of the particle beam until such time as the drive circuit 111 signaled that there are no more particles in the accelerator by the control value from the particle beam influencing device 110. The number of particles stored in the accelerator may be used to optimum effect. This may contribute toward reducing the load on the environment, since otherwise, the stored and accelerated residual beam would be annihilated, resulting in radioactive contamination of an annihilation unit and/or parts of the system. Since the information relating to the particle beam intensity is output by the measuring devices or detectors 103-105 in realtime to the particle beam influencing device 110 for the purpose of regulating the particle beam intensity, a more precise particle beam intensity may be provided at the treatment stations. FIG. 4 shows the variation over time of the particle beam intensity regulated according to the present embodiments. As in FIG. 2, the particle beam intensity was measured at intervals of 50 μs. The regulation function was activated at time instant t=22.7 s. Compared with FIG. 2, FIG. 4 shows that the long-term fluctuations have been successfully reduced to a significant degree as a result of the regulation function. FIG. 5 shows a comparison of the particle beam intensity with and without regulation function. In FIG. 5, the intensity data has been smoothed using an FIR filter by averaging over 20 ms. As can be seen from FIG. 5, not just the slow fluctuations, but also the short-lived fluctuations have been significantly reduced in the regulated particle beam in the time range from approximately 22.7 s to 27.2 s. In the unregulated intensity curve from approximately 31.4 s to 36.3 s, the significantly stronger short-term fluctuations may also be seen in addition to the considerably stronger slow fluctuations. In a particle beam therapy session, tumors may be treated using a raster scan method. In the raster scan method, the tumor is subdivided logically into a plurality of raster scanning volumes (e.g., scanning points), and an irradiation dose is assigned to each of the scanning points in an irradiation plan. In order to irradiate a scanning point, the particle beam is directed onto the scanning point, and particles having a predetermined energy that determines the penetration depth of the particles are delivered into the tumor. The treatment plan may have widely varying particle doses for each of the scanning points. FIG. 6 shows by way of example desired particle doses for a plurality of scanning points of an irradiation plan. In one embodiment, a minimum time per scanning point is used during the time the irradiation is applied onto the scanning point. The minimum time for the irradiation of the scanning point with a minimum dose limits the maximum beam intensity that may be used. In FIG. 6, a scanning point having a minimum dose is identified by an arrow. In a particle therapy system according to the prior art, in which the particle beam intensity is not controlled as described hereintofore, a relatively long radiation exposure time is accepted for many scanning points with a significantly higher dose, the radiation exposure time being dominated by the dose of the scanning point having the lowest dose. In the particle therapy system 100 shown in FIG. 3, which has a regulated particle beam intensity, the particle beam intensity may be varied dynamically from scanning point to scanning point. For example, the setpoint value for the beam intensity may be derived automatically from the irradiation plan and transmitted in a synchronized manner to the particle beam influencing device 110 in realtime during the irradiation of the different scanning points. Since the accelerator system may have the characteristic of responding to a control variable modification of the controller with a delay time of several milliseconds, the specification of the setpoint value for the particle beam influencing device 110 may be output with a “lead time.” If a chronologically sorted list containing the data of the scanning points (irradiation plan) is present in the higher-order control system prior to the start of the irradiation, the optimal time instant for passing the setpoint value to the particle beam influencing device 110 may be determined in a preprocessing stage using a suitable algorithm. The beam intensity may be optimally adapted to the requirements of the respective scanning point at all times. FIG. 7 shows a change in the radiation intensity as a result of a change to a predefined setpoint value. An advantage of the present embodiments derives from the fact that the beam intensity is no longer limited by the scanning point having the lowest dose. By varying the particle beam intensity, scanning points, for which a high beam dose is provided with the designated dose in a considerably shorter time, may be irradiated. The total radiation exposure time may be considerably reduced. The minimum doses per scanning point may be set to a much lower level in the therapy planning systems. An improvement in the homogeneity and conformity of the irradiation is thus achieved. A long radiation exposure time is avoided owing to the active regulation of the particle stream intensity. Thus, the patient throughput of the particle therapy system is significantly increased and the quality of the treatment plans improved. While the present invention has been described above by reference to various embodiments, it should be understood that many changes and modifications can be made to the described embodiments. It is therefore intended that the foregoing description be regarded as illustrative rather than limiting, and that it be understood that all equivalents and/or combinations of embodiments are intended to be included in this description. |
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051026139 | claims | 1. A brake assembly for a control rod drive for selectively preventing travel of a control rod in a nuclear reactor vessel comprising: a shaft having a longitudinal centerline axis; means for translating said control rod upon rotation of said shaft; means for selectively rotating said shaft in a first direction and in a second direction, opposite to said first direction; a stationary housing having a central aperture receiving said shaft; a frame fixedly joined to said housing and having a guide hole; a rotor disc fixedly connected to said shaft for rotation therewith and having at least one rotor tooth extending radially outwardly from a perimeter thereof, said rotor tooth having a locking surface and an inclined surface extending therefrom in a circumferential direction; a brake member disposed adjacent to said rotor disc perimeter and including a base, at least one braking tooth having a locking surface extending radially inwardly from said base and an inclined surface extending therefrom in a circumferential direction, and a plunger extending radially outwardly from said base and slidably joined to said frame through said guide hole; said rotor tooth and said braking tooth being complementary to each other; and means for selectively positioning said brake member in a deployed position abutting said rotor disc perimeter for allowing said braking tooth locking surface to contact said rotor tooth locking surface for preventing rotation of said shaft in said first direction, and in a retracted position spaced radially away from said rotor disc for allowing said rotor disc and said shaft to rotate without restraint from said brake member, said positioning means including a tubular solenoid fixedly joined to said frame and having a central bore disposed around said brake member plunger and effective for sliding said brake member plunger relative to said frame for positioning said brake member in said deployed and retracted positions. a spring disposed in said brake member plunger and being initially compressed for engaging said braking tooth against said rotor tooth in said brake member deployed position when said solenoid is deenergized; and said solenoid being energizable for electromagnetically drawing said plunger further into said solenoid bore and further compressing said spring for positioning said brake member in said retracted position. 2. A brake assembly according to claim 1 wherein said positioning means in effective for resiliently supporting said braking tooth for allowing said rotor tooth inclined surface to displace said braking tooth inclined surface to intermittently free said rotor tooth locking surface from said braking tooth locking surface for allowing said shaft to rotate in said second direction when said brake member is in said deployed position. 3. A brake assembly according to claim 2 wherein said rotor disc includes a plurality of said rotor teeth spaced circumferentially around said rotor disc perimeter. 4. A brake assembly according to claim 3 wherein said brake member base is arcuate and includes a plurality of said braking teeth spaced circumferentially thereon. 5. A brake assembly according to claim 4 further comprising two of said brake members including first and second, circumferentially spaced brake members, and said positioning means is effective for positioning both said first and second brake members in said deployed and retracted positions. 6. A brake assembly according to claim 5 wherein said first brake member is spaced about 180.degree. from said second brake member. 7. A brake assembly according to claim 4 wherein said rotor tooth locking and inclined surfaces form an obtuse angle therebetween. 8. A brake assembly according to claim 4 wherein said plunger is hollow and said positioning means further includes: 9. A brake assembly according to claim 8 wherein said plunger is disposed perpendicularly to said shaft longitudinal centerline axis. 10. A brake assembly according to claim 9 wherein said rotor tooth locking and inclined surfaces form an obtuse angle therebetween. 11. A brake assembly according to claim 9 wherein said rotor tooth locking surface is disposed at an angle of 90.degree. relative to said rotor tooth inclined surface. 12. A brake assembly according to claim 8 wherein said plunger and said guide hole have complementary, square cross-sections. |
061730289 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention will be described in detail in conjunction with what is presently considered as preferred or typical embodiments thereof by reference to the drawings. In the following description, like reference characters designate like or corresponding parts throughout the several views. Also in the following description, it is to be understood that such terms as "right", "left", "top", "bottom" and the like are words of convenience and are not to be construed as limiting terms. FIG. 1 is a partially enlarged vertical sectional view showing in detail a structure of a upper plenum and associated components in a pressurized water reactor in which short-length tubes 5 are employed as the heated fluid guide members of short length according to an embodiment of the present invention. In the figure, the structural or component members of the pressurized water reactor are, for the most part, essentially the same as those of the conventional reactor described hereinbefore. Accordingly, repetitive description thereof will be unnecessary. In this conjunction, it should first be mentioned that the short-length tubes 5 serving as the heated fluid guide member according to the present invention can be installed additionally as fresh members or components at all available locations in an outer peripheral region within the upper plenum 40 which are not occupied by any existing internal component members. Alternatively, the short-length tubes 5 may be disposed restrictively only at locations in an outlet or exit region adjacent to the outlet nozzle 12. The short-length tubes 5 according to an illustrated embodiment of the present invention are formed in a sleeve-like or tubular shape and have a top end and a bottom end both of which are opened. Thus, when the short-length tubes 5 are mounted on the upper core plate 21, the tubes 5 are brought into fluid communication with the reactor core by way of through-holes formed in the upper core plate 21 so that the coolant leaving the reactor core can flow into the short-length tubes 5 to flow therethrough, being ultimately ejected from the top open ends of the short-length tubes 5 into the upper plenum 40. At this juncture, it is to be mentioned that the overall length of the short-length tubes 5 should preferably be so selected that when they are moved on the upper core plate 21, the open top end of the short-length tubes 5 assume a height level or vertical position which is lower than the bottom of the bore of the outlet nozzle 12. More preferably, the length of the short-length tubes 5 should be so determined that the open top ends thereof are positioned substantially midway between the upper surface of the upper core plate 21 and the lower end of the outlet nozzle 12 in the mounted state. When the short-length tubes 5 have an excessively long length, the flow resistance to the coolant stream flowing through the upper plenum 40 toward the outlet nozzle 12 from the center portion of the core will increase to thereby exert undesirable influence on the strength of the structural members such as the upper core support columns 23 and the control rod cluster guide tubes 22. The short-length tubes 5 of the length determined as mentioned above can thus be mounted at all available locations on the upper core plate 21 substantially along the outer periphery of the core which are not occupied by existing structural members such as the upper core support columns, the control rod cluster guide tubes and others. Furthermore, with the length of the short-length tubes 5 mentioned above, the low temperature peripheral coolant stream can be positively discharged or introduced into the high temperature center coolant stream without fail under the effect of inertia of the coolant ejected from the short-length tubes 5. Furthermore, it should also be mentioned that the short-length tubes 5 can be installed within the upper plenum 40 with ease because of the short length thereof. Of course, the short-length tubes 5 can be installed in existing equipment without any appreciable accompanying difficulty. Referring to FIG. 1, it can be seen that a slot-formed tube 1 is mounted adjacent to a short-length tube 5. The structure of the slot-formed tube 1 is disclosed in detail in Japanese Patent Application No. 10-284532 entitled "APPARATUS FOR PROMOTING INTERMIXING OF HEATED FLUID STREAMS IN A NUCLEAR REACTOR" filed by the inventor(s) of the present application. Parenthetically, the disclosure of this preceding application is incorporated herein by reference. Described briefly, the slot-formed tube 1 has an open bottom end so that when it is mounted on the upper core plate 21, the slot-formed tube 1 is brought into fluid communication with the reactor core by way of through-holes formed in the upper core plate 21. Further, a plurality of slots 2 are formed in the slot-formed tube 1 in a window-like fashion at locations corresponding to the height level (vertical position) of the outlet nozzle 12 installed in the nuclear reactor vessel 10. Parenthetically, it should also be mentioned that the slot-formed tube 1 has a reduced diameter portion reduced at a ratio of about 20% of the diameter substantially over two thirds of the whole length thereof in order to prevent excessive hydrodynamic load from being applied to the structural members disposed within the upper plenum 40. Now, description will turn to disposition of the short-length tubes 5 within the upper plenum 40 by reference to FIG. 2. FIG. 2 is a plan view of a quarter cross section of the nuclear reactor vessel 10 and shows only major structural members disposed within the upper plenum 40. In this case, the slot-formed tube 1 mentioned above is also employed in combination with the short-length tube 5 according to the invention. In FIG. 2, each of the control rod cluster guide tubes 22 is shown in the form of a plain rectangle, the upper core support columns 23 are represented by X-like patterns, respectively, and the slot-formed tubes 1 are indicated by hatched circles, respectively. Further, the positions at which the short-length tubes 5 according to the present invention are mounted are indicated by plain circles, respectively. Next, description will be made of the flow behaviors of coolant or light water within the core in the nuclear reactor vessel 10 having the upper plenum 40 within which the short-length tubes 5 with the structure described above are installed. In this conjunction, it is noted that the flow behaviors of the light water is, for the most part, similar to that in the conventional reactor described hereinbefore. Accordingly, the following description will be directed to the flows or streams of light water within the upper plenum 40 and the outlet nozzle 12 by reference to FIG. 1. As described hereinbefore, the stream (indicated by an arrow d) of the coolant or light water flowing through the center portion of the reactor core where the nuclear fission reaction is vigorous is heated at a relatively high rate up to a relatively high temperature. Then, the heated coolant leaves the reactor core into the upper plenum 40 and flows along and through the control rod cluster guide tubes 22. Thus, the coolant reaches the lower surface of the upper core support plate 20 to be thereby deflected in a transverse direction toward the outlet nozzle 12, as indicated by the arrows e and f in FIG. 2. On the other hand, the peripheral stream a of the coolant flowing through the peripheral portion of the reactor core where the neutron flux density is relatively low is heated to a temperature which is relatively low when compared with the center stream of the coolant. Thus, the coolant passed through the peripheral portion of the reactor core enters the upper plenum 40 from the reactor core at a relatively low temperature and flows into the short-length tubes 5 disposed appropriately to be finally ejected therefrom upwardly into the upper plenum 40 through the open top ends of tubes 5. Although the open top end is positioned at a height lower than the bottom end of the outlet nozzle 12, the coolant can be ejected upwardly from the top end of the short-length tubes 5 under the effect of inertia of the coolant flowing therethrough and can flow toward the outlet nozzle 12 within the upper plenum 40, as indicated by arrow b'. As will now be appreciated, owing to the arrangement that at least some part of the coolant of relatively low temperature is ejected upwardly through the short-length tubes 5 in the manner described above, the peripheral coolant stream of relatively low temperature and the center coolant stream of relatively high temperature are intermingled or intermixed to a sufficient extent, whereon the intermixed coolant flows into the outlet nozzle 12, as indicated generally by an arrow A. As a result of this, temperature distribution of the coolant (i.e., light water) flowing through the outlet nozzle 12 can be made uniform. Thus, the mean temperature of the coolant can be measured with high accuracy and reliability. Furthermore, the short-length tubes 5 allow the center coolant stream of high temperature to flow smoothly above the short-length tubes 5 toward the outlet nozzle 12 because of the short length of the tubes 5. Thus, the hydrodynamic load applied to the structural member disposed in the vicinity of the outlet nozzle 12 can be reduced. Furthermore, intermixing of the peripheral coolant stream of low temperature and the center coolant stream of high temperature can be promoted further. Additionally, when the short-length tube 5 according to the present invention is employed in combination with the slot-formed tube 1, there can be obtained advantageous effects provided by the slot-formed tube 1. More specifically, some part of the coolant of relatively low temperature flows into the upper plenum 40 from the slot 2 formed in the slot-formed tube 1 at a location corresponding to the height of the outlet nozzle 12. Thus, the peripheral coolant stream of relatively low temperature and the center coolant steam of relatively high temperature can be intermixed appropriately, whereon the intermixed coolant flows into the outlet nozzle 12, as indicated generally by the arrow A. As a result of this, the temperature distribution of the coolant (i.e., light water) flowing through the outlet nozzle 12 is made more uniform. Thus, the coolant intermixed more sufficiently can flow through the outlet pipe 42 to ensure measurement of the mean temperature of the coolant with enhanced accuracy. In the foregoing, exemplary embodiments of the present invention which are considered preferable at present and other alternative embodiments have been described in detail by reference to the drawings. It should, however, be noted that the present invention is never restricted to these embodiments but other numerous variations and modifications of the structure for promoting intermixing of the heated fluid streams can be easily conceived and realized by those skilled in the art without departing from the spirit and scope of the present invention. |
047986999 | abstract | A nuclear reactor includes a plurality of upstanding guide thimbles and a plurality of control rods received in the guide thimbles and supported for movement relative to the thimbles between inserted and withdrawn positions. The control rods each include a tubular cladding member and an end plug attached to a lower end of the member. The improvement relates to an wear sleeve disposed on the end plug so as to provide a contact interface between the control rod and its respective guide thimble. The sleeve is composed of material similar to that of the guide thimble and attached to the end plug by an interlock connection. The interlock connection includes a circumferential groove formed in the end plug and a circumferential protuberance swaged or mechanically roll formed in the sleeve so as to extend into the end plug groove. |
summary | ||
045338329 | abstract | A self-supporting modular radiation attenuation system formed from a plurality of modules stacked on one another in any desired alignment to protect workers from radiation exposure. The modules include a skin assembled with a radiation attenuation medium therein and shaped to mate with one another when assembled. The medium can be lead particles or compressed lead wool. The outer module skin is sufficiently dimensionally stable to support stacking; however, flexible enough to allow the modules to conform to irregular surfaces. The modules can include a particle binder and the system can include framing to support the modules. |
abstract | An energy beam is irradiated onto a target from an energy beam source, thereby generating an X-ray with an irradiating area to be irradiated onto an object. Then, the X-ray is introduced into a spectrometer, thereby generating an X-ray with parallelism through the selection of wavelength and wavelength range. |
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048428094 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to the drawings, it can be seen in FIG. 1 that the invention is generally indicated by the numeral 10. Rod arraying system 10 is generally comprised of a consolidation canister 12, a rigid support structure 14, and a rod arraying device 16. Consolidation canister 12 is essentially a standard storage and transfer canister for fuel rods and burnable poison rods which has been modified by the addition of several horizontal holes or windows 18 spaced vertically along one side. A portion of rod arraying device 16 enters consolidation canister 12 through each window 18 for alignment of fuel rods 20 as they are inserted into canister 12. As seen in FIG. 11, rod arraying device 16 positions fuel rods 20 in a triangular pitch configuration inside canister 12 to achieve the desired 2 to 1 consolidation ratio. Consolidation canister 12 is supported in a vertical position for receiving fuel rods by rigid support structure 14. Rigid support structure 14 is generally a framework structure sized to receive and support consolidation canister 12. Support structure 14 also has rod arraying device 16 mounted thereon as seen in FIG. 1. Rod arraying device 16 is generally comprised of a mounting base 24 and guide assembly 26. In the preferred embodiment, guide assembly 26 is attached to mounting base 24, as shown in FIG. 5 and 9. Mounting base 24 is guided on support structure 14 by way of wheels 11 and rails 13. The wheels 11, for which there are six, are mounted to the mounting base 24 by way of two mounting blocks 15, one on each side of the mounting base 24. The wheels 11 ride on rails 13 which are mounted to the support structure 14 on an incline, such that the guide assembly 26 and mounting base 24 is predisposed toward the inside of canister 12. In the preferred embodiment, the rod arraying device 16 exerts a force on rods 20, proportional to the angle of inclination and the weight of said device. Retraction of the inclined, rod arraying device 16 from canister 12 for removal of said device or for the purpose of installing or removing canister 12 is accomplished in one of two ways. Cylinder 17, seen in FIG. 5, 9 is attached to support structure 14 and the rod of cylinder 17 bears against mounting base 24 such that extension of said rod pushes rod arraying device 16 out of canister 12. Retraction of rod arraying device 16 is maintained by keeping pressure on cylinder 17. Latch 19 is provided as a redundant feature. Latch 19 is pivotally attached to mounting block 15. Roller 21 of latch 19 rides on rail 13. Upon withdrawal of inclined, rod arraying device 16 from canister 12, latch 19 catches on the upper end of rail 13, and holds rod arraying device 16 in the retracted position. Latch 19 is disengaged by operating the release mechanism 23 mounted on support structure 14, or by a hook on the end of a remotely operated pole (not shown). For exclusive use of cylinder 17, latch 19 can be made inoperative by rotating said latch back against mounting block 15. Guide assembly 26 is comprised of two guide plates 38, 40 as best seen in FIGS. 2, 4, and 7. As seen in FIG. 7, guide plates 38, 40 have identical front scalloped edges which are sized to fit nuclear fuel rods and burnable poison rods. The plates are positioned one over the other in a front-to-back and side-to-side offset configuration such that the projections on the front edges of the plates alternate between the two plates. Lower guide plate 40 is slidably mounted to upper guide plate 38 by means of bearing guides 42 to permit forward and backward cycling of lower guide plate 40 during operation. Bearing guides 42 are mounted to the underside of upper guide plate 38. Upper guide plate 38 is slidably mounted on mounting base 24 by means of guide bars 30. Means for causing forward and backward cycling of lower guide plate 40 relative to upper guide plate 38 is provided in the form of air cylinder 44. As seen in FIG. 2 air cylinder 44 is mounted to a block 46 which extends from the bottom of upper guide plate 38. Cylinder rod 48 extends forward of air cylinder 44 and is attached to plate 50 which extends down from lower guide plate 40. In this manner, cycling of air cylinder 44 causes cylinder rod 48 and lower guide plate 40 to move forward and backward relative to upper guide plate 38. The range of relative motion between the two guide plates is controlled by the cylinder stroke, or by adjustable stops 52, 54 seen in FIG. 7. Forward stops 52 are threadably engaged in and extend forward through lower guide plate 40 to control forward travel by making contact with stop bar 56. Rearward stops 54 are slidably engaged in stop bar 56 and extend rearwardly thereof in threaded engagement with lower guide plate 40 to control rearward travel of lower guide plate 40. In an alternate embodiment of the invention rod arraying device 16 may be positioned in a horizontal orientation relative to canister 12. Mounting base 24 is provided with bearing blocks 28 with bores therethrough sized to slidably receive guide bars 30. As best seen in FIGS. 3 and 4 guide bars 30 are mounted on the top of guide assembly 26 and extend rearwardly therefrom. Bushings may also be mounted in the bores provided to reduce friction and wear between bearing blocks 28 and guide bars 30. For ease of illustration, the structural differences of the alternate embodiment are shown as additional structure in the drawings of the preferred embodiment. A cross bar 32 is attached to and bridges the end of guide bars 30 farthest from guide assembly 26. An attachment point is provided for a spring in the alternate embodiment as illustrated in FIG. 3. A first fixed spring anchor 34 is provided near the forward end of mounting base 24 and may be attached to bearing block 28. A second spring anchor 35 is attached to the cross bar 32. Spring 36 is attached between the two anchor points. Since guide bars 30 are attached to guide assembly 26, the spring and anchor combination serves as a means for biasing guide assembly 26 forward away from mounting base 24 and toward and into consolidation canister 12. The mounting base 24 is prohibited from rolling by way of the two-position locking mechanism 62. In the alternate embodiment means for locking guide assembly 26 in a retracted position to allow removal and installation of consolidation canister 12 is provided in the form of two-position locking mechanism 62 seen in FIG. 6 and 8. The two-position locking mechanism 62 is actuated by rotating a nut 31 connected to cam 33 by way of shaft 37. Rotation of cam 33 provides push-pull motion to locking shafts 39. Cam 33 is provided with slots to which locking shafts 39 are pivotally attached. Locking shafts 39 are slidably engaged to mounting base 24 by way of guide blocks 41. In addition to guide blocks 41, the locking shafts 39 are slidably engaged in holes in mounting base 24. When the locking mechanism 62 is actuated, the locking shafts 39 engage holes 43 in the rails 13 as shown in FIG. 3 and 8. There are two sets of holes in rails 13. One set is for locking the rod arraying device 16 in the outermost position, out of canister 12 for canister installation and removal operations. The second set of holes is used to lock the rod arraying device 16 in place during canister loading. Nut 31 is biased upward with spring 45, so that in the upward position the nut 31 is prevented from turning by pin 47 which is fixed in nut 31 and protrudes from nut 31 into a slot in locking plate 49. Nut 31 is rotated by a socket on the end of a remotely operated pole (not shown). When a small, downward force is applied to nut 31, pin 47 disengages from locking plate 49 and nut 31 is free to rotate. Vertical travel of nut 31 and shaft 37 is limited by the length of the slot in shaft 37 and vertical limit pin 51 which is fixed in guide post 53. In the alternate embodiment means for locking out the spring tension between guide assembly 26 and mounting base 24 is provided in the form of guide extension lock 64 seen in FIG. 10. In the case of remote removal of rod arraying device 16 (as when in need of repair), guide extension lock 64 is used. Guide extension lock 64 provides a way to lock the guide assembly 26 any place in its travel, negating the effect of spring 36, and facilitating subsequent unlatching of the two-position locking mechanism 62. Before remote installation of rod arraying device 16 (such as after repair), guide extension lock 64 is locked so that the extension of guide assembly 26 falls short of the last row of rods 20 when the rod arraying device 16 is latched in the locked-in position (by operating the two-position, locking mechanism 62). After the rod arraying device 16 is latched in the locked-in position, the guide extension lock 64 is released so that the guide assembly 26 can approach and rest against rods 20. Guide extension lock 64 is actuated by rotating nut 55 which is fixed to threaded stud 57. Threaded stud 57 is pivotally attached to clamping arm 59 which is pivotally attached to rear bearing block 28. Rear bearing block 28 is provided with a clearance hole for guide rod 61. Clamping arm 59 is provided with a notch to facilitate clamping of guide rod 61. Guide rod 61 is attached to movable spring anchor 32. Nut 55 is rotated by a socket on the end of a remotely operated tool (not shown). Nut 55 is captured by cap 63 but is free to rotate. Means for dampening the movement of guide assembly 24 is provided in the form of dashpot 60 seen in FIG. 3. After remote installation of rod arraying device 16, and after latching rod arraying device 16 in the locked-in position, guide extension lock 64 is released. Upon release, dashpot 60 works to slow the advancement of guide assembly 26 into canister 12. Dashpot 60 is a cylinder which draws in and expels pool water through orifices mounted in the cylinder ports. Dashpot 60 is trunnion mounted to front bearing block 28 in pivot blocks 25. Cylinder rod 27 is attached to rod mounting block 29 which is attached to upper guide plate 38. During operation, consolidation canister 12 is first placed in position. Before installation of canister 12, each rod arraying device 16 is placed in the locked-out position using either cylinder 17 or two-position locking mechanism 62. With rod arraying device 16 in position on support structure 14, guide assembly 26 is inserted into canister 12 through window 18. Lower guide plate 40 is retracted relative to upper guide plate 38 at the start so that upper guide plate 38 bears against the inside wall of canister 12. It is seen in FIG. 1 that a plurality of rod arraying devices 16 are used in the preferred embodiment along the longitudinal axis of canister 12. Fuel rods 20 are loaded one at a time into canister 12 through the scallop spaces in upper guide plate 38. The loading pattern of the rods provides gaps between adjacent rods in each row which allow the projections (front scalloped edges) to hold each rod in position. After the first row is loaded guide assembly 26 is cycled to cause lower guide plate 40 to extend forward beyond upper guide plate 38. This places the projections on lower guide plate 40 against the first row of fuel rods 20, holds the first row against the inside surface of canister 12, and provides room for loading the second row through the scallop spaces on lower guide plate 40. Once the second row is loaded, lower guide plate 40 is retracted. Pressure from the weight of inclined rod arraying device 16 then biases the projections of upper guide plate 38 against the second row of fuel rods 20. The provides room for loading the third row through the scallop spaces in upper guide plate 38. It can be seen that this is essentially a repetition of the position of the device used in loading the first row. The process is then repeated until canister 12 is full or the desired consolidation ratio is achieved. Because many varying and differing embodiments may be made within the scope of the inventive concept herein taught and because many modifications may be made in the embodiment herein detailed in accordance with the descriptive requirement of the law, it is to be understood that the details herein are to be interpreted as illustrative and not in a limiting sense. |
description | This application is a divisional application of U.S. application Ser. No. 12/574,219 filed Oct. 6, 2009 now U.S. Pat. No. 7,997,078, which is a divisional application of U.S. application Ser. No. 11/340,643, filed Jan. 27, 2006, now U.S. Pat. No. 7,614,233, the entirety of which are incorporated herein by reference. The present application claims priority from Japanese Patent Application No. 2005-021835, filed Jan. 28, 2005 and 2005-066498, filed Mar. 10, 2005, the contents of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to an operation method of a nuclear power plant and to a nuclear power plant. More particularly, the present invention relates to an operation method of a nuclear power plant and to a nuclear power plant, which are suitable for an uprate of normal operation power output of a nuclear power plant. 2. Description of the Related Art In a newly constructed nuclear power plant, electric power has hitherto been uprated, for example, by improving the fuel makeup or the shape and makeup of a fuel assembly so as to increase the flow rate of main steam at a core outlet. Such related art is disclosed in Patent Document 1 (JP,A 9-264983). When the above-mentioned related art is applied to the existing nuclear power plant, the rate of core flow passing through a reactor core is substantially the same as that before the power uprate, while thermal power of the core is increased. In a boiling water reactor (BWR), therefore, an average void rate (proportion of steam with respect to the channel volume) in the core is increased. Accordingly, the flow speed of a coolant is increased and so is a pressure loss in the core. Also, with an increase in the amount of steam generated in the core, a pressure loss in a water-steam two-phase flow section is increased and a margin of core safety tends to reduce. Further, an increase of the average void rate in the core increases an amount of steam condensed in the so-called pressure transient state where pressure rises, for example, when a generator load is cut off, thus resulting in a larger amount of decrease of the average void rate in the core. Generally, the boiling water reactor has a negative void feedback coefficient so that the reactor power reduces as the void rate increases. In the pressure transient state, however, the average void rate in the core is reduced and the reactor power is increased. Thus, the related art has a possibility that, after the power uprate, the amount of decrease of the average void rate in the core is increased in the pressure transient state and a design margin for pressure transient events is reduced. Meanwhile, the flow rate of main steam is increased substantially in proportion to an increase of the power uprate. The increased flow rate of main steam reduces design margins of almost all equipment, such as feedwater equipment including feedwater piping, a feedwater heater, a feedwater pump, etc., pressure vessel internals including a dryer, etc., a main steam line, a high pressure turbine, a low pressure turbine, and a condenser. In a usual nuclear power plant employing a boiling water reactor, the high pressure turbine is one of the equipment with a possibility that the design margin is first lost with the increase in the flow rate of main steam. Also, in a nuclear power system other than the boiling water reactor, a similar problem arises in a plant where the design margin for the high pressure turbine is relatively small. Accordingly, when the related art is applied to the existing nuclear power plant, large-scaled improvement and replacement of plant equipment are required. The increase in the flow rate of main steam can be suppressed by lowering the feedwater temperature. However, such a solution is not realistic for the reason that, if the flow rate of steam extracted for heating the feedwater is simply reduced as a whole, thermal efficiency is noticeably deteriorated and the electric power is not increased in proportion to the core thermal power output. An object of the present invention is to provide an operation method of a nuclear power plant, which can uprate plant power without greatly modifying the construction of plant equipment, while keeping a core's pressure loss characteristic, a safety margin, and a design margin in the transient state substantially the same as those before the power uprate. To achieve the above object, according to one aspect of the present invention, assuming that one operation cycle is defined as a period from a time at which a nuclear power plant starts operation to a time at which the nuclear power plant stops the operation for fuel exchange, second reactor thermal power in a second operation cycle of a nuclear reactor is uprated from first reactor thermal power in a first operation cycle preceding the second operation cycle at least one operation cycle, and a proportion of steam extracted from a steam system and introduced to a feedwater heater, which is in particular extracted from an intermediate point of a high pressure turbine and an outlet thereof (practically, some point in a section ranging from the outlet of the high pressure turbine to an inlet of one of a moisture separator, a moisture separator and heater, and a moisture separator and reheater), with respect to a flow rate of main steam is reduced in the second operation cycle from a proportion in the first operation cycle such that temperature of feedwater discharged from the feedwater heater lowers in the range of 1° C. to 40° C. in the second operation cycle. To achieve the above object, according to another aspect of the present invention, second reactor thermal power in a second operation cycle of a nuclear reactor is uprated from first reactor thermal power in a first operation cycle preceding the second operation cycle at least one operation cycle, and a mass flow rate of steam extracted from a steam system and introduced to a feedwater heater, which is in particular extracted from an intermediate point and an outlet of a high pressure turbine, is reduced in the second operation cycle from a mass flow rate of steam extracted in the first operation cycle such that temperature of feedwater discharged from the feedwater heater lowers in the range of 1° C. to 40° C. in the second operation cycle. To achieve the above object, according to still another aspect of the present invention, second reactor thermal power in a second operation cycle of a nuclear reactor is uprated from first reactor thermal power in a first operation cycle preceding the second operation cycle at least one operation cycle, and a temperature rise in one of a plurality of feedwater heaters, particularly a high pressure feedwater heater installed downstream of a main feedwater pump, is reduced in the second operation cycle such that temperature of feedwater discharged from the feedwater heater lowers in the range of 1° C. to 40° C. in the second operation cycle. To achieve the above object, according to still another aspect of the present invention, second reactor thermal power in a second operation cycle of a nuclear reactor is uprated from first reactor thermal power in a first operation cycle preceding the second operation cycle at least one operation cycle, and at least one of extraction lines for extracting steam from a steam system and introducing the extracted steam to a feedwater heater, which is in particular extended from an intermediate point and an outlet of a high pressure turbine, is shut off such that temperature of feedwater discharged from the feedwater heater lowers in the range of 1° C. to 40° C. in the second operation cycle. To achieve the above object, according to another aspect of the present invention, second reactor thermal power (Q2) in a second operation cycle of a nuclear reactor is uprated A % from first reactor thermal power (Q1) in a first operation cycle preceding the second operation cycle at least one operation cycle, and a proportion of steam extracted from a steam system and introduced to a feedwater heater, which is in particular extracted from an intermediate point of a high pressure turbine and an outlet thereof (practically, some point in a section ranging from the outlet of the high pressure turbine to an inlet of one of a moisture separator, a moisture separator and heater, and a moisture separator and reheater), with respect to a flow rate of main steam in the second operation cycle is kept equivalent to or reduced from a proportion in the first operation cycle such that the following formulae are satisfied;0<A≦5, andT2≦T1−7.7×(Q2−Q1)/(4.5×W)where temperature of the feedwater discharged from the feedwater heater in the first operation cycle is T1 (° C.), temperature of the feedwater discharged from the feedwater heater in the second operation cycle is T2 (° C.), and a core flow rate of the feedwater flowing into the nuclear reactor in the second operation cycle is W (kg/s). To achieve the above object, according to another aspect of the present invention, second reactor thermal power (Q2) in a second operation cycle of a nuclear reactor is uprated A % from first reactor thermal power (Q1) in a first operation cycle preceding the second operation cycle at least one operation cycle, and a proportion of steam extracted from a steam system and introduced to a feedwater heater, which is in particular extracted from an intermediate point of a high pressure turbine and an outlet thereof (practically, some point in a section ranging from the outlet of the high pressure turbine to an inlet of one of a moisture separator, a moisture separator and heater, and a moisture separator and reheater), with respect to a flow rate of main steam in the second operation cycle is kept equivalent to or reduced from a proportion in the first operation cycle such that the following formulae are satisfied;5<A≦10, andT1−40≦T2≦T1−7.7×(Q2×(A+95)/100−Q1)/(4.5×W)where temperature of the feedwater discharged from the feedwater heater in the first operation cycle is T1 (° C.), temperature of the feedwater discharged from the feedwater heater in the second operation cycle is T2 (° C.), and a core flow rate of the feedwater flowing into the nuclear reactor in the second operation cycle is W (kg/s). To achieve the above object, according to another aspect of the present invention, second reactor thermal power (Q2) in a second operation cycle of a nuclear reactor is uprated A % from first reactor thermal power (Q1) in a first operation cycle preceding the second operation cycle at least one operation cycle, and a proportion of steam extracted from a steam system and introduced to a feedwater heater, which is in particular extracted from an intermediate point of a high pressure turbine and an outlet thereof (practically, some point in a section ranging from the outlet of the high pressure turbine to an inlet of one of a moisture separator, a moisture separator and heater, and a moisture separator and reheater), with respect to a flow rate of main steam in the second operation cycle is kept equivalent to or reduced from a proportion in the first operation cycle such that the following formulae are satisfied;10<A<30, andT2≦T1−7.7×(Q2×(A+90)/100−Q1)/(4.5×W)where temperature of the feedwater discharged from the feedwater heater in the first operation cycle is T1 (° C.), temperature of the feedwater discharged from the feedwater heater in the second operation cycle is T2 (° C.), and a core flow rate of the feedwater flowing into the nuclear reactor in the second operation cycle is W (kg/s). To achieve the above object, according to still another aspect of the present invention, a nuclear power plant comprises an extracted flow control valve disposed in at least one extraction line; a temperature sensor disposed in a feedwater system at a point between adjacent two of a plurality of feedwater heaters disposed in the feedwater system or at a point downstream of one of the plurality of feedwater heaters which is positioned most downstream; and an extracted flow controller for outputting an opening request command for the extracted flow control valve based on a measured value from the temperature sensor and a set value of feedwater temperature, wherein the nuclear power plant is operated such that second nuclear thermal power in a second operation cycle of the nuclear reactor is uprated from first nuclear thermal power in a first operation cycle before the second operation cycle, and second feedwater temperature in the second operation cycle is made lower than first feedwater temperature in the first operation cycle. Since the feedwater temperature can be adjusted to a set value through control of the opening of the extracted flow control valve, it is possible to suppress variations in the amount of power generated during the power uprate operation of the nuclear power plant. To achieve the above object, according to still another aspect of the present invention, a nuclear power plant comprises an extracted flow control valve and an extraction flow rate measuring meter disposed in at least one extraction line; and an extracted flow controller for outputting an opening request command for the extracted flow control valve based on a measured value from the extraction flow rate measuring meter and a set value of a flow rate of the extracted steam, wherein the nuclear power plant is operated such that second nuclear thermal power in a second operation cycle of the nuclear reactor is uprated from first nuclear thermal power in a first operation cycle before the second operation cycle, and second feedwater temperature in the second operation cycle is made lower than first feedwater temperature in the first operation cycle. To achieve the above object, according to still another aspect of the present invention, a nuclear power plant comprises an extracted flow control valve disposed in at least one extraction line; at least one main steam flow rate measuring meter disposed in a steam system between the nuclear reactor and the high pressure turbine; and an extracted flow controller for outputting an opening request command for the extracted flow control valve based on a measured value from the main steam flow rate measuring meter and a set value of a flow rate of the main steam, wherein the nuclear power plant is operated such that second nuclear thermal power in a second operation cycle of the nuclear reactor is uprated from first nuclear thermal power in a first operation cycle before the second operation cycle, and second feedwater temperature in the second operation cycle is made lower than first feedwater temperature in the first operation cycle. According to the present invention, in trying to uprate power of the existing nuclear power plant, the power uprate of the nuclear power plant can be realized without greatly modifying the construction of the nuclear power plant, while keeping a core's pressure loss characteristic, a safety margin, a thermal margin, and a design margin in the transient state substantially the same as those before the power uprate. A first embodiment represents the case where the present invention is applied to a boiling water reactor system as one of nuclear power plants. The overall construction of the boiling water reactor system according to this embodiment will be first described with reference to FIG. 1. In FIG. 1, reference numeral 1 denotes a reactor pressure vessel. Recirculation pumps and jet pumps are installed outside and inside the reactor pressure vessel 1 to regulate the rate of flow passing through a core (i.e., a core flow rate). The reactor pressure vessel 1 and its internals constitute a reactor 21, and steam generated in the reactor 21 is supplied to a steam system 22. The steam system 22 comprises a main steam line 2, a high pressure turbine 3 and a low pressure turbine 5 connected to the main steam line 2 in series, and a moisture separator 4 disposed between the high pressure turbine 3 and the low pressure turbine 5. A section including the high pressure turbine 3, which extends from a reactor outlet to an inlet of the low pressure turbine 5, constitutes a high pressure steam system 22A, and a section extending from the inlet of the low pressure turbine 5 to an inlet of a condenser 6 constitutes a low pressure steam system 22B. The condenser 6 condenses steam discharged from the low pressure turbine 5. The condensate condensed by the condenser 6 is supplied as feedwater to a feedwater system 23. The feedwater system 23 heats the feedwater and returns it to the reactor 21. The feedwater system 23 includes a main feedwater pump 8, a low pressure feedwater heater 7 installed downstream of the condenser 6 and upstream of the main feedwater pump 8 and heating the feedwater supplied from the condenser 6, and a high pressure feedwater heater 9 installed downstream of the main feedwater pump 8 and upstream of the reactor 21. The feedwater discharged from the high pressure feedwater heater 9 is introduced to the reactor 21 via a feedwater line 24. Extraction lines 25, 26, 27 and 28 for extracting steam from the steam system 22 and introducing the extracted steam to the high pressure feedwater heater 9 and the low pressure feedwater heater 7 are disposed between the steam system 22 and corresponding one of the high pressure feedwater heater 9 and the low pressure feedwater heater 7. The extraction line 25 extracts steam from an intermediate point of the high pressure turbine 3 and introduces the steam to the high pressure feedwater heater 9. The extraction line 26 extracts steam from an outlet of the high pressure turbine 3 (actually some point downstream of the outlet of the high pressure turbine 3 and upstream of an inlet of the moisture separator 4) and introduces the steam to the high pressure feedwater heater 9. The extraction line 27 extracts steam from an intermediate point of the moisture separator 4 and introduces the steam to the low pressure feedwater heater 7. The extraction line 28 extracts steam from an intermediate point of the low pressure turbine 5 and introduces the steam to the low pressure feedwater heater 7. An extraction line flow adjusting valve 10, i.e., extracted steam amount adjusting means for adjusting an extracted steam amount introduced from the intermediate point of the high pressure turbine 3 to the high pressure feedwater heater 9, is disposed in the extraction line 25. The operation method of the thus-constructed nuclear power plant according to this embodiment will be described below. In FIGS. 1, 2 and 3, to explain heat balances, the reactor thermal power is represented by Q, the mass flow rate of water and steam is represented by G, and the enthalpy of water and steam is represented by H. The thermal power Q and the mass flow rate G are expressed by ratios (%) relative to the reactor thermal power and the steam flow rate at an outlet of a reactor pressure vessel, respectively, when the reactor is in the state before power uprate as shown in FIG. 2. The enthalpy is expressed by a numerical value in units of (kJ/kg). Note that each embodiment of the present invention represents the ordinary operation state and excludes the startup and shutdown states, the transient state, the operation state where the core thermal power is changed by moving a control rod position in a core to change the control rod pattern etc., and the operation state in the event of an accident. In the heat balances shown in FIGS. 1-3, as mentioned above, the thermal power Q and the mass flow rate G are expressed by ratios (%) relative to the reactor thermal power and the steam flow rate at the outlet of the reactor pressure vessel, respectively, when the reactor is in the state before power uprate as shown in FIG. 2. According to the operation method of this embodiment, as seen from the heat balances shown in FIGS. 1-3, when second reactor thermal power in a second operation cycle (FIG. 1) of the reactor 21 is uprated from first reactor thermal power (FIG. 2) in a first operation cycle (FIG. 2) prior to the second operation cycle (i.e., Q=100→105), a proportion (13/100) of the mass flow rate (G=3+10=13) of steam extracted from the high pressure steam system 22A in the second operation cycle with respect to the mass flow rate (G=100) of main steam at the reactor outlet is reduced from a proportion (19/100) of the mass flow rate (G=9+10=19) of steam extracted from the high pressure steam system 22A and introduced to the feedwater heater 9 in the first operation cycle with respect to the mass flow rate (G=100) of main steam at the reactor outlet (i.e., 19/100→13/100). Also, the temperature of the feedwater discharged from the feedwater heater 9 is lowered in the second operation cycle (H=832) from that in the first operation cycle (H=924). As described later, an extent to which the feedwater temperature is lowered in the second operation cycle from that in the first operation cycle is in the range of 1° C. to 40° C. Incidentally, one operation cycle is a period from a time at which the reactor operation is started up from the shutdown state to a time at which the reactor operation is shut down for fuel exchange. Further, the feedwater temperature is one measured at the period of the rated operation (maximum power output operation), but not the period of the partial power output operation such as the process of starting up and shutting down, as stated later. Looking from another aspect, according to the operation method of this embodiment, the mass flow rate (G=3+10=13) of steam extracted from the high pressure steam system 22A is reduced (G=19→13) in the second operation cycle (FIG. 1) from the mass flow rate (G=9+10=19) of steam extracted from the high pressure steam system 22A and introduced to the feedwater heater 9 in the first operation cycle (FIG. 2), and the temperature of the feedwater discharged from the feedwater heater 9 is lowered in the second operation cycle from that in the first operation cycle. Alternatively, it can also be said that, by making a temperature rise in the feedwater heater 9 during the second operation cycle (FIG. 1) smaller than a temperature rise in the feedwater heater 9 during the first operation cycle (FIG. 2), the temperature of the feedwater discharged from the feedwater heater 9 is lowered in the second operation cycle (H=832) from that in the first operation cycle (H=924). The operation method of this embodiment will be described in more detail below. FIG. 4 graphically shows the relationships of operation cycle versus reactor thermal power, flow rate of main steam (amount of steam flowing from the reactor pressure vessel 1 to the main steam line 2), and flow rate of extracted steam in this embodiment as compared with the power uprate by the known method. Note that one operation cycle is defined as a period from a time at which the reactor operation is started up from the shutdown state to a time at which the reactor operation is shut down for fuel exchange. Further, the feedwater temperature is one measured at the period of the rated operation (maximum power output operation), but not the period of the partial power output operation such as the process of starting up and shutting down, as stated later. Of operation cycles shown in FIG. 4, the N-th operation cycle represents a cycle before the power uprating method of the present invention is applied. At that time, the reactor thermal power is Q=100%. FIG. 2 shows one example of heat balance prior to the power uprate. The (N+1)-th operation cycle represents a cycle in which the reactor thermal power is uprated 5% to obtain Q=105%. Means for uprating the reactor thermal power can be realized by increasing the amount of withdrawal of a control rod in the (N+1)-th operation cycle from that in the N-th operation cycle, or by raising the rotation speed of the recirculation pump to increase the core flow rate in the (N+1)-th operation cycle from that in the N-th operation cycle, or by modifying the kind of a fuel assembly. Also, because the temperature of the feedwater supplied to the reactor pressure vessel 1 is lowered with application of the present invention, it is expected that the reactor thermal power is naturally uprated with coolant density feedback as a result of lowering of the coolant temperature at a core inlet. In some plant, the flow rate of the extracted steam and the flow rate of the main steam are changed during one operation cycle as shown in FIG. 5. FIG. 5 shows an example in which, in the (N+1)-th operation cycle, the reactor thermal power is reduced midway one operation cycle with a drop of core reactivity, whereupon the flow rate of the extracted steam and the flow rate of the main steam are reduced (coasted down). Other than the operation cycle shown in FIG. 5, the reactor power is also temporarily reduced, for example, when the amount of insertion of the control rod in the core is changed. For those reasons, in this embodiment, the heat balance, the flow rate of the extracted steam, the flow rate of the main steam, the core flow rate, the feedwater temperature, the reactor thermal power, the extent of heating of the feedwater, etc. are compared at an operation point where the flow rate of the main steam is maximized during the operation cycle, except for the startup and shutdown states, the operation state where the core thermal power is changed by operating the control rod, in the event of an accident or a transient phenomenon, and the test operation. In other words, such an operation point means a point where the reactor thermal power is maximized during the operation cycle. Further, when the thermal power is 100% in the (N−1)-th operation cycle, but the thermal power is largely reduced from the rated power of 100% for some reason in the N-th cycle as shown in FIG. 6, the (N−1)-th operation cycle represents the cycle before the present invention is applied (i.e., the first operation cycle), and the (N+1)-th operation cycle represents the cycle in which the present invention is applied (i.e., the second operation cycle). When the reactor thermal power is uprated, the flow rate of the feedwater has to be increased or the enthalpy difference of a coolant between the inlet and the outlet of the reactor pressure vessel has to be increased in order to remove heat that has increased in amount corresponding to the power uprate. The known power uprating method employs the former manner; namely it increases the flow rate of the feedwater in proportion to the reactor thermal power. An example of heat balance according to the known power uprating method is shown in FIG. 3. As a result of the known power uprating method, the flow rate of the main steam in the (N+1)-th operation cycle is increased to 105% as shown in FIG. 4. The present invention employs the latter manner; namely it intentionally reduces the coolant enthalpy at the inlet of the reactor pressure vessel, to thereby increase the enthalpy difference between the inlet and the outlet of the reactor pressure vessel. The coolant enthalpy at the inlet of the reactor pressure vessel can be reduced by reducing the flow rate of extracted steam from the steam system 22 and supplied to the feedwater heater 9. However, if the flow rate of the extracted steam is simply reduced as a whole, the thermal efficiency is largely reduced and the amount of generated power cannot be so increased. Such a reduction of the thermal efficiency can be suppressed by selectively reducing the extraction of steam from the high pressure steam system 22A, which is constituted by the section including the high pressure turbine 3 and extending from the reactor outlet to the inlet of the low pressure turbine 5. The reason is that the steam in the high pressure steam system 22A has higher energy than the steam in the low pressure steam system 22B, which is constituted by the section extending from the inlet of the low pressure turbine 5 to the inlet of the condenser 6, and a thermal loss is reduced by selectively reducing the extraction of steam from the high pressure steam system 22A, whereby the reduction of the thermal efficiency resulting from the power uprate can be suppressed. To selectively reduce the extraction of steam from a relatively high energy portion in the high pressure steam system 22A and to suppress the reduction of the thermal efficiency, in this embodiment, the flow rate of steam extracted from an intermediate point of the high pressure turbine 3 or the outlet of the high pressure turbine 3 (actually some point between the outlet of the high pressure turbine 3 and the inlet of the moisture separator 4) is selectively reduced so that the flow rate of steam flowing into the low pressure turbine 5 is increased and the amount of generated power is increased. A large part of the steam extracted from the intermediate point of the high pressure turbine 3 or the outlet of the high pressure turbine 3 is used in the high pressure feedwater heater 9 installed downstream of the main feedwater pump 8. Looking from another aspect, therefore, the power uprating method according to the present invention can also be said as a method of reducing the extent of adding the thermal energy to the feedwater in the region downstream of the main feed water pump 8. In a plant where the flow rate of steam extracted from the intermediate point of the high pressure turbine 3 or the outlet of the high pressure turbine 3 is originally small, the flow rate of steam extracted from the low pressure turbine 5 has to be also reduced in order to sufficiently lower the feedwater temperature. Even when the present invention is applied to such a plant, a certain effect can be obtained by reducing the flow rate of steam extracted from the intermediate point of the high pressure turbine 3 or the outlet of the high pressure turbine 3 to a larger extent. With this embodiment, in spite of increasing the reactor thermal power by 5% as compared with the N-th operation cycle, the flow rate of the main steam can be kept the same as that in the N-th operation cycle. Because this embodiment is described in connection with the ideal power uprating method, the flow rate of the main steam is the same in both the N-th operation cycle and the (N+1)-th operation cycle. However, the flow rate of the main steam is not always required to be the same in those operation cycles, and the flow rate of the main steam may be increased within the range of a design margin of the equipment including the high pressure turbine 3. When there are a plurality of extraction points usable to reduce the flow rate of the extracted steam, i.e., when there are a plurality of extraction points midway the high pressure turbine 3 or at the outlet of the high pressure turbine 3, a maximum effect can be obtained by selecting the most upstream extraction point. In such a case, while the extraction line flow adjusting valve 10 for controlling the flow rate of the extracted steam may be disposed to reduce the flow rate of the extracted steam, one or more of the extraction lines may be completely closed as an alternative manner. To close the extraction line, a shut-off valve is disposed midway the extraction line, or the extraction line is plugged. When the extraction line is completely closed, equipment for controlling the flow rate of the extracted steam is not required and the operation control is simplified. Whether to control the flow rate of the extracted steam or completely close the extraction line depends on the heat balance and the extent of power uprate in the plant. (In the case where the flow rate of the extracted steam per extraction line is too large, the feedwater temperature is excessively lowered if the extraction line is completely closed. Therefore, the flow rate of the extracted steam is adjusted in that case.) According to this embodiment, even when the reactor thermal power is uprated to increase the amount of power generated in the nuclear power plant, increases of both the flow rate of the feedwater and the flow rate of the main steam can be suppressed, whereby an increase of burdens imposed on the feedwater line 24, the main steam line 2, and the pressure vessel internals can be suppressed. As compared with the case of simply reducing the flow rate of the extracted steam as a whole, it is possible to suppress decrease of the thermal efficiency and to obtain larger electric power. Further, although the high pressure turbine 3 must be usually replaced when power is uprated to a large extent by the known power uprating method, this embodiment can provide a wider power uprate range available without replacing the high pressure turbine 3 than that provided by the known method. With the operation method of this embodiment, the feedwater temperature is lowered. The lowering of the feedwater temperature lowers the coolant temperature at the core inlet and increase the thermal margin of the core (corresponding to MCPR (Minimum Critical Power Ratio) in BWR), thus resulting in an advantage of ensuring higher safety than that obtained with the known power uprating method. The conventional power uprate increases the core pressure loss and reduces the safety margin if the same fuels are used. In contrast, the power uprating method of the present invention lowers the coolant temperature at the core inlet and reduces the void rate and the absolute value of the void coefficient in the core. Therefore, the core pressure loss is reduced and a reduction of the safety margin of the core is suppressed. Further, the design margin for the pressure rising transient state is increased with the reduction of the void rate and the absolute value of the void coefficient in the core. Thus, the lowering of the feedwater temperature is effective in suppressing the deterioration of core characteristics and the reduction of the design margin in the boiling water reactor during the power update operation. Generally, because feedwater temperature control is not especially performed in the boiling water reactor, the feedwater temperature may change to the extent of smaller than 1° C. even in the same boiling water reactor and at the same core thermal power due to change of heat balance in the entire plant, in particular temperature change of the coolant (seawater in many cases) that is used to condense the steam by the condenser 6 shown in FIG. 1. In this embodiment, an extent to which the feedwater temperature is lowered is set to about 20° C. Regarding the extent to which the feedwater temperature should be lowered to compensate for the deterioration of the core characteristics in the power uprate operation, however, the effect of this embodiment can be obtained with a significant result by lowering the feedwater temperature by a value of not smaller than 1° C. that corresponds to the magnitude at which the feedwater temperature is changed in the ordinary operation. In addition, when the feedwater enters the reactor pressure vessel 1, it is mixed with water at the saturation temperature inside the reactor. Accordingly, there is a temperature difference between the feedwater line 24 and the pressure vessel 1. If the feedwater temperature is too lowered, the temperature difference in such a mixing area is increased, thus causing a risk that a design limit is exceeded from the viewpoint of thermal fatigue. From this point of view, a limit of the extent to which the feedwater temperature should be lowered from the current operation temperature is 40° C. The reduction of the core pressure loss means that an increase of loads imposed on the jet pump and the recirculation pump, which are used for recirculation of the coolant, due to the power uprate can also be suppressed. Further, since the amount of increase in quantity of steam generated in the core is comparatively smaller than that of the thermal power, an influence upon the carry under caused by entrainment with the recirculation water can also be kept small and the flow window can be easily ensured even in the case of large power uprate. Table 1, given below, shows the relationships among the reactor thermal power, the flow rate of the main steam, the flow rate of the extracted steam, and the enthalpy of the feedwater when the power uprating method of this embodiment is applied to the cases of uprating the power at various rates. Each value of the reactor thermal power and the flow rate of the main steam represents a ratio relative to that at 100% of the reactor thermal power, and a value of the flow rate of the extracted steam represents a ratio relative to the flow rate of the main steam at 100% of the reactor thermal power. As seen from Table 1, the power uprating method of this embodiment can be applied over a wide range including even the case where the reactor thermal power is uprated to 110%. The reason why Table 1 shows only the power uprate up to 110% is that the power uprate in excess of 110% requires replacement of the moisture separator 4. If the replacement of the moisture separator 4 is allowed or a combination with, e.g., an increase of the core pressure and/or with employment of a moisture separator and reheater is considered, the power uprating method of this embodiment can be applied over a wider range of power uprate. In the boiling water reactor, uprate of the reactor thermal power up to about 102% is generally feasible just by increasing the measurement accuracy of a feedwater flowmeter, etc. Therefore, the present invention is more effective when applied to the case of uprating the reactor thermal power in excess of 102%. Further, at the power uprate up to about 105% of the reactor thermal power, substantial change of system equipment, e.g., replacement of the high pressure turbine 3, is not required in general. The effect of this embodiment is especially noticeable when applied to the uprate of the reactor thermal power in excess of 105% because the replacement of the high pressure turbine 3 is not required even in that case. TABLE 1Reactor thermalFlow rate ofFlow rate of ex-Enthalpy ofpower (%)main steam (%)tracted steam (%)feedwater (kJ/kg)100100459241031004386910510042831107100407951101003873911010542831 In view of the above-mentioned fact that substantial change of system equipment, e.g., replacement of the high pressure turbine 3, is not required in general at the ordinary power uprate up to about 105% of the reactor thermal power, an application method of this embodiment may be changed between the case of uprating the reactor thermal power at a rate of not larger than 5% and the case of uprating the reactor thermal power at a rate of larger than 5%. More specifically, because the replacement of the high pressure turbine 3 is not required when the increase rate of the reactor thermal power is not larger than 5%, it is primarily intended to keep the above-mentioned core characteristic (core average void rate) the same as that before the power uprate. Assuming that the reactor thermal power before the power uprate is Q1 (kW), the reactor thermal power after the power uprate is Q2 (kW), and the increase rate of the power uprate is A(%), the power uprate of not larger than 5% is expressed by A≦5. Also, on an assumption that the core flow rate after the power uprate is W (kg/s), the specific heat at constant pressure at about 200° C. under 7 MPa, i.e., the operating pressure in a typical boiling water reactor, is about 4.5 (kJ/kg·K), and a proportion of the flow rate of the feedwater with respect to the core flow rate in the typical boiling water reactor is about 13%, the condition for keeping the core average void rate the same as that before the power uprate is given as follows. Change in thermal value of the feedwater per 1° C. of the feedwater temperature is expressed by:W (kg/s)×13(%)/100(%)×4.5 (kJ/kg·K)=W×13/100×4.5 (kW/K)Assuming here that the feedwater temperature before the power uprate is T1 and the feedwater temperature after the power uprate is T2, the feedwater temperature T2 required to reduce the thermal value of the feedwater, which is equivalent to the thermal value (Q2−Q1) (kW) corresponding to the power uprate, is determined by the following equation:Q2−Q1=W×13/100×4.5×(T1−T2) In order to hold the core characteristics equivalent to or better than those before the power uprate, the feedwater thermal value is just required to be reduced by an amount of not smaller than the thermal value increased with the power uprate. The condition to meet such a requirement is expressed by:Q2−Q1≦W×13/100×4.5×(T1−T2)This formula is rewritten into:T2≦T1−7.7×(Q2−Q1)/(4.5×W)Stated another way, by setting the feedwater temperature so as to satisfy the above formula in the case of the power uprate of not larger than 5%, the core characteristics, such as the thermal margin, the pressure loss characteristic, the safety margin, and the design margin in the transient state of the core, can be basically held equivalent to or better than those before the power uprate. Further, since the flow rate of the main steam is kept equivalent to or reduced from that before the power uprate, the design margin of the main steam system, including the high pressure turbine and the dryer, can also be held equivalent to or better than those before the power uprate. A description is made of the feedwater temperature when the nuclear thermal power is uprated at a rate of larger than 5%, but not larger than 10%. When the nuclear thermal power is uprated at a rate of larger than 5% by the known power uprating method, the flow rate of the main steam is also increased in excess of 5%, thus generally resulting in that the design margin of equipment, e.g., the high pressure turbine, is exceeded. Therefore, such equipment has to be replaced. In that case, by lowering the feedwater temperature as in this embodiment, the increase in the flow rate of the main steam can be held not larger than 5%. Assuming here that the increase in the flow rate of the main steam up to 5% is allowed, a reduction of the feedwater thermal value required when the power uprate is increased at A(%) corresponds to the power increase of (A−5) % and is expressed by:Q2×(A−5+100)/100−Q1=Q2×(A+95/100)−Q1On the assumption that the feedwater temperature before the power uprate is T1 and the feedwater temperature after the power uprate is T2, the above thermal value can be offset when T1 and T2 satisfy the following formula:Q2×(A+95/100)−Q1=W×13/100×4.5×(T1−T2)Thus, the increase in the flow rate of the main steam can be held not larger than 5% by satisfying:Q2×(A+95/100)−Q1≦W×13/100×4.5×(T1−T2)With rewrite of this formula, T2 is expressed by:T2≦T1−7.7×(Q2×(A+95)/100)−Q1)/(4.5×W)In the case of the power uprate being larger than 5%, but not larger than 10%, therefore, if the feedwater temperature T2 satisfies the above formula, the power uprate can be realized within the range of the design margin of the high pressure turbine, etc. and the replacement of such equipment is not required. Further, the design margins of the high pressure turbine and the core can be held equivalent to or larger than those resulting at the power uprate of 5% by the known method. At the power uprate in excess of 10%, it is generally required in the boiling water reactor to replace, e.g., the moisture separator in addition to the high pressure turbine, etc. This problem can be overcome by lowering the feedwater temperature in a similar manner to hold the increase in the flow rate of the main steam to be not larger than 10% so that the replacement of the moisture separator is not required. The condition of meeting such a requirement in this case is given as follows for the rate A(%) of the power uprate based on the same concept as that when the power uprate is larger than 5%, but not larger than 10%:Q2×(A−10+100)/100−Q1≦W×13/100×4.5×(T1−T2)With rewrite of this formula, T2 is expressed by:T2≦T1−7.7×(Q2×(A+90)/100−01)/(4.5×W)This indicates an extent of lowering of the feedwater temperature within which the replacement of the moisture separator is not required at the power uprate of larger than 10%. In that case, the design margins of the high pressure turbine and the core can also be held equivalent to or larger than those resulting at the power uprate of 10% by the known method. At any of the power uprate of not larger than 5%, the power uprate of larger than 5%, but not larger than 10%, and the power uprate of larger than 10%, the extent of lowering of the feedwater temperature in excess of 40° C. is undesired in practice from the viewpoint of thermal fatigue. Also, as the power uprate increases, the thermal margin of the core is reduced. It can be generally said that the thermal margin of the core is bearable for the power uprate up to about 20% by employing new fuel. Another conceivable solution is to improve, e.g., a pump for increasing the core flow rate. Even in consideration of such an improvement as well, the power uprate at a rate of about 30% is regarded to be a limit from the viewpoint of the core characteristics. Further, looking at the equipment side, the power uprate over 30% is also not desired in practice because such large power uprate exceeds the design limits of the low pressure turbine and the condenser, which are more expensive than the high pressure turbine, and hence requires replacement of those other units of equipment. According to this embodiment, for the power uprate of not larger than 5%, the power uprate can be realized while holding the design margins of the high pressure turbine and the core equivalent to or larger than those before the power uprate. For the power uprate of larger than 5%, but not larger than 10%, the power uprate can be realized up to 10% while holding the design margins of the high pressure turbine and the core equivalent to or larger than those resulting at the power uprate of 5% by the known method. For the power uprate of larger than 10%, the power uprate can be realized in excess of 10% with no need of replacing the moisture separator, etc. while holding the design margins of the moisture separator and the core equivalent to or larger than those resulting at the power uprate of 10% by the known method. The present invention can be modified in various ways without being restricted to the above-described embodiment. For example, in the boiling water reactor system, a moisture separator and heater or a moisture separator and reheater 11, shown in FIG. 7, may be used instead of the moisture separator 4. Even in that case, although a steam extraction line 31 and a drain line 32 are added, the operation method of the present invention can be applied as in the above-described embodiment and can provide similar advantages without causing substantial change with regards to main parameters, such as the feedwater temperature and the flow rate of the main steam. While the above embodiment is described as applying the present invention to the boiling water reactor power plant, the present invention is applicable to a pressurized water reactor system as well. Another embodiment of the present invention in which the invention is applied to the boiling water reactor power plant as one of nuclear power plants will be described below with reference to FIGS. 8 and 9. The same components as those in the first embodiment are denoted by the same numerals. As in the first embodiment, the boiling water reactor power plant of this embodiment comprises a reactor pressure vessel 1, a main steam line 2, a high pressure turbine 3 and a low pressure turbine 5 connected to the main steam line 2 in series, and a moisture separator 4 (or a moisture separator and heater) disposed in the main steam line 2 between the high pressure turbine 3 and the low pressure turbine 5. A low pressure feedwater heater 7, a feedwater pump 8, and a high pressure feedwater heater 9 are installed in a feedwater system 23 downstream of a condenser 6. When the reactor thermal power is uprated, the flow rate of the feedwater has to be increased or the enthalpy difference of the coolant between the inlet and the outlet of the reactor pressure vessel 1 has to be increased in order to remove heat that has increased in amount corresponding to the power uprate. The known power uprating method employs the former manner; namely it increases the flow rate of the feedwater in proportion to the reactor thermal power. On the other hand, as a new power uprating method, there is also proposed a method of suppressing increases of both the flow rate of the main steam and the flow rate of the feedwater in the power uprate operation based on the latter manner by intentionally reducing the coolant enthalpy (temperature) at the inlet of the reactor pressure vessel, to thereby increase the enthalpy difference between the inlet and the outlet of the reactor pressure vessel. This embodiment is adapted for such a new power uprating method and requires additional equipment for widening a feedwater temperature controllable range so that the feedwater temperature is lowered to a value beyond the range estimated in the stage of plant construction. The necessity of widening the feedwater temperature controllable range toward the lower temperature side in turn requires the flow rate of steam extracted for heating the feedwater to be reduced in comparison with that before the power uprate. The steam extracted from the high pressure turbine 3 for heating the feedwater is introduced to the high pressure feedwater heater 9 via extraction lines 25 and 26. Also, the steam extracted from the low pressure turbine 5 is sent to the low pressure feedwater heater 7 via an extraction line 28. An extracted flow control valve 38 is disposed in the extraction line 25 to adjust the flow rate of the extracted steam. In the boiling water reactor power plant of this embodiment, a plurality of main extraction lines are installed downstream of the inlet of the high pressure turbine and upstream of the outlet of the low pressure turbine. In the boiling water reactor power plant of this embodiment to which the above-mentioned new power uprating method is applied, it is important that the feedwater temperature be surely lowered to a preset value. For that reason, a feedwater temperature sensor 39 is disposed in the feedwater system 23 downstream of the high pressure feedwater heater 9 that is located most downstream in the feedwater system 23. The feedwater temperature sensor 39 measures the temperature of the feedwater discharged from the high pressure feedwater heater 9 and outputs a feedwater temperature measured value signal 41. An extracted flow controller 40 controls the opening of the extracted flow control valve 38 to adjust the flow rate of the extracted steam. The feedwater temperature sensor 39 is disposed in the feedwater system 23 downstream of the high pressure feedwater heater 9, to which the extracted steam is supplied at the adjusted flow rate, and upstream of the inlet of the reactor pressure vessel 1. Alternatively, the feedwater temperature sensor 39 may be disposed between the high pressure feedwater heater 9 and another suitable feedwater heater installed downstream of the former. One example of control logic executed by the extracted flow controller 40 in the second embodiment will be described below with reference to FIG. 9. The extracted flow controller 40 receives the feedwater temperature measured value signal 41 outputted from the feedwater temperature sensor 39 and a feedwater temperature set value signal 42. Based on the feedwater temperature measured value signal 41 and the feedwater temperature set value signal 42, the extracted flow controller 40 produces an opening demand signal 43 and outputs the opening demand signal 43 to the extracted flow control valve 38. If the measured value of the feedwater temperature is lower than the set value of the feedwater temperature, this means that the flow rate of the extracted steam is insufficient. Therefore, the extracted flow controller 40 outputs the opening demand signal 43 to increase the opening of the extracted flow control valve 38. Conversely, if the measured value of the feedwater temperature is higher than the set value, this means that the flow rate of the extracted steam is too large. Therefore, the extracted flow controller 40 outputs the demand signal 43 to reduce the opening of the extracted flow control valve 38. According to this embodiment, since the feedwater temperature can be adjusted to the set value through control of the opening of the extracted flow control valve 38, it is possible to suppress variations in the amount of power generated during the power uprate operation of the nuclear power plant. Also, according to this embodiment, since the feedwater temperature can be always held at the set value, it is possible to suppress the increases of both the flow rate of the main steam and the flow rate of the feedwater by lowering the feedwater temperature in the power uprate operation of the reactor. Further, since the feedwater temperature can be adjusted in real time, even in the case of changing the thermal power of the nuclear power plant, the operation method of this embodiment is adaptable for the load following operation of the nuclear power plant by modifying the set value of the feedwater temperature depending on the change of the thermal power, while the flow rate of the main steam and the flow rate of the feedwater are held constant. A boiling water reactor power plant according to still another embodiment (third embodiment) of the present invention will be described below with reference to FIG. 10. This third embodiment differs from the second embodiment in that, instead of the feedwater temperature sensor 39, an extraction flowmeter 44 is disposed in the extraction line 25 in which the extracted flow control valve 38 is disposed. The extracted flow controller 40 receives a measured value from the extraction flowmeter 44 and a set value of the flow rate of the extracted steam. The extraction flowmeter 44 measures the flow rate of the extracted steam supplied to the high pressure feedwater heater 9. The extracted flow control valve 38 and the extraction flowmeter 44 may be disposed in the extraction line 25 irrespective of which one of them is positioned upstream of the other. When the extraction line 25 is merged midway with another extraction line (e.g., the extraction line 26), one of the extracted flow control valve 38 and the extraction flowmeter 44, which is positioned on the downstream side, may be disposed in the extraction line 25 downstream of the merging point between the two extraction lines. Also, when the extraction line 25 is branched midway, one of the extracted flow control valve 38 and the extraction flowmeter 44, which is positioned on the downstream side, may be disposed in a line after being branched. If the reactor thermal power, the flow rate of the feedwater, and the flow rate of the extracted steam are known, the feedwater temperature is uniquely decided from the heat balance of the nuclear power plant. Accordingly, measuring the flow rate of the extracted steam by the extraction flowmeter 44 disposed in the extraction line 25, as in this embodiment, is equivalent to measurement of the feedwater temperature. In this embodiment, the extracted flow controller 40 receives an extracted flow measured value signal 45 outputted from the extraction flowmeter 44 and a set value signal 42A for the flow rate of the extracted steam. Based on the extracted flow measured value signal 45 and the extracted flow set value signal 42A, the extracted flow controller 40 outputs an opening demand signal 43 for the extracted flow control valve 38. The opening of the extracted flow control valve 38 is controlled in accordance with the opening demand signal 43. If the measured value of the flow rate of the extracted steam is smaller than the set value of the flow rate of the extracted steam, this means that the flow rate of the extracted steam is insufficient. Therefore, the extracted flow controller 40 outputs the opening demand signal 43 to increase the opening of the extracted flow control valve 38. Conversely, if the measured value of the flow rate of the extracted steam is larger than the set value, this means that the flow rate of the extracted steam is too large. Therefore, the extracted flow controller 40 outputs the opening demand signal 43 to reduce the opening of the extracted flow control valve 38. This third embodiment can also provide similar advantages to those obtainable with the second embodiment. A boiling water reactor power plant according to still another embodiment (fourth embodiment) of the present invention will be described below with reference to FIG. 11. In this fourth embodiment, both the technical ideas of the second and third embodiments are combined with each other. More specifically, the feedwater temperature sensor 39 is disposed in the feedwater system 23 as in the second embodiment, and the extraction flowmeter 44 is disposed in the extraction line 25 as in the third embodiment. With such an arrangement, the extracted flow controller 40 in this fourth embodiment performs the control based on the flow rate of the extracted steam and the control based on the feedwater temperature. FIG. 12 shows one example of control logic used in this fourth embodiment. A certain time delay occurs from adjustment of the extracted flow control valve 38 to actual change of the feedwater temperature. In this embodiment, therefore, the control based on the flow rate of the extracted steam is performed with priority when the flow rate of the extracted steam is fluctuated at a short cycle, and the control based on the feedwater temperature is performed with priority when the feedwater temperature continues to lower or rise for a relatively long time. In practice, as shown in FIG. 12, the set value of the flow rate of the extracted steam is determined through the control based on the feedwater temperature, and the opening command signal 43 for the extracted flow control valve 38 is outputted based on the difference between the determined set value and the measured value of the flow rate of the extracted steam. This fourth embodiment can also provide similar advantages to those obtainable with the second embodiment. A boiling water reactor power plant according to still another embodiment (fifth embodiment) of the present invention will be described below with reference to FIG. 13. In this fifth embodiment, a main steam flowmeter 46 is disposed in the main steam system 2 between the reactor pressure vessel 1 and the high pressure turbine 3. If the reactor thermal power and the flow rate of the main steam are known, the feedwater temperature is uniquely decided from the heat balance of the nuclear power plant. Accordingly, measuring the flow rate of the main steam by the main steam flowmeter 46 disposed in the main steam system 22, as in this fifth embodiment, is equivalent to the measurement of the feedwater temperature in the second embodiment. The arrangement of this fifth embodiment is similar to that of the second embodiment except that the main steam flowmeter 46 is disposed instead of the feedwater temperature sensor 39 and the extracted flow controller 40 controls the extracted flow control valve 38 in accordance with the measured value from the main steam flowmeter 46. In this embodiment, the extracted flow controller 40 receives a main-steam flow measured value signal 47 outputted from the main steam flowmeter 46 and a set value signal 48 for the flow rate of the main steam. Then, the extracted flow controller 40 produces an opening demand signal 43 based on those two signals and outputs it to the extracted flow control valve 38. If the measured value of the flow rate of the main steam is smaller than the set value of the flow rate of the main steam, this means that the feedwater temperature is too low, namely the flow rate of the extracted steam is insufficient. Therefore, the extracted flow controller 40 outputs the opening demand signal 43 to increase the opening of the extracted flow control valve 38. Conversely, if the measured value of the flow rate of the main steam is larger than the set value, this means that the feedwater temperature is too high, namely the flow rate of the extracted steam is too large. Therefore, the extracted flow controller 40 outputs the opening demand signal 43 to reduce the opening of the extracted flow control valve 38. This fifth embodiment can also provide similar advantages to those obtainable with the second embodiment. A boiling water reactor power plant according to still another embodiment (sixth embodiment) of the present invention will be described below with reference to FIG. 14. In this sixth embodiment, the technical ideas of the second, third and fifth embodiments are combined together. More specifically, the feedwater temperature sensor 39 is disposed in the feedwater system 23 as in the second embodiment, the extraction flowmeter 44 is disposed in the extraction line 25 as in the third embodiment, and the main steam flowmeter 46 is disposed in the main steam system 2 as in the fifth embodiment. With such an arrangement, the extracted flow controller 40 in this sixth embodiment performs the control based on the flow rate of the main steam, the control based on the flow rate of the extracted steam, and the control based on the feedwater temperature. FIG. 15 shows one example of control logic used in this sixth embodiment. A certain time delay occurs from adjustment of the extracted flow control valve 38 to actual change of the flow rate of the main steam, and another certain time delay also occurs, though being shorter than the time delay regarding the flow rate of the main steam, from adjustment of the extracted flow control valve 38 to actual change of the feedwater temperature. On the other hand, a time delay from adjustment of the extracted flow control valve 38 to actual change of the flow rate of the extracted flow is short. In this embodiment, therefore, the control based on the flow rate of the extracted steam is performed with priority when the flow rate of the extracted steam is fluctuated at a short cycle, and the control based on the feedwater temperature is performed with priority when the feedwater temperature continues to lower or rise for a relatively long time. Further, the control based on the flow rate of the main steam is performed with priority when the flow rate of the main steam continues to increase or reduce for a relatively long time. In practice, as shown in FIG. 15, the set value of the feedwater temperature is determined through control based on the flow rate of the main steam, and the set value of the flow rate of the extracted steam is determined based on the difference between the determined set value and the measured value of the feedwater temperature. Then, the opening demand signal 43 for the extracted flow control valve 38 is outputted based on the difference between the determined set value and the measured value of the flow rate of the extracted steam. This sixth embodiment can also provide similar advantages to those obtainable with the second embodiment. Instead of the above-described arrangement, the main steam flowmeter 46 and the feedwater temperature sensor 39 may be disposed to perform the control based on the flow rate of the main steam and the control based on the feedwater temperature. Further, the main steam flowmeter 46 and the extraction flowmeter 44 may be disposed to perform the control based on the flow rate of the main steam and the control based on the flow rate of the extracted steam. Generally, when the flow rate of the extracted steam is reduced to lower the feedwater temperature, the thermal efficiency of the plant is reduced. In order to suppress such a reduction of the thermal efficiency, it is preferable to reduce the flow rate of the steam extracted from the extraction point that is positioned as close as possible to the uppermost side. Therefore, a greater effect can be obtained by installing the extracted flow control valve 38 in the extraction line for extracting the steam from a point downstream of the inlet of the high pressure turbine and upstream of the inlet of the lower pressure turbine. While the extracted flow control valve 38 is disposed only in one extraction line in the second to sixth embodiments, the feedwater temperature cannot be sufficiently lowered in some cases by reducing the flow rate of the extracted steam through only one extraction line. In such a case, the extracted flow control valve 38 is disposed in plural extraction lines. In the boiling water reactor power plant, as mentioned above, uprate of the reactor thermal power up to about 102% is generally feasible just by increasing the measurement accuracy of the feedwater flowmeter 39, etc., and the feedwater temperature is not required to be lowered in such an uprate range. In the boiling water reactor power plant, therefore, the control of the feedwater temperature described in the second to sixth embodiments is more effective when applied to the case of uprating the reactor thermal power in the range of larger than 102%, but smaller than 105%. Further, at the uprate of the nuclear thermal power in the range of 105% to 120%, substantial change of system equipment, e.g., replacement of the high pressure turbine 3, is not required in general. The effects of those embodiments are especially noticeable when applied to the uprate of the reactor thermal power in excess of 105% because the replacement of the high pressure turbine 3 is not required even in the power uprate operation in excess of 105% by employing the power uprating method of lowering the feedwater temperature according to the present invention. While the second to sixth embodiments have been described in connection with the case where the power uprating method of lowering the feedwater temperature is applied to the boiling water reactor power plant, the following description is made of an example in which the power uprating method is applied to a pressurized water reactor power plant as one of reactor power plants. The pressurized water reactor power plant according to still another embodiment (seventh embodiment) of the present invention will be described below with reference to FIG. 16. Feedwater temperature control logic used in this seventh embodiment is the same as that shown in FIG. 9. In this seventh embodiment, a steam generator 49 is newly installed in addition to the construction of the second embodiment such that a primary loop and a secondary loop are formed. The primary loop is a circulation loop starting from the reactor pressure vessel 1, passing the steam generator 49, and returning to the reactor pressure vessel 1. The secondary loop is formed by connecting both the main steam system 2 and the feedwater system 23 in the second embodiment to the steam generator 49. The high-temperature coolant generated from the reactor pressure vessel 1 is supplied to the steam generator 49 and is returned to the reactor pressure vessel 1 after the coolant temperature has been lowered. In the steam generator 49, the feedwater supplied from the feedwater system 23 is heated by the high-temperature coolant to become steam. The secondary steam delivered from the steam generator 49 is introduced to the high-pressure turbine 3, the moisture separator and heater 24, and the low-pressure turbine 5 through the main steam line 2. The steam discharged from the low-pressure turbine 5 is condensed by the condenser 6 to become water. This water is fed to the steam generator 49 through the feedwater system 23 in which the low pressure feedwater heater 7, the feedwater pump 8, and the high pressure feedwater heater 9 are disposed. Note that one operation cycle is defined as a period from startup of the reactor to the time at which the reactor is shut down for fuel exchange. When the reactor thermal power is uprated, the amount of heat exchange in the steam generator 49 is increased substantially proportional to the amount of increase of the reactor thermal power. In order to offset the amount of heat exchange in the steam generator 49, the flow rate of the feedwater supplied to the steam generator 49 has to be increased or the enthalpy difference of the coolant between the inlet and the outlet of the steam generator 49 has to be increased. The known power uprating method employs the former manner; namely it increases the flow rate of the feedwater in proportion to the amount of heat exchange in the steam generator 49. On the other hand, this embodiment employs the latter manner, i.e., the new power uprating method. More specifically, in this embodiment, increases of both the flow rate of the main steam and the flow rate of the feedwater in the power uprate operation are suppressed by intentionally reducing the coolant enthalpy (temperature) at the inlet of the steam generator 49, to thereby increase the enthalpy difference between the inlet and the outlet of the steam generator 49. Thus, this embodiment is adapted for the new power uprating method and requires additional equipment for widening a feedwater temperature controllable range so that the feedwater temperature is lowered to a value beyond the range estimated in the stage of plant construction. The necessity of widening the feedwater temperature controllable range toward the lower temperature side in turn requires the flow rate of steam extracted for heating the feedwater to be reduced in comparison with that before the power uprate. The extracted steam for heating the feedwater is extracted from the main steam system 2 including the high pressure turbine 3 and the low pressure turbine 5, and is introduced to the high pressure feedwater heater 9 and the low pressure feedwater heater 7 via the extraction lines 25, 26 and 28, etc. In the pressurized water reactor power plant of this embodiment, a plurality of main extraction lines are installed downstream of the inlet of the high pressure turbine and upstream of the outlet of the low pressure turbine. To reduce the flow rate of the extracted steam, an extracted flow control valve 38 is disposed in the extraction line 25 to adjust the flow rate of the extracted steam. When the above-mentioned new power uprating method of increasing the enthalpy difference of the coolant is employed, it is important that the feedwater temperature be surely lowered to a preset value. For that reason, as in the second embodiment, a feedwater temperature sensor 39 is disposed in the feedwater system 23 downstream of the high pressure feedwater heater 9 that is located most downstream in the feedwater system 23, and the opening of the extracted flow control valve 38 is controlled to adjust the flow rate of the extracted steam. The feedwater temperature sensor 39 is disposed in the feedwater system 23 downstream of the high pressure feedwater heater 9, to which the extracted steam is supplied at the controlled flow rate, and upstream of the inlet of the steam generator 28. As an alternative, the feedwater temperature sensor 39 may be disposed between the high pressure feedwater heater 9, to which the extracted steam is supplied at the controlled flow rate, and another suitable feedwater heater installed downstream of the former. Further, the sensor 39 may be disposed between an outlet of a high pressure feedwater heater installed most downstream and an inlet of the steam generator 49. One example of control logic executed by the extracted flow controller 40 in the seventh embodiment will be described below with reference to FIG. 9. As in the second embodiment, the extracted flow controller 40 controls the opening of the extracted flow control valve 38 in accordance with the feedwater temperature measured value signal 41 and the feedwater temperature set value signal 42, thereby controlling the flow rate of the extracted steam for heating the feedwater, which is supplied to the high pressure feedwater heater 9. According to this embodiment, since the feedwater temperature can be adjusted to the set value through control of the opening of the extracted flow control valve 38, it is possible to suppress variations in the amount of power generated during the power uprate operation of the nuclear power plant. Also, according to this embodiment, since the feedwater temperature can be always held at the set value, it is possible to suppress the increases of both the flow rate of the main steam and the flow rate of the feedwater by lowering the feedwater temperature in the power uprate operation of the reactor. Further, since the feedwater temperature can be adjusted in real time, the operation method of this embodiment is adaptable for the load following operation of the nuclear power plant, as with the second embodiment, while the flow rate of the main steam and the flow rate of the feedwater are held constant. A pressurized water reactor power plant according to still another embodiment (eighth embodiment) of the present invention will be described below with reference to FIG. 17. As in the third embodiment, this eighth embodiment includes, instead of the feedwater temperature sensor 39, an extraction flowmeter 44 disposed in the extraction line 25 in which the extracted flow control valve 38 is disposed. Comparing with the arrangement of the seventh embodiment that the extracted flow controller 40 controls the extracted flow control valve 38 in accordance with the measured value from the feedwater temperature sensor 39, this eighth embodiment is modified such that the extracted flow controller 40 controls the extracted flow control valve 38 in accordance with the measured value from the extraction flowmeter 44. The extracted flow control valve 38 and the extraction flowmeter 44 may be disposed in the extraction line 25 irrespective of which one of them is positioned upstream of the other. When the extraction line 25 is merged midway with another extraction line, one of the extracted flow control valve 38 and the extraction flowmeter 44, which is positioned on the downstream side, may be disposed in the extraction line 25 downstream of the merging point between the two extraction lines. Also, when the extraction line 25 is branched midway, one of the extracted flow control valve 38 and the extraction flowmeter 44, which is positioned on the downstream side, may be disposed in a line after being branched. If the reactor thermal power, the flow rate of the feedwater, and the flow rate of the extracted steam are known, the feedwater temperature is uniquely decided from the heat balance of the nuclear power plant. As in the third embodiment, the extracted flow controller 40 controls the opening of the extracted flow control valve 38 in accordance with the extracted flow measured value signal 45 and the extracted flow set value signal 42A. This eighth embodiment can also provide similar advantages to those obtainable with the seventh embodiment. A pressurized water reactor power plant according to still another embodiment (ninth embodiment) of the present invention will be described below with reference to FIG. 18. In this ninth embodiment, both the technical ideas of the seventh and eighth embodiments are combined with each other. More specifically, the feedwater temperature sensor 39 is disposed in the feedwater system 23 as in the seventh embodiment, and the extraction flowmeter 44 is disposed in the extraction line 25 as in the eighth embodiment. With such an arrangement, the extracted flow controller 40 in this seventh embodiment performs the control based on the flow rate of the extracted steam and the control based on the feedwater temperature. Control logic used in this ninth embodiment is the same as that used in the fourth embodiment and shown in FIG. 12. A certain time delay occurs from adjustment of the extracted flow control valve 38 to actual change of the feedwater temperature. In this embodiment, therefore, the control based on the flow rate of the extracted steam is performed with priority when the flow rate of the extracted steam is fluctuated at a short cycle, and the control based on the feedwater temperature is performed with priority when the feedwater temperature continues to lower or rise for a relatively long time. In practice, as shown in FIG. 12, the set value of the flow rate of the extracted steam is determined through the control based on the feedwater temperature, and the opening demand signal 43 for the extracted flow control valve 38 is outputted based on the difference between the determined set value and the measured value of the flow rate of the extracted steam. This ninth embodiment can also provide similar advantages to those obtainable with the seventh embodiment. A pressurized water reactor power plant according to still another embodiment (tenth embodiment) of the present invention will be described below with reference to FIG. 19. This tenth embodiment employs the main steam flowmeter 46 used in the fifth embodiment. The main steam flowmeter 46 is disposed in the main steam system 2 downstream of the steam generator 49 and upstream of the inlet of the high pressure turbine 3. Comparing with the arrangement of the seventh embodiment that the extracted flow controller 40 controls the extracted flow control valve 38 in accordance with the measured value from the feedwater temperature sensor 39, this tenth embodiment is modified such that the extracted flow controller 40 controls the extracted flow control valve 38 in accordance with the measured value from the main steam flowmeter 46. As in the fifth embodiment, the extracted flow controller 40 receives a main-steam flow measured value signal 47 outputted from the main steam flowmeter 46 and a set value signal 48 for the flow rate of the main steam. Then, the extracted flow controller 40 produces an opening demand signal 43 based on those two signals and controls the extracted flow control valve 38 in accordance with the produced opening demand signal 43. This tenth embodiment can also provide similar advantages to those obtainable with the seventh embodiment. A pressurized water reactor power plant according to still another embodiment (eleventh embodiment) of the present invention will be described below with reference to FIG. 20. In this eleventh embodiment, the extraction flowmeter 44 and the main steam flowmeter 46 are added to the arrangement of the seventh embodiment, as in the sixth embodiment. The extraction flowmeter 44 is disposed in the extraction line 25, and the main steam flowmeter 46 is disposed in the main steam system 2 between the steam generator 49 and the high pressure turbine 3. With such an arrangement, the extracted flow controller 40 in this embodiment performs the control based on the feedwater temperature, the control based on the flow rate of the main steam, and the control based on the flow rate of the extracted steam. Control logic used in this embodiment is the same as that shown in FIG. 15, and the extracted flow controller 40 executes the control in the same manner as that in the sixth embodiment. This eleventh embodiment can also provide similar advantages to those obtainable with the seventh embodiment. Instead of the above-described arrangement, the main steam flowmeter 46 and the feedwater temperature sensor 39 may be disposed to perform the control based on the flow rate of the main steam and the control based on the feedwater temperature. Further, the main steam flowmeter 46 and the extraction flowmeter 44 may be disposed to perform the control based on the flow rate of the main steam and the control based on the flow rate of the extracted steam. While the seventh, eighth and tenth embodiments of the present invention have been described, by way of example, in connection with the pressurized water reactor, the present invention can also be applied to an indirect-cycle plant other than the pressurized water reactor. Generally, when the flow rate of the extracted steam is reduced to lower the feedwater temperature, the thermal efficiency of the plant is reduced. In order to suppress such a reduction of the thermal efficiency, it is preferable to reduce the flow rate of the steam extracted from the extraction point that is positioned as possible as close to uppermost side. In the seventh to eleventh embodiments, therefore, a greater effect can be obtained by installing the extracted flow control valve 38 in the extraction line for extracting the steam from a point downstream of the inlet of the high pressure turbine and upstream of the inlet of the lower pressure turbine. While the extracted flow control valve 38 is disposed only in one extraction line in the seventh to eleventh embodiments, the feedwater temperature cannot be sufficiently lowered in some cases if the flow rate of the extracted steam through only one extraction line is reduced. In such a case, the extracted flow control valve 38 is disposed in plural extraction lines. In the pressurized water reactor power plant, uprate of the reactor thermal power up to about 102% is generally feasible just by increasing the measurement accuracy of the feedwater flowmeter 39, etc., and the feedwater temperature is not required to lowered in such an uprate range, as mentioned above. In the pressurized water reactor power plant, therefore, the control of the feedwater temperature described in the seventh to eleventh embodiments is more effective when applied to the case of uprating the reactor thermal power in the range of larger than 102%, but smaller than 105%. Further, at the uprate of the nuclear thermal power in the range of 105% to 120%, substantial change of system equipment, e.g., replacement of the high pressure turbine 3, is not required in general. The effects of those embodiments are especially noticeable when applied to the uprate of the reactor thermal power in excess of 105% because the replacement of the high pressure turbine 3 is not required even in the power uprate operation in excess of 105% by employing the power uprating method of lowering the feedwater temperature according to the present invention. |
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description | The entire disclosure of Japanese Patent Application No. 2015-066676 filed on Mar. 27, 2015 including description, claims, drawings, and abstract are incorporated herein by reference in its entirety. Field of the Invention The present invention relates to a scintillator panel having excellent luminance, sharpness, and formability, and a method for manufacturing the same. Description of the Related Art Conventionally, a radiation image such as an X-ray image has been widely used for diagnosis of a disease at a medical site. Particularly, an intensifying paper-film type radiation image has enhanced sensitivity and image quality thereof in the long history. As a result, the intensifying paper-film type radiation image is still now used widely at a medical site in the world as an imaging system having both high reliability and excellent cost performance. However, this image information is so-called analog image information, and cannot perform image processing freely or cannot perform electrical transmission instantaneously unlike digital image information which is developing now. As one of digital technologies on an X-ray image, computed radiography (CR) is now accepted at a medical site. However, an X-ray image obtained by CR has insufficient sharpness and insufficient spatial resolution compared to an image obtained by a screen film system such as a silver salt photography method. The image level of CR has not reached that of the screen film system. Therefore, as a new digital X-ray image technology, for example, a flat panel X-ray detector (FPD) using a thin film transistor (TFT) has been developed. In order to convert an X-ray into visible light, the above FPD principally uses a scintillator panel including a scintillator layer formed with an X-ray phosphor which converts an irradiation X-ray into visible light to emit light. However, in X-ray imaging using an X-ray source having a low dose, in order to increase a ratio (SN ratio) between a signal and a noise detected by the scintillator panel, it is necessary to use a scintillator panel having a high luminous efficiency (conversion ratio of an X-ray into visible light). In general, the luminous efficiency of a scintillator panel depends on the thickness of a scintillator layer and an X-ray absorption coefficient of a phosphor. The thicker the scintillator layer is, the more easily the light emitted by X-ray irradiation in the scintillator layer is scattered. An excessively thick scintillator layer deteriorates sharpness of an X-ray image obtained via the scintillator panel disadvantageously. Therefore, when sharpness required for an image is determined, the film thickness is determined automatically. Therefore, a scintillator plate which has an excellent luminous efficiency, that is, has both excellent luminance and excellent sharpness (MTF), and can form a high image quality, has been desired. JP 2007-292583 A discloses a scintillator plate using at least one kind selected from gadolinium oxide containing an activation material and gadolinium oxysulfide containing an activation material as a phosphor. Each of the gadolinium oxide and the gadolinium oxysulfide is a mixture of particles having different average particle diameters. However, the scintillator plate described in JP 2007-292583 A requires further improvement in emission luminance and sharpness. JP 5340444 B1 discloses a radiation image detector including a wavelength conversion layer having a first phosphor layer and a second phosphor layer in such an order that the spatial filling ratio of the phosphor particles increases on aside of the detector in order to improve sharpness. The first phosphor layer and the second phosphor layer each have phosphor particles dispersed in a binder. The average particle diameter of the phosphor particles in the second phosphor layer is smaller than that of the phosphor particles in the first phosphor layer. JP 2013-217913 A discloses, a radiation image detector including a wavelength-converting layer having a monolayer phosphor layer in which first phosphor particles having a first average particle diameter and second phosphor particles having a second average particle diameter are mixed in a binder in order to improve sharpness. The second average particle diameter is smaller than that of the first average particle diameter. The weight of the phosphor particle is gradually decreased as the distance from the solid detector is increased. However, the technology disclosed in JP 5340444 B1 or JP 2013-217913 A distributes phosphor particles in a phosphor layer nonuniformly in a direction perpendicular to a surface of a support, requires a very high ability for controlling a process in order to control a dispersion state of the phosphor with a fixed order, and is not necessarily suitable for industrial mass production. Because of the nonuniform distribution of the phosphor particles in the phosphor layer, it cannot be said that light emitted by a phosphor particle existing on the opposite side to a sensor panel can be received efficiently. In such a situation, appearance of a new scintillator panel which has high luminance and sharpness and does not require complicated management of a process is desired strongly. An object of the present invention is to provide a scintillator panel having excellent luminance, sharpness, and formability, and a method for manufacturing the same. To achieve the abovementioned object, according to an aspect, a scintillator panel reflecting one aspect of the present invention comprises a support and a scintillator layer, wherein the scintillator layer includes scintillator particles, a binder resin, and a void, and the porosity of the scintillator layer is from 14 to 35% by volume. According to the scintillator panel, when the scintillator layer is divided equally into two layers parallel to a plane of the support, a difference in the porosity between the layers is preferably 5% by volume or less. According to the scintillator panel, when the scintillator layer is divided equally into three to five layers parallel to a plane of the support, a variance in the porosity between the layers is preferably 5% by volume or less. According to the scintillator panel, the diameter of a circumscribed sphere circumscribed to the void of the scintillator layer is preferably from 0.2 to 15 μm. According to the scintillator panel, at least a part of the void is preferably formed by introducing air bubbles into the scintillator layer. According to the scintillator panel, at least a part of the void is preferably formed by introducing hollow particles into the scintillator layer. According to the scintillator panel, the light transmittance of the binder resin in a wavelength range of 400 to 600 nm is preferably 80% or more. According to the scintillator panel, the refractive index of the binder resin is preferably from 1 to 2.2, and more preferably from 1 to 1.5. According to the scintillator panel, an area of the scintillator particles in contact with the void is preferably larger than that of the scintillator particles in contact with the binder resin in the scintillator layer. According to the scintillator panel, the refractive index of the binder resin is preferably from 3 to 12% by volume in the scintillator layer. According to the scintillator panel, the filling ratio of the scintillator particles is preferably from 55 to 73% by volume in the scintillator layer. According to the scintillator panel, the scintillator particles preferably include at least two kinds of scintillator particles having different average particle diameters, of a first scintillator particle having a first average particle diameter, and a second scintillator particle having a second average particle diameter, the average particle diameter of the first scintillator particle is preferably from 0.5 to 5 μm, the average particle diameter of the second scintillator particle is preferably from 7 to 20 μm, and a particle diameter ratio between the first scintillator particle and the second scintillator particle is preferably three or more. According to the scintillator panel, the film thickness of the scintillator layer is preferably 500 μm or less. According to the scintillator panel, at least a part of the scintillator layer is preferably covered with a protective layer. According to the scintillator panel, the scintillator particle preferably includes a component having a melting point of 800° C. or higher as a main component. According to the scintillator panel, the scintillator particle preferably includes gadolinium oxysulfide as a main component. According to the scintillator panel, a light reflection layer which reflects 80% or more of light in a wavelength region of 400 to 600 nm is preferably provided between the support and the scintillator layer. According to the scintillator panel, a protective layer having humidity resistance is preferably provided on the opposite side of the scintillator layer to the side on which the support is provided. To achieve the abovementioned object, according to an aspect, a method for manufacturing a scintillator panel reflecting one aspect of the present invention comprises: preparing a coating liquid for a phosphor layer including scintillator particles, a binder resin, and a void-forming component; and forming a scintillator layer having a porosity of 14 to 35% by volume by applying the coating liquid for a phosphor layer on a support. According to the method for manufacturing a scintillator panel, the void-forming component is preferably at least one selected from a volatile solvent, air bubbles, and inert gas, or the void-forming component is preferably a hollow particle. Hereinafter, an embodiment of the present invention will be described with reference to the drawings. However, the scope of the invention is not limited to the illustrated examples. [Scintillator Panel] A scintillator panel according to an aspect of the present invention includes a support and a scintillator layer. The scintillator panel may further include at least one selected from a light reflection layer and a protective layer, if necessary. <Support> In the present invention, the support means a member playing a dominant role in order to hold the scintillator layer in components of the scintillator panel. Examples of a material of the support used in the present invention include various kinds of glass, a polymer material, and metal which can transmit radiation such as an X-ray. More specific examples thereof include plate glass such as quartz, borosilicate glass, or chemically reinforced glass; ceramic such as sapphire, silicon nitride, or silicon carbide; a semiconductor such as silicon, germanium, gallium arsenide, gallium phosphide, or gallium nitride; a polymer film (plastic film) such as a cellulose acetate film, a polyester resin film, a polyethylene terephthalate film, a polyamide film, a polyimide film, a triacetate film, a polycarbonate film, or a carbon fiber-reinforced resin sheet; a metal sheet such as an aluminum sheet, an iron sheet, or a copper sheet; a metal sheet having a cover layer of an oxide of the metal; and a bionanofiber film. The material of the support may be used singly or in combination of two or more kinds thereof. Among the materials of the support, a flexible polymer film is particularly preferable. The thickness of the support depends on the thickness of a scintillator panel used, but is preferably from 100 to 1000 μm, and more preferably from 100 to 500 μm in terms of handling. The support may include a light-shielding layer and/or a light-absorbing pigment layer, for example, in order to adjust a reflectivity thereof in addition to a layer formed of the above materials. The support may have a light-absorbing property and/or a light-reflecting property or may be colored, for example, in order to adjust the reflectivity thereof. <Scintillator Layer> The scintillator layer used in the present invention includes scintillator particles, a binder resin, and a void. <Scintillator Particle> As the scintillator particle according to an aspect of the present invention, it is possible to appropriately use a substance which can convert radiation such as an X-ray into light having a different wavelength such as visible light. Specifically, a scintillator and a phosphor described at pp. 284 to 299 of “Phosphor Handbook” (edited by Phosphor Research Society, Ohmsha, Ltd., 1987) and a substance described in “Scintillation Properties (http://scintillator.lbl.gov/)” (Web homepage of U.S. Lawrence Berkeley National Laboratory) can be used. However, even a substance not described here can be used as a scintillator particle as long as the substance “can convert radiation such as an X-ray into light having a different wavelength such as visible light”. Specific examples of a composition of the scintillator particle include the following. First, examples thereof include a metal halide phosphor represented by a basic composition formula (I): MIX.aMIIX′2.bMIIIX″3:zA. In the above basic composition formula (I), MI represents an element which can become a monovalent cation, that is, at least one selected from the group consisting of lithium (Li), sodium (Na), potassium (K), rubidium (Rb), cesium (Cs), thallium (Tl), silver (Ag), and the like. MII represents an element which can become a divalent cation, that is, at least one selected from the group consisting of beryllium (Be), magnesium (Mg), calcium (Ca), strontium (Sr), barium (Ba), nickel (Ni), copper (Cu), zinc (Zn), cadmium (Cd), and the like. MIII represents at least one selected from the group consisting of scandium (Sc), yttrium (Y), aluminum (Al), gallium (Ga), indium (In), and elements belonging to lanthanoid. X, X′, and X″ each represent a halogen element, and may represent different elements or the same element. A represents at least one element selected from the group consisting of Y, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Na, Mg, Cu, Ag (silver), Tl, and Bi (bismuth). a, b, and z independently represent values within ranges of 0≦a<0.5, 0≦b<0.5, and 0<z<1.0, respectively. Examples of the composition of the scintillator particle include a rare earth activated metal fluorohalide phosphor represented by a basic composition formula (II): MIIFX:zLn. In the above basic composition formula (II), MII represents at least one alkaline earth metal element, Ln represents at least one element belonging to lanthanoid, and X represents at least one halogen element. z represents a value within a range of 0<z≦0.2. Examples of the composition of the scintillator particle include a rare earth oxysulfide phosphor represented by a basic composition formula (III): Ln2O2S:zA. In the above basic composition formula (III), Ln represents at least one element belonging to lanthanoid, and A represents at least one element selected from the group consisting of Y, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Na, Mg, Cu, Ag (silver), Tl, and Bi (bismuth). z represents a value within a range of 0<z<1. Particularly, Gd2O2S using gadolinium (Gd) as Ln is preferable because it is known that by using terbium (Tb), dysprosium (Dy), or the like as an element of A, Gd2O2S exhibits high luminous characteristics in a wavelength region in which a sensor panel receives light most easily. Examples of the composition of the scintillator particle include a metal sulfide phosphor represented by a basic composition formula (IV): MIIS:zA. In the above basic composition formula (IV), MII represents an element which can become a divalent cation, that is, at least one element selected from the group consisting of an alkaline earth metal, zinc (Zn), strontium (Sr), gallium (Ga), and the like, and A represents at least one element selected from the group consisting of Y, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Na, Mg, Cu, Ag (silver), Tl, and Bi (bismuth). z represents a value within a range of 0<z<1. Examples of the composition of the scintillator particle include a metal oxoate phosphor represented by a basic composition formula (V): MIIa(AG)b:zA. In the above basic composition formula (V), MII represents a metal element which can become a cation, (AG) represents at least one oxo acid group selected from the group consisting of a phosphate, a borate, a silicate, a sulfate, a tungstate, and an aluminate, and A represents at least one element selected from the group consisting of Y, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Na, Mg, Cu, Ag (silver), Tl, and Bi (bismuth). a and b represent any value which can be according to a valence of a metal or an oxo acid group. z represents a value within a range of 0<z<1. Examples of the composition of the scintillator particle include a metal oxide phosphor represented by a basic composition formula (VI): MaOb:zA. In the above basic composition formula (VI), M represents at least one element selected from metal elements which can become cations. A represents at least one element selected from the group consisting of Y, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Na, Mg, Cu, Ag (silver), Tl, and Bi (bismuth). a and b represent any value which can be according to a valence of a metal or an oxo acid group. z represents a value within a range of 0<z<1. Examples of the composition of the scintillator particle include a metal acid halide phosphor represented by a basic composition formula (VII): LnOX:zA. In the above basic composition formula (VII), Ln represents at least one element belonging to lanthanoid, X represents at least one halogen element, and A represents at least one element selected from the group consisting of Y, Ce, Pr, Nd, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Na, Mg, Cu, Ag (silver), Tl, and Bi (bismuth). z represents a value within a range of 0<z<1. As the scintillator particle, it is possible to appropriately use a substance which can convert radiation such as an X-ray into light having a different wavelength such as visible light. However, it is particularly preferable to use a substance having a melting point of a main component of 800° C. or higher. Examples of a substance having a melting point of 800° C. or higher include gadolinium oxysulfide. The reason is as follows. That is, a substance having a melting point of 800° C. or higher has an excellent handling property. In addition, it is difficult due to suppression of evaporation of a scintillator raw material caused by the high melting point to use a thermal PVD method (physical vapor deposition method with heat) known as a method for manufacturing a scintillator. Therefore, it is easy to use a method for manufacturing a scintillator layer by preparing a coating liquid for a phosphor layer and applying the coating liquid on a support. Here, the main component means a component of 50% by mass or more in 100% by mass of the components constituting the scintillator particle. The scintillator particles preferably include at least two kinds of scintillator particles having different average particle diameters, of a first scintillator particle having a first average particle diameter and a second scintillator particle having a second average particle diameter. By using at least two kinds of scintillator particles having different average particle diameters, it is possible to increase a filling ratio of scintillator particles in the scintillator layer. The average particle diameter of the first scintillator particle is preferably from 0.5 to 5 μm, and more preferably from 0.5 to 3 μm. The average particle diameter of the second scintillator particle is preferably from 7 to 20 μm, and more preferably from 12 to 20 μm. A particle diameter ratio between the first scintillator particle and the second scintillator particle is preferably three or more, and more preferably six or more. Here, the particle diameter ratio means “average particle diameter of the second scintillator/average particle diameter of the first scintillator particle”. The filling ratio of the scintillator particles in the scintillator layer is preferably from 55 to 73% by volume, and more preferably from 58 to 70% by volume. A filling ratio of the scintillator particles in the scintillator layer, lower than the lower limit value of the above range, is not preferable because it is possible to obtain only an emission amount to a degree that is not suitable for practical use. A filling ratio higher than the upper limit value of the above range is not preferable because the scintillator layer is not suitable for forming a coating film due to reduction in fluidity of a mixture with a binder resin, and at the same time, an emission amount is reduced due to not being capable of extracting emission well at a location far away from a light-receiving side. [Binder Resin] The binder resin is not particularly limited as long as the object of the present invention is not impaired, and may be a commercially available binder resin obtained appropriately or a binder resin manufactured appropriately. Examples of the binder resin include a natural polymer such as protein (for example, gelatin), a polysaccharide (for example, dextran), or gum arabic; and a synthetic polymer such as polyvinyl butylal, polyvinyl acetate, nitrocellulose, ethylcellulose, a vinylidene chloride-vinyl chloride copolymer, polyalkyl (meth)acrylate, a vinyl chloride-vinyl acetate copolymer, polyurethane, cellulose acetate butyrate, polyvinyl alcohol, linear polyester, or an epoxy resin. The binder resin may be used singly or in combination of two or more kinds thereof. Among these binder resins, nitrocellulose, linear polyester, poly (meth)acrylate, polyvinyl butylal, a mixture of nitrocellulose and linear polyester, a mixture of nitrocellulose and poly (meth)acrylate, polyurethane, and a mixture of polyurethane and polyvinyl butylal are preferable from a viewpoint of transparency (light transmittance). These binder resins may be crosslinked by a crosslinking agent. The binder resin preferably includes an epoxy resin as an yellowing preventing agent. In general, 0.01 to 1 part by mass of the binder resin is used with respect to one part by mass of the scintillator particles. However, a smaller amount of the binding agent is more preferable in terms of sensitivity and sharpness of a scintillator plate obtained. 0.03 to 0.2 parts by mass of the binder resin is more preferable due to balance with easiness of applying. The volume ratio of the binder resin is preferably from 0.01 to 0.5, and more preferably from 0.03 to 0.3 with respect to the scintillator particles. The filling ratio of the binder resin in the scintillator layer is preferably from 3 to 12% by volume, and more preferably from 3 to 10% by volume. When the filling ratio of the binder resin in the scintillator layer is equal to or more than the lower limit value of the above range, it is possible to form a coating film easily. When the filling ratio is equal to or less than the upper limit value of the above range, it is possible to reduce emission loss in the scintillator layer. The present invention uses a resin having a light transmittance in a wavelength range of 400 to 600 nm, usually of 80% or more and more preferably of 83% or more. When the light transmittance in a wavelength range of 400 to 600 nm is within the above range, for example, it is possible to reduce emission loss caused by reflection on the air layer or attenuation when scintillator particles including gadolinium oxysulfide are used. The refractive index of the binder resin is usually from 1 to 2.2, and preferably from 1 to 1.5. The refractive index within the above range can suppress refraction between the binder resin and the void, and can suppress scattering of emission passing through the binder resin and the void. [Void] In the scintillator panel according to an aspect of the present invention, the porosity of the void in the scintillator layer is from 14 to 35% by volume, and preferably from 20 to 30% by volume. The porosity means a volume ratio of the void in the scintillator layer. By providing the void in the scintillator layer, it is possible to detect emitted light by a scintillator particle located far away from the sensor panel without any loss. As a result, even when the scintillator layer is thin, the scintillator layer has sufficient luminance. In addition, a thinner scintillator layer improves sharpness and an image quality advantageously. A method for providing a void in a scintillator layer is not particularly limited, but can be selected appropriately according to an object. However, for example, when a scintillator layer is formed from a coating liquid for a phosphor layer including scintillator particles, a binder resin, a solvent, or the like, if necessary, examples of the method include (1) a method of using a volatile solvent for the coating liquid for a phosphor layer and vaporizing the solvent, (2) a method of mechanically stirring the coating liquid for a phosphor layer to generate air bubbles, (3) a method of introducing inert gas into the coating liquid for a phosphor layer, (4) a method of adding a foaming agent to the coating liquid for a phosphor layer, (5) a method of adding hollow particles to the coating liquid for a phosphor layer, and (6) a method of adding a component to be subjected to a chemical reaction to generate gas to the coating liquid for a phosphor layer. In the above method of (1), examples of the volatile solvent include benzene, chloroform, diethyl ether, ethyl acetate, acetone, methyl ethyl ketone, methyl isobutyl ketone, ethanol, toluene, and cyclohexanone, described in [Method for manufacturing scintillator layer]. In the above method of (2), examples of the method of mechanically stirring the coating liquid to generate air bubbles include a method of introducing the air into a liquid in a form of bubbles by stirring the liquid with a stirrer, a whisk, or the like. In the above method of (3), as the inert gas, a substance in a gas or liquid state at the time of mixing is used. Examples thereof include nitrogen gas, argon, helium, and carbon dioxide gas. It is possible to appropriately change a rate of introducing an inert gas according to the kind, the amount, and the like of the coating liquid for a phosphor layer. In the above method of (4), the foaming agent can be selected appropriately from known foaming agents according to an object. However, preferable examples thereof include a carbon dioxide gas-generating compound, a nitrogen gas-generating compound, an oxygen gas-generating compound, and a microcapsule type foaming agent. Examples of the carbon dioxide gas-generating compound include a bicarbonate such as sodium bicarbonate. Examples of the nitrogen gas-generating compound include a mixture of NaNO2 and NH4Cl; an azo compound such as azobisisobutyronitrile or diazoaminobenzene; and a diazonium salt such as p-diazodimethylaniline chloride zinc chloride, morpholino benzene diazonium chloride zinc chloride, morpholino benzene diazonium chloride, fluoroborate, p-diazo ethyl aniline chloride zinc chloride, 4-(p-methyl benzoylamino)-2,5-diethoxy benzene diazonium zinc chloride, or 1,2-diazonaphthol 5-sodium sulfonate. Examples of the oxygen gas-generating compound include a peroxide. Examples of the microcapsule type foaming agent include a foaming agent of microcapsule particles encapsulating a substance (may be in a liquid state or a solid state at normal temperature) having a low boiling point, vaporized at a low temperature. Examples of the microcapsule type foaming agent include a microcapsule type foaming agent having a diameter of 10 to 20 μm, obtained by encapsulating a volatile substance having a low boiling point, such as propane, butane, neopentane, neohexane, isopentane, or isobutylene, into a microcapsule wall material made of polystyrene, polyvinyl chloride, polyvinylidene chloride, polyvinyl acetate, poly(meth)acrylate, poly(meth)acrylonitrile, polybutadiene, or a copolymer thereof. In the above method of (5), the hollow particle is not particularly limited as long as the hollow particle includes a void. Examples thereof include a single hollow particle having one hollow part in the particle, a multi hollow particle having many hollow parts in the particle, and a porous particle. These particles can be selected appropriately according to an object. Among these hollow particles, the single hollow particle and the multi hollow particle, in which the void is not filled with a binder resin or the like, are preferable. Here, the hollow particle means a particle having a void such as a hollow part or a pore. The “hollow part” means a hole (air layer) in a particle. The multi hollow particle means a particle having a plurality of holes in a particle. The porous particle means a particle having a pore. The pore means a part recessed from the surface of a particle toward the inside of the particle. Examples of a shape of the pore include a cavity shape, a needle shape, a shape recessed toward the inside or the center of a particle, such as a curve shape, and a shape in which these shapes pass through the particle. The size or volume of the pore may be large or small, and is not particularly limited. A material of the hollow particle is not particularly limited, but can be selected appropriately according to an object. However, examples thereof include a wall material of the above microcapsule type foaming agent. Preferable examples thereof include a thermoplastic resin such as a styrene-(meth)acrylate copolymer. The hollow particle may be manufactured appropriately or may be a commercially available one. Examples of the commercially available one include Ropaque HP1055 and Ropaque HP433J (manufactured by Zeon Corporation) and SX866 (manufactured by JSR Corporation). Preferable examples of the multi hollow particle include Sylosphere (registered trademark) and Sylophobic (registered trademark) manufactured by Fuji Silysia Chemical Ltd. Among these hollow particles, a single hollow particle is particularly preferable in terms of a magnitude of the porosity. Examples of the method of (6) include a method of allowing diisocyanate to react with a polyol which is a reaction type liquid for forming. This is a method utilizing gas generated in a reaction of generating a polymer. Examples of a component to be reacted for forming include a combination of a polyether polyol, a polyester polyol, or the like and an aromatic diisocyanate, an aliphatic diisocyanate, or the like. In the present invention, it is preferable to form at least a part of the void by introducing air bubbles into the scintillator layer. Specific examples of introducing air bubbles include the methods of (1) to (3). As in the method of (5), it is also preferable to format least a part of the void by mixing hollow particles into the scintillator layer. A void-forming component means a component used for forming a void. For example, a volatile solvent corresponds thereto in the method of (1), air bubbles correspond thereto in the method of (2), an inert gas corresponds thereto in the method of (3), a foaming agent corresponds thereto in the method of (4), a hollow particle corresponds thereto in the method of (5), and a component to be subjected to a chemical reaction to generate gas corresponds thereto in the method of (6). [Method for Manufacturing Scintillator Layer] A method for manufacturing a scintillator layer preferably includes a step of adding scintillator particles and a binder resin to a proper solvent and mixing these sufficiently to prepare a coating liquid for a phosphor layer having the scintillator particles and the binder resin scattered uniformly. Examples of the solvent used in preparing the coating liquid for a phosphor layer include a lower alcohol such as methanol, ethanol, isopropanol, or n-butanol; a ketone such as acetone, methyl ethyl ketone, methyl isobutyl ketone, or cyclohexanone; an ester of a lower alcohol and a lower aliphatic acid such as methyl acetate, ethyl acetate, or n-butyl acetate; an ether such as dioxane, ethylene glycol monoethyl ether, or ethylene glycol monomethyl ether; an aromatic compound such as triol or xylol; a halogenated hydrocarbon such as methylene chloride or ethylene chloride; and a mixture thereof. The coating liquid for a phosphor layer may include various additives such as a dispersant for improving dispersiveness of a phosphor in the coating liquid or a plasticizer for improving a bonding force between the binder resin and the scintillator particles in a phosphor layer formed. Examples of the dispersant used for such an object include phthalic acid, stearic acid, caproic acid, and a hydrophilic surfactant. Examples of the plasticizer include a phosphate such as triphenyl phosphate, tricresyl phosphate, or diphenyl phosphate; a phthalate such as diethyl phthalate or dimethoxyethyl phthalate; a glycolate such as ethylphthalylethyl glycolate or butylphthalylbutyl glycolate; and a polyester of a polyethylene glycol and an aliphatic dibasic acid, such as a polyester of triethylene glycol and adipic acid or a polyester of diethylene glycol and succinic acid. A method for providing a void in the scintillator layer is described in [Void]. For example, a coating film of a coating liquid is formed by applying the coating liquid for a phosphor layer prepared as described above on a surface of the support uniformly. This applying operation is performed using a normal applying unit such as a doctor blade, a roll coater, or a knife coater such that the porosity is from 14 to 35% by volume after the scintillator layer is formed. Subsequently, the coating film formed is heated gradually and is thereby dried to complete formation of the scintillator layer. The scintillator layer may be formed of one layer or two or more layers. The film thickness of the scintillator layer depends on the characteristics of an aimed scintillator plate, but is usually 500 μm or less, and preferably from 150 to 300 μm. The film thickness within the above range makes it possible to obtain a scintillator layer having excellent luminance and sharpness. When the scintillator layer in the present invention is divided equally into two layers parallel to a plane of the support, a difference in the porosity between the layers is preferably 5% by volume or less. When the scintillator layer is divided equally into three to five layers parallel to a plane of the support, a variance in the porosity between the layers is preferably 5% by volume or less. When the difference in the porosity between the layers is within the above range, it is also possible to extract emission by a scintillator particle farthest from the sensor panel. The diameter of a circumscribed sphere circumscribed to the void of the scintillator layer is usually from 0.2 to 15 μm, and preferably from 0.2 to 13 μm. The diameter of the circumscribed sphere circumscribed to the void of the scintillator layer can be measured with a scanning electron microscope. When the diameter of the circumscribed sphere circumscribed to the void of the scintillator layer is within the above range, a path for light transmission is proper, a scattering amount of the emitted light parallel to the film thickness of the phosphor layer can be suppressed, and a path in the film thickness direction can be obtained sufficiently. Therefore, a sufficient emission amount can be held. In the scintillator layer in the present invention, an area of the scintillator particles in contact with the void is preferably larger than that of the scintillator particles in contact with the binder resin. The areas can be measured with a scanning electron microscope. When the area of the scintillator particles in contact with the void is larger than that of the scintillator particles in contact with the binder resin, attenuation of emission does not occur easily. <Light Reflection Layer> The scintillator panel according to an aspect of the present invention may include a light reflection layer between the support and the scintillator layer. The light reflection layer may be formed of one layer or two or more layers. By providing a light reflection layer, it is possible to extract emission by a phosphor very efficiently to improve luminance. The light reflection layer reflects preferably 80% or more, more preferably 85% or more, and still more preferably 90% or more of light with a wavelength of 400 to 600 nm. The surface reflectivity of the light reflection layer is preferably 80% or more, more preferably 85% or more, and still more preferably 90% or more. The surface reflectivity is a value calculated from a spectral reflectivity in a range of 300 to 700 nm based on JIS Z-8722. Unless a reflection wavelength is particularly specified, the reflectivity means a reflectivity at a wavelength of 550 nm. Examples of the light reflection layer include a reflection layer (1) containing a metal and a reflection layer (2) containing light-scattering particles and a binder. The reflection layer (1) containing a metal preferably contains a metal material such as aluminum, silver, platinum, palladium, gold, copper, iron, nickel, chromium, cobalt, or stainless steel as constitutional materials thereof. Among these materials, the reflection layer (1) particularly preferably contains aluminum or silver as a main component from a viewpoint of reflectivity or corrosion resistance. Two or more layers of such a metal thin film may be formed. Examples of a method for covering the support with a metal include deposition, sputtering, and sticking a metal foil without any particularly limitation. However, sputtering is most preferable from a viewpoint of adhesion. The thickness of the reflection layer (1) is preferably from 0.005 to 0.3 μm, and more preferably from 0.01 to 0.2 μm from a viewpoint of an efficiency of extracting the emitted light. In the present invention, the light reflection layer includes at least light-scattering particles and a binder, and may be the reflection layer (2) applied on the support. Examples thereof include a reflection layer described in JP 2014-17404 A. Examples of the light-scattering particles include a white pigment such as TiO2 (anatase type or rutile type), MgO, PbCO3.Pb(OH)2, BaSO4, Al2O3, M(II)FX (M(II): at least one atom selected from Ba, Sr, and Ca, X: Cl atom or Br atom), CaCO3, ZnO, Sb2O3, SiO2, ZrO2, lithopone (BaSO4.ZnS), magnesium silicate, basic silisulfate, basic lead phosphate, or aluminum silicate. These white pigments have a high covering power and a high refractive index, and therefore can scatter emission of a scintillator easily through reflection or refraction of light and can enhance sensitivity of a radiation image conversion panel obtained. Other examples of the light-scattering particle include a glass bead, a resin bead, a hollow particle having a hollow part in the particle, a multi hollow particle having many hollow parts in the particle, and a porous particle. These substances may be used singly or in combination of two or more kinds thereof. The film thickness of the reflection layer (2) is preferably from 10 to 500 μm. When the film thickness of the reflection layer (2) is less than 10 μm, sufficient luminance is not necessarily obtained. When the film thickness is more than 500 μm, smoothness of the surface of the reflection layer (2) may be deteriorated. The reflection layer (2) includes preferably 40 to 95% by mass of titanium oxide, and particularly preferably 60 to 90% by mass of titanium oxide. When the content is less than 40% by mass, luminance may be reduced. When the content is more than 95% by mass, adhesion to the support or the phosphor may be reduced. <Protective Layer> The scintillator panel according to an aspect of the present invention may be provided with a protective layer for protecting the phosphor layer physically or chemically, if necessary. In this case, it is preferable to cover at least a part of the scintillator layer with a protective layer, and it is more preferable to cover the entire surface of the scintillator layer on the opposite side to the support with a continuous protective layer. The protective layer preferably has humidity resistance. The protective layer may be formed of a single material, a mixed material, or a plurality of films formed of different materials. Various transparent resins can be used for the protective layer. Specifically, the protective layer can be formed by laminating a transparent resin film made of polyethylene terephthalate, polyethylene, polyvinylidene chloride, polyamide, polyimide, or the like on the phosphor layer. Alternatively, the protective layer can be formed by preparing a protective layer coating liquid having a proper viscosity by dissolving a transparent resin such as a cellulose derivative, polyvinyl chloride, polyvinyl acetate, a vinyl chloride-vinyl acetate copolymer, polycarbonate, polyvinyl butylal, polymethyl methacrylate, polyvinyl formal, or polyurethane, applying the protective layer coating liquid on a scintillator, and drying the protective layer coating liquid. The thickness of the protective layer on the phosphor layer is preferably from 1 to 10 μm in terms of an influence on an image and scratch resistance. This protective layer intercepts a substance (for example, a halogen ion) emitted from the phosphor of the scintillator panel or the like, and prevents corrosion on a side of the sensor panel caused by contact between the scintillator layer and the sensor panel. The light transmittance of the protective layer is preferably 70% or more with respect to light of 550 nm considering a photoelectric conversion efficiency of the scintillator panel, a wavelength of emission by the phosphor (scintillator), or the like. However, it is difficult to industrially obtain a material (film or the like) having a light transmittance of 99% or more. Therefore, the light transmittance is preferably from 99% to 70% substantially. A moisture permeability of the protective layer measured under the conditions of 40° C. and 90% RH in conformity with JIS 20208 is preferably 50 g/m2·day or less, and more preferably 10 g/m2·day or less from a viewpoint of protection of the scintillator layer, deliquescence, or the like. However, it is difficult to industrially obtain a film having a moisture permeability of 0.01 g/m2·day or less. Therefore, the moisture permeability is preferably 0.01 g/m2·day or more and 50 g/m2·day or less, and more preferably 0.1 g/m2·day or more and 10 g/m2·day or less. [Method for Manufacturing Scintillator Panel] The scintillator panel according to an aspect of the present invention can be manufactured by a method of a scintillator layer, including preparing a coating liquid for a phosphor layer including scintillator particles, a binder resin, and a void-forming component and forming a scintillator layer having a porosity of 14 to 35% by volume by applying the coating liquid for a phosphor layer on a support. At this time, it is possible to form a void having a predetermined ratio by a method using a volatile solvent, air bubbles, inert gas, or the like as a void-forming component, a method using hollow particles as a void-forming component, and various methods described above. Details of the method for manufacturing a scintillator layer have been described above. In the scintillator panel according to an aspect of the present invention, after a light reflection layer is formed on a support, if necessary, a scintillator layer may be formed on a surface of the support on which the light reflection layer has been formed. After the scintillator layer is formed, a protective layer may be formed on a surface of the scintillator layer, not in contact with the support. Hereinafter, the present invention will be described more specifically based on Examples, but is not limited to these Examples. [Transmittance] The light transmittance in a wavelength range of 400 to 600 nm was determined using a spectrophotometer U-4100 manufactured by HITACHI, Ltd. [Refractive Index] A refractive index was measured using KPR-2000 manufactured by Shimadzu Corporation. [Viscosity] A viscosity was measured using a B-type viscometer (BLII: manufactured by Toki Sangyo Co., Ltd.) based on JIS Z 8803. [Average Particle Diameter] An average value based on a volume measured using a particle diameter distribution measuring apparatus LA-920 manufactured by HORIBA was used as an average particle diameter. [Film Thickness] The film thickness of a phosphor layer was measured using a film thickness meter SP-1100D manufactured by Toyo Corporation. As a binder resin, 10 parts by mass of a polyurethane resin (Pandex T5265 manufactured by DIC Corporation, light transmittance in a wavelength range of 400 to 600 nm: 85% or more, refractive index: 1.5) and 2 parts by mass of a yellowing preventing agent: epoxy resin (EP1001 manufactured by YUKA SHELL EPOXY KABUSHIKI KAISHA, light transmittance in a wavelength range of 400 to 600 nm: 85% or more, refractive index: 1.5) were added to methyl ethyl ketone (boiling point: 79.5° C.) as a solvent for dissolution, and were dispersed with a propeller mixer to prepare a coating liquid for forming a phosphor layer having a solid content of 77%. Here, the “solid content” means a total content of components obtained by removing the solvent for dissolution from the components of the coating liquid for forming a phosphor layer. Next, first phosphor particles having an average particle diameter of 15 μm and formed of Gd2O2S: Tb (refractive index: 2.2) and second phosphor particles having an average particle diameter of 1 μm and formed of Gd2O2S: Tb (refractive index: 2.2) were mixed such that a mass ratio thereof was 7:3 to prepare mixed phosphor particles. The coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 10:90, and were dispersed with a propeller mixer. Furthermore, in order to adjust the viscosity, methyl ethyl ketone was added thereto to prepare a phosphor coating liquid having a viscosity of 100 CP. As a support, a white polyethylene terephthalate film (PET film, thickness: 250 μm, Lumirror E20 manufactured by Toray Industries, Inc.) was used. The phosphor coating liquid was applied on the support using a doctor blade, and then was dried at 60° C. for 20 minutes to manufacture a phosphor sheet including a phosphor layer having a thickness of 250 μm. A phosphor sheet was manufactured in a similar manner to Example 1 except the following. That is, as a solvent for dissolution, a solvent obtained by mixing cyclohexanone (boiling point: 155.6° C.) and methyl ethyl ketone at a mass ratio of 4:6 was used in place of methyl ethyl ketone. The viscosity of the phosphor coating liquid was adjusted to 20 CP in place of 100 CP. The phosphor coating liquid was applied on a support using a doctor blade, and then was dried at 30° C. for 30 minutes in place of being dried at 60° C. for 20 minutes. A phosphor sheet was manufactured in a similar manner to Example 1 except that the viscosity of the phosphor coating liquid was adjusted to 40 CP in place of 100 CP. A phosphor sheet was manufactured in a similar manner to Example 1 except that the phosphor coating liquid was applied on the support using a doctor blade, and then was dried at 80° C. for 10 minutes in place of being dried at 60° C. for 20 minutes. A phosphor sheet was manufactured in a similar manner to Example 1 except that the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 6:94 in place of 10:90. A phosphor sheet was manufactured in a similar manner to Example 1 except that the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 12:88 in place of 10:90. A phosphor sheet was manufactured in a similar manner to Example 1 except that the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 20:80 in place of 10:90. The phosphor coating liquid manufactured in Example 1 was applied on the support using a doctor blade, and then was dried at 80° C. for 10 minutes to manufacture a phosphor sheet including a phosphor layer having a thickness of 120 μm. The phosphor coating liquid manufactured in Example 2 was further applied on the surface of the phosphor sheet including the phosphor layer using a doctor blade, and then was dried at 60° C. for 20 minutes to manufacture a phosphor sheet including a phosphor layer having a thickness of 250 μm. A phosphor sheet was manufactured in a similar manner to Example 1 except that the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 25:75 in place of 10:90. A phosphor sheet was manufactured in a similar manner to Example 1 except that the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 4:96 in place of 10:90. A phosphor sheet was manufactured in a similar manner to Example 1 except the following. That is, as a solvent for dissolution, a solvent obtained by mixing cyclohexanone and methyl ethyl ketone at a mass ratio of 4:6 was used in place of methyl ethyl ketone. The coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 16:84 in place of 10:90. The viscosity of the phosphor coating liquid was adjusted to 20 CP in place of 100 CP. The phosphor coating liquid was applied on the support using a doctor blade, and then was dried at 30° C. for 30 minutes in place of being dried at 60° C. for 20 minutes. A phosphor sheet was manufactured in a similar manner to Example 1 except that when the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed and dispersed, the coating liquid for forming a phosphor layer and the mixed phosphor particles were stirred with a propeller mixer for 10 minutes while nitrogen gas was introduced at a rate of 500 g/min, and a coating liquid having nitrogen gas dispersed in the coating liquid for forming a phosphor layer was prepared. A phosphor sheet was manufactured in a similar manner to Example 1 except the following. That is, the viscosity of the phosphor coating liquid was adjusted to 200 CP in place of 100 CP. When the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed and dispersed, the coating liquid for forming a phosphor layer and the mixed phosphor particles were stirred with a propeller mixer for 10 minutes while nitrogen gas was introduced at a rate of 500 g/min, and a coating liquid having nitrogen gas dispersed in the coating liquid for forming a phosphor layer was prepared. The phosphor coating liquid was applied on the support using a doctor blade, and then was dried at 80° C. for 10 minutes in place of being dried at 60° C. for 20 minutes. A phosphor sheet was manufactured in a similar manner to Example 1 except that the coating liquid for forming a phosphor layer and the mixed phosphor particles were mixed such that a volume ratio of the solid content of the coating liquid for forming a phosphor layer:the mixed phosphor particles was 2:98 in place of 10:90. [Evaluation] Physical Properties were Measured as Follows. [Phosphor Filling Ratio, Resin Filling Ratio, and Porosity] In each of the phosphor sheets manufactured in Examples 1 to 10 and Comparative Examples 1 and 2, a phosphor layer was peeled from a PET film. The total volume of the phosphor layer was measured. Subsequently, a resin component was dissolved, and the volume of remaining phosphor particles was measured. The phosphor filling ratio (% by volume) was calculated from the total volume of the phosphor layer and the volume of the phosphor particles. The resin filling ratio (% by volume) was calculated using the phosphor filling ratio based on the mixing ratio between the solid content of the coating liquid for forming a phosphor layer and the mixed phosphor particles at the time of preparation of the phosphor coating liquid. The porosity (% by volume) was determined by a relation of “porosity”=1−(“phosphor filling ratio”+“resin filling ratio”) using the phosphor filling ratio and the resin filling ratio. Results are shown in Table 1. [Variance of Void] In each of the phosphor layers of the phosphor sheets manufactured in Examples 1 to 10 and Comparative Examples 1 and 2, a cross section perpendicular to the support was observed using a microtome (manufactured by Leica Microsystems) and a scanning electron microscope (manufactured by Hitachi High-Technologies Corporation). Using an image of the cross section, porosities of regions obtained by equally dividing the phosphor layer into upper and down parts perpendicularly to the support were calculated by image processing, and a variance in the porosity (% by volume) between the upper and down parts was calculated. Similarly, using an image of the cross section, porosities of regions obtained by equally dividing the phosphor layer into three to five parts perpendicularly to the support were calculated by image processing, and a variance in the porosity (% by volume) between the layers was calculated. At this time, almost the same result as the variance in the porosity obtained by equal division into two parts was obtained. Results are shown in Table 1. [Diameter of Circumscribed Sphere] A cross section obtained by equally dividing each of the phosphor layers of the phosphor sheets manufactured in Examples 1 to 10 and Comparative Examples 1 and 2 into two layers parallel to a plane of the support using a microtome (manufactured by Leica Microsystems) was observed using a scanning electron microscope (manufactured by Hitachi High-Technologies Corporation), and the diameter of the circumscribed sphere circumscribed to the void was measured. Results are shown in Table 1. [Film Formability] After the phosphor coating liquid was applied on the support using a doctor blade, and dried, a case in which a phosphor layer was formed on the support and a film was formed was evaluated as AA, and a case in which a film was not formed was evaluated as DD. Results are shown in Table 1. [Relative Luminance] A flat panel display (FPD) was manufactured using each of the phosphor sheets manufactured in Examples 1 to 10 and Comparative Examples 1 and 2, was irradiated with an X-ray having a tube voltage of 80 kVp, and an average signal value of image data obtained was used as an emission amount. A relative luminance obtained by assuming the luminance of the scintillator sheet manufactured in Comparative Example 1 as 100% is shown in Table 1. [Total Evaluation] A case in which the film formability was AA and the relative luminance was more than 100% was evaluated as AA, and the other cases were evaluated as DD. Results are shown in Table 1. TABLE 1Example 1Example 2Example 3Example 4Example 5Example 6Example 7ManufacturingSolvent for dissolution*1MEKCYC/MEKMEKMEKMEKMEKMEKphosphorSolid content of coating10:9010:9010:9010:906:9412:8820:80sheetliquid for formingphosphor layer:Mixedphosphor particlesIntroductionViscosity of phosphor100 CP20 CP40 CP100 CP100 CP100 CP100 CPcoating liquidDrying60° C.30° C.60° C.80° C.60° C.60° C.60° C.20 min30 min20 min10 min20 min20 min20 minEvaluationPhosphor filling ratio64%77%70%60%68%62%62%Resin filling ratio 7% 9% 8% 7% 4% 8%16%Porosity29%14%22%33%28%30%23%Variance in porosity5% or less5% or less5% or less5% or less5% or less5% or less5% or lessDiameter of0.5 um to 130.5 um to 130.5 um to 130.5 um to 130.5 um to 130.5 um to 130.5 um to 13circumscribed sphereumumumumumumumArea ratio5235642(approximately)*2Film formabilityAAAAAAAAAAAAAA Relative luminance117%105% 118% 117% 119% 112% 105% Total evaluationAAAAAAAAAAAAAAComparativeComparativeComparativeComparativeExample 8Example 9Example 10Example 1Example 2Example 3Example 4ManufacturingSolvent for dissolution*1MEKMEKMEKCYC/MEKMEKMEKMEKphosphorSolid content of coating10:9025:754:9616:8410:9010:902:98sheetliquid for formingphosphor layer:Mixedphosphor particlesIntroductionN2N2Viscosity of phosphor100 CP100 CP100 CP20 CP100 CP200 CP100 CPcoating liquidDrying60° C.60° C.60° C.30° C.60° C.80° C.60° C.20 min20 min20 min30 min20 min10 min20 minEvaluationPhosphor filling ratio64%58%70%76%55%NotNotmeasurablemeasurableResin filling ratio 7%19% 3%14% 6%NotNotmeasurablemeasurablePorosity29%23%27%10%39%NotNotmeasurablemeasurableVariance in porosity10%5% or less5% or less5% or less5% or less——Diameter of0.5 um to 130.5 um to 130.5 um to 130.5 um to 130.5 um to 130.5 um to 250.5 um to 13circumscribed sphereumumumumumumumArea ratio41.581.56——(approximately)*2Film formabilityAAAAAAAAAADDDDRelative luminance102%101% 121% 100% 97%——Total evaluationAAAAAADDDDDDDD*1The solvent for dissolution represents MEK: methyl ethyl ketone or CYC: cyclohexanone.*2The area ratio represents a ratio of (an area of the scintillator particles in contact with the void)/(an area of the scintillator particles in contact with the binder resin). According to an embodiment of the present invention, it is possible to obtain a scintillator panel having excellent luminance, sharpness, and formability. Although the present invention has been described and illustrated in detail, it is clearly understood that the same is by way of illustrated and example only and is not to be taken by way of limitation, the scope of the present invention being interpreted by terms of the appended claims. |
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052710459 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT ______________________________________ Outline of Contents ______________________________________ I. Overview Description of Control Complex 12 II. Panel Overview 15 A. Alarm and Messages 15 B. Indicator 16 C. CRT 17 D. Controller 17 E. Display Formats 19 F. Display Integration 23 III. DIAS 25 A. Discreet Indicators 25 B. Validity Algorithm Summary 31 C. Alarm Processing and Display 35 1. Mode and Equipment Dependence 38 2. Subfunction Grouping 38 3. Shape and Color Coding 40 4. Alarms on CRT 41 5. Determining Alarm Conditions 42 6. Acknowledging Alarms 42 IV. DPS A. CRT 51 B. IPSO 58 V. Control Room Integration 67 VI. Panel Modularity 72 APPENDIX (Validity Algorithm) 76 ______________________________________ I. Overview Description of Control Complex FIG. 1 shows a control room complex in accordance with the preferred embodiment of the present invention. The heart of the main control room 10 is a master control console 12 which allows one person to operate the nuclear steam supply system from the hot standby to the full power condition. It should be appreciated that the control room, equipment and methods described herein, may be advantageously used with light water reactors, heavy water reactors, high temperature gas cooled reactors, liquid metal reactors and advanced passive light water reactors, but for present purposes, the description will proceed on the basis that the plant has a pressurized water NSSS. For such an NSSS, the master control console 12 typically has five panels, one each for the reactor coolant system (RCS) 14, the chemical volume and control system (CVCS) 16, the nuclear reactor core 18, the feed water and condenser system (FWCS) 20, and the turbine system 22. As will be described more fully below, the monitoring and control for each of these five plant systems, is accomplished at the respective panel in the master control console. Immediately overhead behind the core monitoring and control panel 18, is a large board or screen 24 for displaying the integrated process status overview (IPSO). Thus, the operator has five panels and the overhead IPSO board within easy view while sitting or standing in the center of the master control console 12. To the left of the master control console is the safety related console 26, typically including modules associated with the safety monitoring, engineered safeguard features, cooling water, and similar functions. To the right of the master control console is the auxiliary system console 28 containing modules associated with the secondary cycle, auxiliary power and diesel generator, the switch yard, and the heating and ventilation system. Preferably, the plant computer 30 and mass data storage devices 32 associated with the control room are located in distributed equipment rooms 31 to improve fire safety and sabotage protection. The control room complex 10 also has associated therewith, a shift supervisor's office 34, which has a complete view of the control room, an integrated technical support center (TSC) 36 and viewing gallery outside the control area, and other offices 38 in which paper work associated with the operation of the plant may be performed. Similarly, desk, tables, and the like 40 are located on the control room floor for convenient use by the operators. A remote shut-down room 42 (FIG. 2) is also available on site for post-accident monitoring purposes (PAM). FIG. 2 is a schematic of the information links between the plant components and sensors, which for present purposes are considered conventional, and the various panels in the main control room. It is evident from FIG. 2 that information flows in both directions through the dashed line 46 representing the nuclear steam supply system and turbo generating system boundary. NSSS status and sensor information 48 that is used in the plant protection system 50 and the PAMS 58, passes directly through the NSSS boundary 46. Control signals 52 from the power control system pass directly through the NSSS boundary. Other control system signals 60,62 from the engineered safeguard function component control system 56 and the normal process component control system 64, are interfaced through the NSSS boundary via remote multiplexors 6. Each of the plant protection system, ESF component control system, process component control system, power control system and PAMs, is linked to the main control room 42, to each other, to the data processing system (DPS) 70 and to the discrete indication and alarm system (DIAS) 72. FIG. 2 illustrates one significant aspect of the present invention, namely, the integration of monitoring, control and protection information, during both normal and accident conditions, so that the operator's task in determining an appropriate course of action is considerably simplified. The way in which this is accomplished will be described in the following sections. II. Panel Overview FIGS. 3(a) and 3(b) is schematics of a sit/stand panel such as the reactor coolant system panel 14 from the master control console 12 in accordance with one embodiment of the invention. FIGS. 3(c) and 3(d) show an alternative embodiment for stand up only. The substantially flat upper portion or wall 74 of the panel is vertically oriented and the substantially flat lower or desk portion 76 is substantially horizontal, with the monitoring and alarm interfaces carried by the upper portion, and the control interfaces carried on the lower portion. A. Alarm and Messages The alarm Functionality (see FIGS. 9, 15-18) includes alarm and message (A & M) interface 78 having a multiplicity of tiles 80 each having particular acronym or similar cue 81 associated therewith, whereby an alarm condition is indicated by the illumination of that tile and the generation of an accompanying audible signal. The operator is required to acknowledge the alarm by either pushing the tile or some other interface provided for that purpose. The number of tiles associated with a particular panel is dependent on the number of different alarm conditions that can arise with respect to the monitored system, e.g., the reactor coolant system. Typically, hundreds of such tiles are associated with each panel. The alarms are prioritized into three (3) alarm classes (Priority 1, Priority 2, and Priority 3, prompting immediate action, prompt action and cautionary awareness). This RCS panel alarms are equipment status and mode dependent (Normal RCS, Heatup/Cooldown, Cold Shutdown/Refueling and Post Trip). When a high priority alarm actuates coincidentally with a low priority alarm on the same parameter, the lower priority alarm is automatically cleared. On improving conditions, the higher priority alarm will flash and sound a reset tone. The operator will acknowledge that the higher priority alarm has cleared. If the lower priority alarm still exists, its alarm window or indicator will turn on in the acknowledged state after the operator acknowledges that the higher priority alarm has cleared. B. Indicator The second monitoring interface are the process variable indicators, for example reactor coolant hot and cold leg temperatures, pressurizer level and pressure and other RCS parameters. Discrete indicators 82 (see also FIGS. 7 and 8) provide an improved method of presenting the RCS panel parameters. Some RCS panel parameters require continuous validated display and trending on the master control console. Plant process and category 1 parameters like pressurizer level and RCS cold leg temperature fall into this category. Other RCS panel parameters are used less frequently. The discrete indicators 82 provide indication on parameters needed for operation when the Data Processing System (CRT information displays) is unavailable. These include Regulatory Guide 1.97 category 1 and 2 parameters, parameters associated with priority 1 or priority 2 alarms, other parameters needed for operation due to inaccessibility of local gages and parameters that the operator must view for surveillance when the Data Processing System is unavailable for a period of up to twenty-four (24) hours. These less frequently viewed parameters would be available on discrete indicators, with a menu available by operator selection. The menu would show alphanumeric listings of available data points. Lastly, parameters displayed on process controllers need not be available on discrete indicators. C. CRT Additionally, a CRT display 84 generates an image of the major vessels, pipes, pumps, valves and the like associated with, e.g., the reactor coolant system, and displays the alarms and values of the parameters which may be shown in bar, graph, trend line or other form on the other displays 78,82 (see FIGS. 4-6, 10, 12-14 and 19-23). From this CRT, the operator has access to all NSSS information. The information is presented in a three level structured hierarchy that is consistent with the operator's system visualization. FIG. 4 illustrates the NSS primary side page directory 84, which accesses all CRT pages related to the functions of the RCS panel. D. Controller In the control portion 76 of the panel 14, a plurality of discrete, on-off switches 86 are provided at the left, for example, each switch pattern being associated with a particular reactor cooling pump whose operating parameters are displayed immediately above it, and analog control interfaces which can be in the form of conventional dials or the like (not shown), or touch screen, discrete control as indicated at 88. Process controllers are provided on the RCS panel to provide the operator with the ability to automatically or manually control process control loops. The process controllers allow control of throttling or variable position devices (such as electro-pneumatic valves) from a single control panel device. Process controllers are used for closed loop control of the following RCS panel process variables: pressure level, pressurizer pressure, RCP Seal Injection Flow and RCP Seal Injection Temperature. Process controllers are designed for each specific control loop utilizing a consistent set of display and control features. In a conventional control room, each process control loop has its own control device, usually referred to as a MANUAL/AUTO Station. For example, the RCP Seal Injection Sub-System has five process control loops, a seal injection flow control loop for each of the four RCPs and a seal injection temperature control loop for the entire sub-system. These five control loops each have their own MANUAL/AUTO station which occupy a large amount of control panel space and make cross loop comparisons cumbersome. Although these five process loops are controlled independently, process variations in one controlled parameter affect the other four process parameters. Conventional MANUAL/AUTO stations make it difficult for the operator to simultaneously interact with the five MANUAL/AUTO stations. The RCS panel process controllers for similar processes (related by function or system) are operated from a single control station, called a process controller. This single control station saves panel space, accommodates convenient cross channel checking and allows easier control loop interaction for multiple related controls. Component control features (i.e., actuation of switches controls) provide the primary method by which the operator actuates equipment and systems on the RCS panel. The RCS panel has forty-three components controlled from momentary type switches. Each switch contains a red status indicator for active or open and a green status indicator for inactive or closed. Blue status indicator lights/switches are used to indicate and select automatic control or control via a process controller. In addition to color coding, the red switch is always located above the green switch to reinforce color distinction. Each switch generates an active control signal when depressed and is inactive when released. Each switch is backlit to indicate equipment status/position. E. Display Formats Process display formats use standard information placement for similar processes and equipment. Fluid system piping representations are where possible standardized, top to bottom, left to right, with avoidance of crossovers. Incoming and outgoing flow path connections are placed at the margins. Related data is grouped by task and analysis specifications for comparison, sequence of use, function, and frequency. Process representations/layout are based on the operator's process visualization to maximize the efficiency of his data gathering tasks. The operator's visualization of a system is often based on diagrams used with learning materials and plant design documentation associated with system descriptions. Graphic information is presented on display page formats to aid in rapid operator comprehension of processes. Graphic information includes the use of bar graphs, flow charts, trends, and other plots, (e.g., Temp. vs. Press.). Bar graphs are primarily used to represent flows, pressures and levels. Since level corresponds to a tank, the bar graph is placed with consistent spatial orientation with respect to the tank symbol. Level bar graphs are oriented vertically. Flow bar graphs when used are oriented horizontally. Bar graphs are also helpful for comparison of numeric quantities. Flowcharts are used when they aid in the operator's process visualization. Flowcharts are helpful for understanding control system, processes such as the Turbine Control System. Operator's learning materials for process control systems are frequently in a flowchart format, and thus a similar format on a display page is easy to comprehend. Trends are used on display page formats when task analysis indicates that the operator should be informed about parameter changes over time. Additionally, the operator is able to establish trends of any data base points in the plant computers data base. In some situations, task analysis may indicate that more than one trend is important to monitor process comparisons. In other situations such as heatup/cooldown curves, two parameters may be placed on the different ordinate axis of a graph. When more than one trend curve occupies the same coordinate axes, two ordinate vertical axes can be used for parameters that have different units. Scale labels are divisible by 1, 2, 5 or 10. Tick marks between scale labels are also divisible by 1, 2, 5 or 10. Trended information is typically presented on display pages with a scale of 30 minutes. However, the operator is able to adjust the scale to suit his needs. Logarithmic axes may be established using multiples of 10. If full range is less than 10, an intermediate range label is located to fall near the middle of the scale. Different colors are used for trends occupying the same coordinates. When, multiple curves use a common scale, the scale is gray and the curves are color coded. When multiple ordinate scales are used, they are color coded in correspondence to the curve. The colors used for trends will not include the alarm color or normal status color to avoid associating process parameter with normal or alarm conditions. Color is used to aid the operator in rapidly discriminating between different types of information. Since the benefits of color coding are more pronounced with fewer colors, coding on informational displays (i.e., IPSO, CRTs, alarm tiles) is limited to seven colors. In addition, color coded information has other representational characteristics to aid in discrimination of data and discrimination by color deficient observers. The following colors are used in the information display to represent the following types of information. The colors used have been carefully selected to yield satisfactory contrast for red-green deficient color observers. ______________________________________ Color Representation Characteristics ______________________________________ Black Background color. Green Component Off/Inactive, Valve Closed and Operable. Red Component On/Activated, Valve Open and Operable. Yellow Alarm Status-Good attention-getting color. Grey Text, labels, dividing lines, menu options, piping, inoperable and non-instrumented valves, graph grids, and other applications not covered by other coding conventions. Light Blue Process parameter values. White System's response to operator touch, e.g., menu selection until appropriate system response occurs. ______________________________________ Shape coding is used in the information system to aid the operator to identifying component type, operational status, and alarm status. Component shape coding is based on symbology studies which included shape coding questionnaires given to nuclear power plant personnel. FIGS. 5 and 6 show the shapes used to represent components in the control room. An attribute of shape, hollow/solid, is reflective of the status of the component. Hollow shape coding indicates that the component is active, whereas solid shape coding is used to represent inactive components. An example of shape coding for a pump and valve is described as follows. Pump--A hollow pump indicates that the pump has been activated by the operator to automatic control signal. A solid pump indicates that the pump has been deactivated by the operator or automatic control signal. PA0 Valve--A hollow valve indicates that the valve is fully open and a solid valve indicates that the valve is fully closed. A valve not fully open or closed has a mixed solid/hollow shape, i.e., left side solid/right ride hollow. PA0 Valve Open and Operable--Red Color Coding. PA0 Valve Closed and Operable--Green Color Coding. PA0 Non-Instrumented Valve--Grey Color Coding (Position is Operator Inputted). PA0 Valve Not Operable--Grey Color Coding with Alarm Coding. PA0 Loss of Indication--Grey Color Coding with Alarm Coding and mixed hollow/solid shape; PA0 The critical function information provided on the 1st level display page that is associated with the critical function. PA0 Information related to success path availability and performance of the success paths that can support that critical function. PA0 High level information presented using a mimic format with the critical function/success path related information. PA0 A time trend of the most representative critical function parameter. PA0 1. RCP 1A PA0 2. RCP 1B PA0 3. RCP 2A PA0 4. RCP 2B PA0 5. RCP SealBleed PA0 6. RCS PA0 7. .sup.T hot PA0 8. .sup.T cold PA0 9. Pressurizer Pressure PA0 10. Pressurizer Level PA0 1. When validation fails and a "FAULT SELECT" sensor is selected for the "process representation". PA0 2. When the "Valid" output does not correlate to the PAMI sensor(s). PA0 1. The "process representation" is always displayed on the applicable DIAS display and/or CRT page(s) where a single "process representation" is needed as opposed to multiple sensor values. Each plant process parameter is evaluated individually to determine the type of display required and location (DIAS and CRT or CRT only). PA0 2. The "process representation" is always a "valid" value unless there is a: PA0 3. The "process representation" is always used for alarm calculations and trending (where a single value is normally trended). This can be "valid", "fault select" or "operator select" data, depending on the results of the algorithm calculations as described below. PA0 4. Using a menu on DIAS or the CRT, the operator may view any of the values (A,B,C,D or calculated output) without changing the "process representation". PA0 5. A "Fault Select" value will be displayed automatically as the "process representation" when the validation algorithm is unable to yield "valid" data. The "fault select" value is the output of the sensor closest to the last "valid" signal at the time validation initially failed. On DIAS (if applicable), this information will be labeled "fault select". On the CRT(s) graphic pages, this information is preceded by an asterisk(*) to indicate suspect data. The "fault select" "process representation" is automatically returned to a "valid" process representation" when the validation algorithm is able to calculate "valid" data. PA0 6. An "operator select" sensor may be selected for the "process representation" only when there is a: PA0 The "operator select" "process representation" will replace the "valid" or "fault select" "process representation". On DIAS (if applicable), this information will be labeled "operator select". On the CRT(s), this information will be preceded by an asterisk(*) on graphic displays and labelled "operator select" in the data base. The "operator select" "process representation" is automatically replaced by the calculated "valid" signal when both the "Validation Fault" and the "PAMI Fault" clear. PA0 1. Conditions that may cause a trip in less than 10 minutes. PA0 2. Conditions that may cause major equipment damage. PA0 3. Personnel/Radiation hazard. PA0 4. Critical Safety Function violation. PA0 5. Immediate Technical Specification Action Required. PA0 6. First-Out Reactor/Turbine Trip. PA0 1. Conditions that may cause a trip in greater than 10 minutes. PA0 2. Technical specification action items that are not Priority 1. PA0 3. Possible equipment damage. PA0 1. Sensor deviations. PA0 2. Equipment status deviations. PA0 3. Equipment/process deviations not critical to operation. PA0 1. Normal operation PA0 2. Heatup/cooldown. PA0 3. Cold shutdown/refueling. PA0 4. Post-trip. PA0 1. Unacknowledged Alarm--If there is an unacknowledged alarm associated with an alarm tile, the alarm tile will flash at a fast rate (i.e., 4 times/sec using a 50/50 duty cycle as depicted by the long rays in FIG. 9). This condition takes precedence over all other alarm tile states for group alarms PA0 2. Cleared Alarm/Return to Normal (Reset Alarm)--When an alarm condition clears, the corresponding alarm tile flashes at a slow rate (i.e., 1 time/sec using a 50/50 duty cycle as depicted by the short rays in FIG. 9) until this condition has been acknowledged. This condition takes precedence over the remaining two states for grouped alarms. PA0 3. Alarm--If an alarm condition exists and alarm states 1 and 2 above do not exist, then the alarm tile is lit without flashing (as depicted by the absence of rays in FIG. 9). PA0 4. No Alarm--If there is no alarm condition associated with an annunciator tile, then the alarm tile is not lit (not depicted in FIG. 9). To indicate that the alarm tile's bulb is functioning, a lamp test tile is not 1 indicate that the feature is provided. PA0 A) First Level Display Page Set (Major Plant System/Function Groupings 142) PA0 B) Control Room Workstation 144 PA0 C) Alarm tiles 146 PA0 1) The operator selects the "Alarm List" menu option 140 (FIG. 4) followed by the "Elec." menu option 148 (FIG. 12). This accesses the categorized alarm listing of the type shown in FIG. 14 beginning with the electrical alarms. PA0 2) If the operator wishes to view alarms associated with a specific alarm, e.g., RCPIA he selects the following menu options from page 84 (FIGS. 4 and 12): PA0 A. Categorized Alarm List--The operator selects "Alarm List" followed by the tile, e.g., "RCPIA", menu option. The categorized alarm list is accessed with RCPIA alarms at the top of the page. PA0 B. Alarm Messages--The operator can use the alarm tile menu options in the same method that the control panel alarm tiles are used. The selection of an alarm tile menu option provides the alarm message and a menu with display pages that can provide supporting information about the alarm condition. PA0 1 ) Alarm acknowledgement via the annunciator tiles--Alarms can be acknowledged by depressing alarming/unacknowledged annunciator tiles or a CRT annunciator tile representation. This action changes the annunciator tile from a flashing condition to a solid condition when all alarm conditions associated with the tile have been acknowledged and silences any audible sound (described later) associated with the alarm condition. Alarm messages are viewed on the message window (when using the physical tile) and the workstation's CRT message line (see FIG. 16) . PA0 2 ) Alarm acknowledgement using alarm listing pages--Alarms can be acknowledged on the categorized listing by touching alarm tile touch targets associated with the alarm tile categories (see FIG. 14). Upon touching the alarm tile's representation, all alarms associated with that tile are acknowledged. This means of alarm acknowledgement may be the most useful for acknowledging multiple alarms remote to the operator's location. PA0 1. Unacknowledged Priority 1 or 2 Alarms. PA0 2. An Alarm Reminder Tone for Priority 1 or 2 Unacknowledged or Cleared Conditions. PA0 3. Cleared Priority 1 Alarms, or Cleared Priority 2 Alarms. PA0 All new/unacknowledged priority 2, 3 and operator aid features change from a fast flash rate to a steady highlighted condition, i.e., tiles and CRT alarm representations. PA0 Any cleared alarm conditions, i.e., slow flash rate, are not presented as alarm information. PA0 Any new alarm condition or cleared alarm condition coming in after the "STOP FLASH" button has been activated, is normally displayed to the operator (i.e., flashing). However, the operator may redepress the alarm "STOP FLASH" button to suppress these conditions. PA0 1) Primary Systems (example, see FIG. 19) PA0 2) Secondary Systems PA0 3) Power Conversion PA0 4) Electrical Systems PA0 5) Auxiliary Systems PA0 6) Critical Functions PA0 1) The next higher level (when applicable) display page in the hierarchy, item (c). This feature is more meaningful on a 3rd level display page since the next higher level page is a level 2 display page which is not normally on the menu. PA0 2) Display pages of systems that are connected to or support the process of the presently displayed page (h,i). PA0 3) All six first level display pages (b,c,d,e,f,g). PA0 4) The IPSO display page (a). PA0 5) The last page viewed on the monitor (j). PA0 (1) Display Page Access Using Alarm Tiles--This mechanism for display page access may be most useful for obtaining display pages associated with the workstation's process. By pressing a workstation alarm tile from display 78, such as 80 (FIG. 15), region 4 of the workstation CRT's display page menu changes to a new menu with display page options associated with the alarm tile's descriptor. For example as shown in FIG. 23, an RCP1A alarm tile provides menu options associated with RCP1A. The desired display page will then be a direct access menu option. PA0 (2) Accessing CRT Information from the Discrete Indicators--Each discrete indicator 82 such as shown in FIG. 7, has a CRT access touch target 158. This button provides for access to supporting information for the process parameter that is presently displayed on the discrete indicator. By touching the CRT target on the discrete indicator, region 4 of the menu options on the workstation's CRT changes to menu options containing display pages with supporting and diagnostic information associated with the process parameter. PA0 (3) Display Page Access Using a Display Page Directory--Any display page of the display page hierarchy can be accessed using the presently displayed menu. For example, if the operator is viewing the Feedwater System display page and wants to access the CVCS display page, the following sequence takes place (refer to FIGS. 22 and 4): The operator selects "by touch" the "DIRECTORY" menu option (option 1 in region 2 on FIG. 22) followed by the "PRIMARY" menu option (option b in region 3 on FIG. 22). This accesses the primary section of the display page hierarchy from the display page library (see FIG. 4). Each display page within the primary section of the display page hierarchy is a touch target on this display page, and now the operator can select the CVCS display page. Any page in the display page hierarchy can be accessed using this feature. The "DIRECTORY" menu option is followed by the desired hierarchy associated with one of the six first level display pages, menu options b,c,d,e,f or g on FIG. 22. PA0 Failure to satisfy the safety function status checks, (post-trip). PA0 Poor performance of a success path/system that is being used to support a critical function. PA0 An undesirable priority 1 deviation in a power production function (pre-trip). PA0 Unavailability of a safety system (less than minimum availability as defined by Reg. Guide 1.47). PA0 (a) Feedwater and Condensate System Status Information (i.e., operational status, alarm status) PA0 (b) Steam Generator Levels, Dynamic Representation PA0 (c) Steam Generator Safety Valve Status PA0 (d) Atmospheric Dump Valve Status PA0 (e) Main Steam Isolation Valve Status PA0 (f) Turbine Bypass System Status PA0 (a) Plant net electric output, digital value. PA0 (b) Alarm information for deviations in important processes associated with the main turbine and turbine generator. PA0 (c) Power distribution operational and alarm status to the plant busses and site grid. PA0 (a) Circulation water system status. PA0 (b) Alarm information for critical deviations in condenser pressure conditions. PA0 Containment Isolation Actuation PA0 Safety Injection Actuation PA0 Main Steam Isolation PA0 Containment Purge Isolation PA0 High Containment Airborne Radiation PA0 High Activity Associated, with Any Release Path PA0 High Coolant Activity PA0 (a) Diesel Generator Status PA0 (b) Status of Power Distribution within the Power Plant PA0 (c) Instrument Air System Status PA0 (d) Service Water System Status PA0 (e) Component Cooling Water System Status PA0 CCW--Component Cooling Water PA0 CD--Condensate PA0 CI--Containment Isolation PA0 CS--Containment Spray PA0 CW--Circulating Water PA0 EF--Emergency Feedwater PA0 FW--Feedwater PA0 IA--Instrument Air PA0 SDC--Shutdown Cooling PA0 RCS--Reactor Coolant PA0 SI--Safety Injection PA0 SW--Service Water PA0 TB--Turbine Bypass Information coding on valves is provided by these additional characteristics/representations: F. Display Integration Information associated with safety related concerns is integrated as a part of the control room information to allow the operator to use safety related information, where possible, during normal operation. This is a better design from a human factors view than that of previous control rooms because in stressful situations, people tend to use information that they are most familiar with. In many situations, safety related parameters are only a subset of the parameters that monitor a particular process variable. Operators of present control room designs typically use control or narrow range indications during process control and should use separate safety related indications when monitoring plant safety concerns. In this invention, the parameters typically used for monitoring and control are validated for accuracy against the safety related parameter(s), where available. If a parameter deviates beyond expected values from the associated safety related information, a validation alarm is presented to the operator. In response to an alarm condition, the operator can review the individual channels associated with the parameter on either a diagnostic CRT page or the discrete indicator displaying that parameter. At this time, he can select the most appropriate sensor for display. The operator is informed when the validation algorithm is able to validate the data. The resultant output of the validation algorithms are used on IPSO, the normally displayed format of a discrete indicator, and the higher level display pages on the CRT display system that contain the parameter. The Regulatory Guide 1.97 category 1 information is also displayed, by discrete indication display, at a single location on the safety monitoring panel. Critical Function and Success Path (availability and performance) information is accesible throughout the information hierarchy (see FIGS. 10, 24, 25, 26, 27, 32-35). Alarms provide guidance to unexpected deviation in critical functions as well as success path unavailability or performance problems. Priority 1 alarms alert the operator to the inability to maintain a critical function as well as the inability of a success path to meet minimum functional requirements. Lower priority alarms provide subsystem/train and component unavailability or poor performance. IPSO provides overview information that is most useful for operator assessment of the Critical Functions Priority 1 alarms associated with the Critical Functions, or Success Paths supporting the critical function are presented on IPSO critical function matrix. Supporting information relating to these alarm conditions is available by using the alarm tiles or the critical function section of the CRT display page hiearchy. The critical function section of the display page hierarchy contains the following information: Level 1 Display Page--"Critical Functions: this page provides more detail on the critical function matrix presented on IPSO. Specifically, more detail on alarm conditions (descriptor, priority). This will help guide the operator to the appropriate level two critical function display page. A 2nd level page exists for each of the 12 critical functions. Each page contains: The 3rd level display pages in the critical function hierarchy are a duplicate of display page existing elsewhere in the hierarchy. For example, a safety injection display page display page under Inventory Control also exists within the primary section of the display page hierarchy. III. DISCRETE INDICATOR AND ALARM SYSTEM A. Discrete Indicators The discrete indicators 82 provide an improved method of presenting safety related parameters. Major process parameters such as Regulatory Guide 1.97 Category 1, require continuous validated display and trending on the master control console. The discrete indicators also provide indication and alarms on parameters needed for operation when the Data Processing System (DPS) is unavailable. These include Regulatory Guide 1.97 Category 1, 2 and 3 parameters, parameters associated with priority 1 or priority 2 alarms, and other surveillance related parameters. Though the DPS is a highly reliable and redundant computer system, its unavailability is considered for a period of up to twenty-four hours. The less frequently viewed parameters are available on discrete indicators, with a menu available by operator selection. Each discrete indicator has the capability to present a number of parameters associated with a component, system, or process. The discrete indicators present various display formats that are based on fulfilling certain operator information requirements. When monitoring or controlling a process such as pressurizer pressure, it is desirable that the operator use a "process representation" value in the most accurate range. For this type of information, the discrete indicator 82 such as shown in FIG. 7 and 8 presents a bold digital value 90 in field 92 and an analog bar graph 94 of the validated average of the sensors in the most accurate range. The preferred validation technique is described in the Appendix, and validated status is indicated in field 96. This validated data is checked against post-accident monitoring indication (PAMI) sensors when applicable. When in agreement with the PAMI, as shown at field 98, the indicator may be used for post-accident monitoring. This has the advantage of continuing to allow the operator to utilize the indicator he is most familiar with and uses on a day-to-day basis. The operator, upon demand, can display any individual channel on the discrete indicator digital display, by touching a sensor identification such as 102. The use of validated parameters is a benefit to operators by reducing their stimulus overload and task loading resulting from presentation of multiple sensor channels representing a single parameter. When the parameter cannot be validated, the discrete indicator displays the sensor reading that is closest to the last validated value. A validation alarm is generated for this condition. The discrete indicator continues to display this sensor's value until the operator selects another value for indication. The field 96 on the discrete indicator that usually read "VALID" displays "FAULT SEL" in reverse image. This indicates that the value is not validated and has been selected by the computer. In this circumstance the operator should review the available sensors that can be used for the "process representation". If the operator makes a sensor selection (which is enabled by a validation fault or failure of the "VALID" signal to agree with PAMI), the field 96 with "FAULT SEL" will be replaced by the message "OPERATOR SELECT", which is displayed in reverse image. When the validation algorithm can validate the data and all faults have cleared, the validation fault alarm will clear and the algorithm will replace the "FAULT SELECT" or "OPERATOR SELECT" "process representation" in field 92 with the "VALID" "calculated signal". Parameters that are required for monitoring the overall performance of plant processes or responding to priority 1 or 2 alarms are provided on discrete indicators. The most representative process parameter is the normally displayed value. Through menu options, the operator can view the other process related parameters. There are ten discrete indicators provided for the RCS panel. The indicators are: FIG. 7 illustrates that two related discrete indicators can be shown on a single display 82. On the left side of the display 82 validated pressurizer pressure is shown whereas at the right, pressurizer level is shown. The pressure display includes the following: digital "process representation" value 90 with units of measurement (2254 psig), quality 96 of the display (VALID), indication 98 that the display is acceptable for post accident monitoring (PAMI), bar chart 94 with the process value, a 30 minute trend 104, normal operating range (NORMAL) 106, instrument range (1500-2500) and units of measurement for the bar chart (psig). In the upper right hand corner of the PRESS display, there are two buttons, "CRT" and "MENU". When touched, the selected button backlights, indicating selection. When the operator removes his hand, the actual selection is processed. The "CRT" button changes the CRT 84 menu options on the CRT located at the same panel as the discrete indicator where the button is pushed e.g., RCS panel 14 as shown in FIG. 3. This "CRT" option identifies the CRT pages most closely associated the parameters on the discrete indicator. The "MENU" button selects the discrete indicator menu (FIG. 8). The upper section of the menu page is nearly identical to the normal display. It contains the digital "process representation" value 90 with units of measurement (2254 psig), quality of display (valid), indication that the display is acceptable for post accident monitoring (PAMI), CRT and MENU buttons. The lower section of the menu page contains selector buttons, such as 102, for all sensor inputs and "calculated signals" of this discrete indicator. The selector buttons 102 backlight when touched, indicating selection. When the operator removes his finger, the actual processing of the selection takes place. There are 13 buttons for pressure: four for 0-1600 psig pressurizer pressure: P-103, P-104, P-105 and P-106; six for 1500-2500 psig pressure: P-101A, P-101B, P-101C, P-101D, P-100X and P-100Y; two for O-4000 psig RCS pressure; P-190A and P-190B; and one four the "calculated signal" pressure: CALC PRESS. When selected, the CALC PRESS button displays the "calculated signal" (i.e., the output of the algorithm). The "calculated signal" of the algorithm can be a "valid" signal. If the algorithm were to fail and select an individual sensor for the "calculated signal", the "valid" message would be replaced by the message "fault select". This message "fault select" would be displayed in reverse image on the discrete indicator. This message would be displayed on the discrete indicator any time "CALC PRESS" is selected until the algorithm outputs a "VALID" signal to replace the "FAULT SELECT" sensor. To change the display, the operator would touch the button containing the sensor he wished to view. For example: by touching the button marked "P-103", the digital display would display the output from the 0-1600 psig range sensor P-103. The message "VALID" below the digital value would be replaced by the message "P-103". Additionally, the "PAMI" message would be removed because P-103 is not a PAMI sensor. The button "ANAL/ALARM OPER SEL" selects the signal used for the "process representation" in DIAS. It selects whatever sensor is displayed on the digital display. The signal select button gives the operator the option to "operator select" any of the sensors for analog display and alarm processing when a fault exists, such as: If a fault were present and the operator elected to select P-103 for the "process representation", he would select the menu, select P-103 for display and then touch the "ANAL/ALARM OPER SEL" button. The message in field 96 below the digital display would read "P-103 OP SEL" in reverse image. Any time P-103 was selected for display, it would have the message "OP SEL" displayed in reverse image, indicating that the output from P-103 is being used for the "process representation". After selecting an "operator select" sensor for the "process representation", it is expected that the operator will depress the button marked "ANALOG DISPLAY". This would return to the analog 94 and trend display 104 (FIG. 7) for the operator selected sensor with the message "OP SEL" in reverse image. The "ANAL/ALARM OPER SEL" button is not normally displayed on the discrete indicator menu page; it automatically displays when the "operator select permissive" is enabled after a fault. The "ANAL/ALARM OPER SEL" button is removed from the menu page when the "operator select permissive" is disabled after all faults are corrected. The button "ANALOG DISPLAY" removes the menu page and replaces it with the bar graph (analog) and trend display for whatever sensor or "calculated signal" is currently selected as the "process representation" (normally the "valid" "calculated signal" output). Other validated process parameter discrete indicators operate in an identical manner. Menu driven discrete indicators contain all level 1 and 2 displays for a functional group of indication. B. Validation Algorithm Summary To reduce an operator's task loading and to reduce his stimulus overload, a generic validation algorithm is used. This algorithm takes the outputs of all sensors measuring the same parameter and generates a single output representative of that parameter, called the "Process Representation". A generic validation approach is used to ensure that it is well understood by operators. This avoids an operator questioning the origin of each valid parameter. This generic algorithm averages all sensors [(A,B,C and D) (sensor quantity may be parameter specific)] and deviation checks all sensors against the average. If the deviation checks are satisfactory, the average is used as the "Process Representation" and is output as a "valid" signal. If any sensors do not successfully pass the deviation check against the average, the sensor with the greatest deviation from the average is taken out and the average is recalculated with the remaining sensors. When all sensors used to generate the average deviation check satisfactorily against the average, this average is used as the "valid process representation". This "valid process representation" is then deviation checked against the post-accident monitoring system sensors (if present). If this second deviation check is satisfactory, the "process representation" is displayed with the message "Valid PAMI" (Post-Accident Monitoring Indication), indicating that this signal is suitable for monitoring during emergency conditions, since it is in agreement with the value as determined by the PAMI sensors. As long as agreement exists, this indicator may then be utilized for post-accident monitoring rather than utilizing the dedicated PAMI indicator. This provides a Human Factors Engineering advantage of allowing the operator to use the indicator he normally uses for any day-to-day work and which he is most familiar with. The validation process, as described, reduces the time an operator takes to perform the tasks related to key process related parameters. To insure timely information, all validated outputs are recalculated at least once every two seconds. Additionally, redundancy and hardware diversity are provided in the calculating devices insuring reliability. The following section describes the algorithm and display processing on the DIAS and CRT displays. a. "Fault Select" value or PA1 b. "Operator Select" value. PA1 Both of these are explained below. PA1 a. "Validation Fault" or PA1 b. "PAMI Fault". PA1 "Alarm Tiles 150" PA1 "Primary 152" It should be appreciated that the discrete validation is accomplished using a generic algorithm that is applicable to different parameters. In this manner, the operators understand how the validated reading has been determined for every parameter and, again, this reinforces their confidence. This algorithm always has an output and allows the operator selection for display when validation is not possible. The discrete indicators continuously display all vital information yet allow easy access via a function or organized menu system to enable the operator to access less frequently needed information. There is no need for separate backup displays, since the backups are integrated in the subsidiary levels of retrieval. Such displays vastly reduce the amount of indicator locations required on the panel and yet provide all vital indication in a easy to use format, thereby reducing stimulus overload. The Appendix in conjunction with FIGS. 37 and 38 provide additional details on the preferred implementation of the algorithm. C. Alarm Processing and Display: Another feature of the monitoring associated with each panel, is the reduction of the number of alarms that are generated, in order to minimize the operator information overload. Cross channel signal validation is accomplished prior to alarm generation, and the alarm logic and set points are contingent on the applicable plant mode. The alarms are displayed with distinct visual cueing in accordance with the priority of the required operator response. For example, priority 1 dictates immediate action, priority 2 dictates prompt action, priority 3 is cautionary, and priority 4, or operator aid, is merely status information. The types of alarm conditions that exist within each category are described below: Priority 1 Priority 2 Priority 3 The alarms are displayed using techniques that help the operator quickly correlate the impact of the alarm on plant safety or performance. These techniques include grouping of displays which highlight the nature of the problem rather than the symptom denoted by the specific alarm condition. Another is the fixed spatial dedication of alarm displays allowing pattern recognition. Another is the plant level pictorial overview display on the IPSO board which shows success paths and critical functions impacted by the priority 1 alarms. To insure that all alarms are recognized by the operator without task overload, all alarms can be either individually acknowledged, or acknowledged in small functionally related groups. All alarms can be acknowledged at any control panel. Momentary audible alerts for alarm state changes require no operator action to silence. Periodic momentary audible reminders are provided for unacknowledged conditions. The operator can affectuate a global alarm stop flash which will automatically resume in time, to allow for deferred acknowledgement. In addition to alarms, an information notification category "Operator Aids" has been established for information that may be helpful for operations but is not, representative of deviations from abnormal conditions. Conditions classified as "Operator Aids" include: channel bypass conditions, approach to interlocks and equipment status change permissive. Some parameters have more than one alarm on the same parameter (i.e., Seal Inlet Temperature Hi Hi and Hi). To limit the operator's required response, the lower priority is automatically cleared without a reset tone or slow flash rate when the higher priority alarm actuates after actuation of the lower priority alarm. The Hi Hi alarm will be acknoweldged by the operator; therefore, the operator acknowledgement of the cleared lower priority alarm is unnecessary. When the condition improves to the point where the higher priority alarm clears, the condition will sound a reset tone and the alarm window will flash slowly. The operator will acknowledge that the higher priority alarm has cleared. If the lower priority alarm condition still exists, its alarm tile or indicator will turn on in the acknowledged state after the operator acknowledges that the higher priority alarm has cleared. If the condition improves such that it clears both the high and low priority alarms before operator acknowledgement, then operator acknowledgement of the cleared high priority alarm will also clear the lower priority condition. 1. Mode and Equipment Dependency A key feature of the alarm system is its mode dependent and equipment status dependent logic. These features combine to greatly reduce the number of alarms received during significant events and limit those alarms to conditions that actually represent process or conditions that actually represent process or component deviations pertinent to the current plant state. Mode and equipment dependency is implemented both through alarm logic changes and setpoint changes. An alarm of mode dependency is the reduction in the low pressurizer alarm setpoint to avoid a nuisance alarm on a normal reactor ring. Equipment dependent logic is used to actuate a low flow alarm only when an upstream pump is supposed to be operating. Four modes have been selected which correspond to significant changes in the alarm logic based on the plant state. These modes are: 2. Subfunction Grouping The RCS panel has over 200 conditions that can cause an alarm. To reduce the operator's stimulus overload due to the quantity of alarms and improve his alarm comprehension, many alarms are grouped into subfunctional groups 108, 110, 112 (FIG. 15). The subfunctional group alarm tiles have a variety of related subfunctional group alarm messages that are read on the panel alarm message window 114 (adjacent to the alarm tile) or CRT. In cases where key process related parameters are alarmed, there is a single alarm message for each alarm tile (i.e., RCS Pressure Low). This single alarm message allows the operator to quickly identify the specific process related problem. As shown in FIG. 16, some alarms are grouped by similar component rather than process function, and are augmented by a message such as 116. As shown in FIG. 9, each alarm tile can be in one of the following states: 3. Shape and Color Coding Alarm information is identified by a unique tile color, preferably yellow 118. The parameter/component descriptor or concise message 120 within the tube is shown in blue. Grey color coding is used for the tile color 122 for Return to Normal conditions. Shape coding is used to identify alarm priority, i.e., 1, 2 or 3. A single bright color is used for alarm information to maximize the attention-getting quality of this information. The shape coding used for identifying alarm priorities uses representational features of decreasing levels of salience. Shape coding of alarm priorities also allows retention of priority information for Return to Normal conditions. For priority 1 alarms, the alarm tiles, mimic diagram components, symbols, process parameters, and menu option fields have their descriptor presented in reverse image (i.e., blue letters 12 on a yellow 178 solid rectangular background 124) using the alarm color coding. The descriptor is presented in blue to provide good contrast for readability. In addition, the alarm tiles and menu option fields on the CRT use the same representation. For priority 2 alarms, the alarm tiles, mimic diagram parameters, components, menu options, and symbols have a thin (1 line) box 12b using the yellow alarm color code around their descriptor, which is blue. For priority 3 alarms, the alarm tiles, mimic diagram parameters, components, menu options, and symbols have brackets 128 around their descriptors 120. For all alarms, English Descriptors on the CRT's message line are also represented with the alarm representation formats when they are in alarm. 4. Alarms on CRT Each CRT page in the data processing system provides the operator with an overview of the existence of any unacknowledged alarm conditions and a general overview of where they exist within the plant. The standard menu provided with each display page contains the IPSO and all first level display: pages as menu options (see FIG. 10 menu region 130). These menu option fields provide the existence of unacknowledged alarms in their sector of the display page hierarchy and their alarm status/priority by using the alarm highlighting feature as described above. If an alarm tile (i.e., in the dias) is in alarm, a first level display page menu option field 2 such as 132, in the menu options 130 shows that an alarm condition exists in an associated area of the display page hierarchy. The alarm tiles in menu 130 are categorized into the first level display page set corresponding to the console groupings or by critical function as shown in FIG. 11. In addition to alarm information represented on the first level display page menu options, the following display page features are also used to represent the existence of alarms. Display page menu option 134 that provide access to levels 2 and 3 display pages are lit with the above described alarm representation if information on the corresponding page is in alarm (e.g., if an unacknowledged alarm exists, the display page menu option is highlighted to show the highest priority unacknowledged condition). The operator can by selecting option 136, call up a level 2 display page, directory containing a pictorial diagram of the level 3 display pages in a hierarchical format associated with a first level display page (see FIG. 12). Each of the level 2 and 3 display pages represented on this diagram provide alarm notification if information on that display page is in an unacknowledged alarm state. This alarm information is most useful for determining where alarms exist within an area of the display page hierarchy. For example, the operator would be notified by the display page menu 130 (FIG. 10) that an unacknowledged alarm(s) exists in the auxiliary systems by grey alarm shape coding (return to normal) and slow flashing of alarm coding on the "PRI" menu option field. He can then access that directory/hierarchy to see what page(s) contains alarm information by touching the menu option "DIRECTORY" 136 followed by "PRI". When the Primary display directory comes up (FIG. 12), the field(s) representing the display page(s) that contains the alarm condition(s) (such as P2R level 138) will be highlighted. The desired page that contains the alarm information (similar to FIG. 15) is accessed by touching the flashing field. The descriptors of components and plant data on the process display pages of the CRT (FIG. 13) are alarm coded and flashed to provide indication of alarms and their acknowledgement status. A component's descriptor can provide this alarm information if a parameter associated with the component is in alarm. This is true even if the parameter in alarm is not represented on the display pages, e.g., low pump lube oil pressure is represented by alarm coding of the associated component's symbol. To view the exact information that is in alarm, the operator can access a lower level display page, or use the alarm system features that are described later. 5. Determining Alarm Conditions and Acknowledging Alarms With reference again to FIG. 16, each category 1 and 2 alarm annunciator tile in the DIAS may notify the operator of more than one possible alarm condition. To quickly determine the actual alarm condition, a message window 114 is provided in the display. area 78 on the panel. By depressing an unacknowledged alarming annunciator tile such as 134, an English description 116 of the specific alarm condition is provided on the message window 114. The alarm tile 134 remains flashing until all alarm conditions associated with the alarm tile have been acknowledged. The English descriptors of additional alarms can be accessed by redepressing the alarm tile 134. At the same time that a message appears on the message window of a DIAS alarm display 78, an alarm message is presented on another field 132 at the bottom of the display page 84 on the panels CRT (see FIG. 13). The CRT alarm message contains the following information: Time, Priority, Severity (e.g., Hi, Hi-Hi), Descriptor, Setpoint, and real time process value (coded as described to show the alarm priority and alarm condition). If additional unacknowledged alarms exist that are associated with the tile, the number of additional unacknowledged alarms is specified within a circle 136 the right hand side of the message area (see FIG. 13). In addition to this alarm message, menu options/fields appear on the display page menu (Region 4) and provide direct access to the display pages that can be used to obtain supporting or diagnostic information of the alarm condition. The display regions are shown in FIG. 22. The alarm tiles that are in alarm on the DIAS display 78 of given panel can be accessed and acknowledged on any CRT panel by procedure, similar to accessing and acknowledging the alarms via the alarm tiles. By selecting the "Alarm Tiles" menu option followed by an alarming display page menu option, i.e., first level display page set (region 3), the alarm tiles that are in alarm, that are associated with the display page, are provided in region 4 of the display page menu. One tile is depicted and is a touch target that provides access to other tiles. The operator acknowledges and reviews these CRT alarm tiles by touch and obtains alarm messages and supporting display page touch targets in the same format as described above. This means of responding to alarming alarm tiles is most useful for responding to alarms at workstations that are remote to the operator's location. All alarm conditions associated with an annunciator tile in the DIAS display are held in a buffer. The buffer containing alarm conditions is arranged in the following format: ______________________________________ 1. First-In Unacknowledged 2. . . . . . N Last-In Unacknowledged N + 1 First-In Cleared/Return to Normal N + 2 . . . . . . . n Last-In Cleared/Return to Normal n + 1 Acknowledged Alarms n + 2 . . . . . ______________________________________ Depressing an alarm tile provides access to the alarm condition that is at the top of the buffer. Acknowledging unacknowledged alarms moves these alarm conditions to the bottom of the buffer. Acknowledging cleared alarms drops them from the buffer. Previously acknowledged alarm(s) (n+1,n+2,..) can be reviewed when there are no unacknowledged or cleared unacknowledged alarm conditions present. Upon reviewing these alarms, they move to the bottom of the buffer. Alarm messages for priority 3 alarms and operator aids are only generated by the computer and only appear on the message line 132 of the CRT page (FIG. 3); there will be no English descriptor provided on the message window of the DIAS display 78. One annunciator tile is provided at each annunciator workstation for all priority 3 alarms and 1 alarm tile is provided on the workstation for operator aids that are associated with these workstation. When an alarm condition changes priority, the following changes occur in the alarm handling system. When a higher priority alarm comes in on the same parameter, the previous alarm is automatically cleared (i.e., no operator acknowledgement necessary since he will need to acknowledge the higher priority condition) without a reset tone or slow flash rate. When an alarm condition improves to the point where the high priority alarm clears, the operator will need to acknowledge that the higher priority alarm has cleared; however, if the lower priority alarm still exists, it will turn on (upon operator acknowledgement of the higher priority cleared condition) and automatically go to the acknowledged state (i.e., no operator action required). The new lower priority alarm condition will be observed by the operator when reading the alarm message in response to clearing the highest priority alarm. The invention provides a means of listing and categorizing alarms, and accessing supporting display pages. In this system, alarms are provided on alarm listing display pages accessible from the fields 138 of the DIAS display 78 and 140 of the CRT display 84 shown in FIGS. 15 and 13, respectively. The categories of alarms in this listing are as follows (see FIG. 14): A workstation's alarm tiles in alarm are listed by priority. Alarms associated with the alarm tiles are listed as they are contained in the alarm tile's alarm buffer. These alarm categories provide alarm data consistent with operator's information needs in response to alarm conditions. When accessing the Categorized Alarm Listing 78 via page 84 (FIGS. 4 and 12), the operator can easily select the data in the category he wishes to see. Using the "Alarm List" menu option by 14 (FIG. 4) followed by a display page feature that represents alarm condition (s), (FIG. 12) the operator can view the specific alarm conditions that he is interested in (FIG. 14). Three examples of accessing alarm data in the categorized list from page 84 (FIG. 4) follow. The display page's menu changes to a representation of the alarm tiles that are in alarm and are associated with the Primary Systems (see FIG. 14). At this time, the operator can request one of two different types of information formats associated with the displayed alarm tiles: Alarm information is also provided on all process display mimic diagrams which contain a component or parameter which is in an alarm condition. Color, and shape coding is used to indicate alarm conditions, as described earlier. Parameters in alarms that are associated with a component can cause the represented component's descriptor to be highlighted to indicate an alarm condition if the parameter is not visible on the display page, e.g., pump lube oil pressure may not be listed on a level two display page, so the pump's descriptor may be alarm coded. If the operator desires to see the exact alarm condition associated with a component, he would access the appropriate lower level display page. Alternatively, he could touch the "Alarm Tiles" menu option followed by touching the component's descriptor and respond to the alarm using alarm tile representations. This action also accesses menu options associated with display pages that provide more detail about the component. The following means of alarm acknowledgement is provided with the invention. Each of these methods of alarm acknowledgement clears unacknowledged alarm indicators in the other alarm formats. When an alarm condition clears, the operator needs to be notified. Notification is accomplished by flashing the annunciator tiles and associated process display page information at a slow rate. Acknowledging or resetting the cleared alarm indications takes place in a mechanism similar to acknowledgement of new alarms, i.e., touching an alarm tile or CRT alarm representation/feature. Distinct sounds/tones are provided in the control room to indicate the following alarm information: An audible alarm, tone 1 or 3, is only present for 1 second and tone 2 will repeat periodically, once every minute, until all new or cleared alarms are acknowledged. In situations where multiple unacknowledged alarms exist, the operator needs to direct his attention at the highest priority new alarm conditions. In this situation, all other unacknowledged alarms, i.e., new priority 2, 3 and all cleared alarm conditions, are added noise that distracts the operator from most important alarm conditions. In the control room, a "STOP FLASH" and "RESUME" button exists at the MCC, ACC and ASC. When the "STOP FLASH" button is depressed, the alarm system's behavior exhibits the following characteristics: The alarm reminder tone informs the operator about any unacknowledged new or cleared alarm conditions that exist. To identify these conditions for acknowledgment, the operator selects a "resume" button which returns all unacknowledged and cleared conditions to their normal representational alarm status. The alarm suppression button is backlit after selection to show that the alarm suppression feature is active. So that the operator can provide quick, direct access to supporting information thereby enhancing the operator response to alarm conditions, a single operator action provides alarm acknowledgement, display of alarm parameters, and selection options for CRT display pages appropriate for the alarm condition. The invention provides redundancy and diversity in alarm processing and display such that the operators have confidence in intelligent alarm processing techniques and such that plant safety and availability are not impacted by equipment failures. Priority 1 and 2 alarms are processed and displayed by two independent systems. Two-system redundancy is invisible to the operators through continuous cross-checking and integrated operator interfaces. FIGS. 16-18 show a schematic alarm response using the tiles in accordance with the invention. The illustrated group of tiles is associated with the reactor coolant pump seal monitoring in the reactor cooling system panel shown in FIG. 3. The priority 2 seal/bleed system trouble alarm is illuminated to alert the operator, who then can read a more complete message in the message window, which indicates a high control bleed-off pressure. Such a message is provided for priority 1 and 2 alarms. The same message in more complete form is displayed on the panel CRT. The CRT also identifies menu options that indicate useful supporting display pages. Alternatively, the operator may directly access a listing of all the alarms in a particular group. Thus, overview of the alarm conditions is provided with the tiles, and the detail is provided with the associated messages. A given alarm is rendered more or less important at a particular point in time, depending on the equipment status and the mode of operation of the NSSS. Alarm handling is reduced by validation of the parameter signals, and clearing automatically lower priority alarms when one of the higher priority alarms is actuated on the same condition. IV. DATA PROCESSING SYSTEM A. The CRT Display The CRT shown 84 in the center of the panel in FIG. 3 is part of the data processing system which processes and displays all plant operational data. Thus, it is linked to all other instrumentation and control systems in the control room. FIGS. 2, 28 and 30 schematically show the relationship of the data processing system with the control system, plant protection system, and discrete indication and alarm, system. The data processing system 70 receives from the control system 64, the same sensor data that is used by the control system for executing the control logic. Likewise, it receives from the discrete indication and alarm system 72 the validated sensor data that is used by the discrete indication and alarm system for generating the discrete alarms and displays. The plant protection system 50 does not use internally validated data for its trip logic, and this "raw" signal is for each channel passed along to the data processing system 70 which performs its own signal validation logic 154 on the plant protection system signals, and passes on the internally validated signal to the validated signal comparison logic 156. In that functional area, the validated signals from the control system 64, the plant protection system 50 and the discrete indication and alarm system 72 are compared and displayed on the CRT 84. It should be appreciated that both the validated signal from the comparison logic 156 and the validated signal from the plant protection system are available for display on the CRT 84. Thus, the CRT display within each panel includes signal validation and all CRTs in the plant are capable of accessing any information available to the other CRTs in the plant. Moreover, on any given CRT, the alarm tile images from any other panel may be generated and the alarms acknowledged. Detailed display indicator windows may be accessed as well. The CRTs have a substantially real time response, with at most a two-second delay. The CRT display pages contain all the power plant information that is available to the operator, in a structured, hierarchic format. The CRT pages are very useful for information presentation because they allow graphical layouts of power plant processes in formats that are consistent with operator visualization. In addition, CRT formats can aid operational activities, where appropriate, by providing trends, categorized listing, messages, operational prompts, as well as alert the operator to abnormal processes. The primary method the operator obtains information formats on the CRTs is through a touch screen interface which operates in a known manner. The touch screens are based on infrared beam technology. Horizontal and vertical beams exist in a bezel mounted around the face of each color monitor. When the beams are obstructed by the user, the coordinates are cross-referenced with the display page data base to determine the selected information. Messages and Supporting Display page option touch targets can be accessed onto panel CRTs by touching other panel features, e.g., discrete indicators and alarm tiles. IPSO is available as a display page and forms the apex of the display page hierarchy (See FIGS. 10, 22 and 24). Three levels exist below IPSO, where each level of the hierarchy provides consistent information content to satisfy particular operational needs. The structure of the hierarchical format is based on assisting the operator in the performance of his tasks as well as providing quick and easy access to all information displayed via the CRTs. The display formats on the top level provide information for general monitoring activities, while the lowest level formats contain information that is most useful for supporting diagnostic activities. Level 1 display pages provide information that is most useful for general monitoring activities associated with a major plant process. These display pages inform the operator of major system performance and major equipment status and provide direction to lower level display pages for supportive or diagnositc information. The level 1 display pages are as follows: Level 2 display pages provide information that is most useful for controlling plant components and systems. These pages contain all information necessary to control the system's processes and functions. Parameters which must be observed during controlling tasks appear on the same display, even though they may be parts of other systems. Proposed operating procedures or guides for controlling components are utilized for determining which parameters to display. FIG. 20 is a sample display for Reactor Coolant Pump 1A and 1B Control. The operator would normally monitor the "Primary System" display page to assess RCS performance. If the operator wishes to operate or adjust RCP 1A or 1B, the operator would access the control display page. All information for Reactor Coolant Pump Control is on the control display to preclude unnecessary jumping between display pages. Level 3 display pages provide information that is most useful for diagnostic activities of the component and processes represented in level 2 display pages. Level 3 display pages provide data useful for instrument cross-channel comparisons, detailed information for diagnosing equipment or system malfunctions, and trending information useful for determining direction of system performance changes, degradation or improvement. FIG. 21 shows a diagnostic display of the Seal and Cooling section of RCP1A; the pump portion, the supporting oil system, and the motor section are presented on a separate display page due to display page information density limits. Display page access is accomplished through the use of menus placed on the bottom of the display pages. Each display page contains one standard menu format that provides direct, i.e., single touch, access to all related display pages in the information hierarchy. The menu has fields (see FIG. 10) where display page title are listed. By selecting a field (a thru j), the specified display page is accessed. The menu option fields associated with a display page includes the following (see FIG. 22). To access a display page described by a menu option, the operator would select the menu option (a-k) by touching the desired menu option field on the monitor. The menu option is highlighted (using black letters on a white background) until the display page appears. Since the menu options provide direct access to a minimum set of display pages in the display page hierarchy, alternate means are available for quickly accessing other display pages. Three options are available to the operator: In addition to the menu options described above, menu options exist for "LAST PAGE", "ALARM LIST", "ALARM TILES", "OTHER", and horizontal paging options ("Keys"). The "LAST PAGE" (option j on FIG. 22) provides direct access to the last page that was on the monitor. This is very useful to operators for comparison of information between two display pages, or retrieval of information that the operator was previously involved with. The "ALARM LIST" (option n on FIG. 22) provides for quick access to the alarm listing display pages. The "ALARM TILES" (option m on FIG. 22) provides for quick access to the alarm tile representations of active alarm tiles in the area above Region 4(see FIG. 23) of the workstation's CRT menu. This allows an operator to access alarm information associated with specific tiles on any workstation's CRT. This method of alarm access is further described in Section 5 of this document. The "OTHER" (option k on FIG. 22) provides access to display pages or information that does not fall into the categories of information described by the presently displayed menu options. B. IPSO Another part of the data processing system is the integrated process status overview (IPSO board). Although the number of displays and alarms stimulating the operator at any one time can be considerably reduced using the panels having the discrete alarm, discrete display, and CRT displays described above, the number of stimuli is still relatively high and, particularly during emergency operations, may cause delay in the operator's understanding of the status and trends of the critical systems of the NSSS. A single display is needed that presents only the highest level concerns to the operator and helps guide the operator to the more detailed information as it is needed. Although some attempts have been made in the past to present a large board or display to the operator, such displays to date have not included a significant consolidation of information in the nature to be described below. The IPSO board presents a high level overview of all high level concerns including overview of the plant state, critical safety and power functions, symbols representing key systems and processes, key plant data, and key alarms. IPSO information includes trends, deviations, numeric values of most representative critical function parameters, and the existence and system location of priority 1 alarms including availability and performance status for systems supporting the critical functions. This is otherwise known as success path monitoring. The IPSO board also can identify the existence and plant area location of other unacknowledged alarms. Thus, IPSO bridges the gap between an operator's tendency toward system thinking and a more desirable assessment of critical functions. This compensates for reduction in the dedicated displays to help operators maintain a field plant conditions. It also helps operators maintain an overview of plant performance while being involved in detailed diagnostic tasks. IPSO provides a common mental visualization of the plant process to facilitate better communication among all plant personnel. In FIG. 25, the condition illustrated is a reactor trip. At the instance illustrated, the temperature rise in the reactor is 27.degree. and the average temperature rise is higher than desired and rising as indicated by the arrow and "+". The pressurizer pressure is higher than desired, but it is falling. Likewise, the steam generator water level is higher than desired but falling. FIG. 24 shows a CRT display page hierarchy wherein the IPSO is at the apex, the first level display page set contains generic monitoring information for each of the secondary, electrical, primary, auxiliary, power conversion and critical function systems, the second level of display pages relates to system and/or component control, and the third level of display pages provides details and diagnostic information. IPSO is a continuous display visible from any control room workstation, the shift supervisor's office, and Technical Support Center. The IPSO is centrally located relative to the master control console. The IPSO also exists as a display page format that is accessible from any control room workstation CRT as well as remote facilities such as the Emergency Operations Facility. The IPSO large panel format is 4.5 feet high by 6 ffet wide. Its location, above and behind the MCC workstation, is approximately 40 feet from the shift supervisor's office (the furthest viewable point). One of the beneficial aspects of IPSO is the use of IPSO information to support operator response to plant disturbances, particularly when a disturbance effects a number of plant functions. IPSO information supports the operator's abaility to respond to challenges in plant power production as well as safety-related concerns. IPSO supports the operator's ability to quickly assess the overall plant's process performance by providing information to allow a quick assessment of the plant's critical safety functions. The concept of monitoring plant power and safety functions allows a categorization of the power and safety-related plant processes into a manageable set of information that is representative of the various plant processes. The critical functions are: ______________________________________ Critical To: Function Power Safety ______________________________________ 1. Reactivity Control X X 2. Core Heat Removal X X 3. RCS Heat Removal X X 4. RCS Inventory Control X X 5. RCS Pressure Control X X 6. Steam/Feed Conversion X 7. Electric Generation X 8. Heat Rejection X 9. Containment Environment Control X 10. Containment Isolation X 11. Radiological Emissions Control X X 12. Vital Auxiliaries X X ______________________________________ A 3.times.4 alarm matrix block 160 containing a box 162 for each critical function exist and the CRT display of ISPO in FIG. 10 in the upper right hand corner of ISPO (see FIG. 25). The matrix provides a single location for the continuous display of critical function status. If a priority 1 alarm condition exists that relates to a critical function, the corresponding matrix box 164 will be highlighted in the priority 1 alarm presentation technique. Critical Function alarms are representative of one of the following priority 1 conditions: The 3.times.4 matrix representation is an overview summary of the 1st level critical function display page information (FIG. 32). The operator obtains the details associated with critical function and Success Path alarms in the Critical Function section of the display page. Each critical function can be maintained by one or more plant systems. Information on IPSO is most representative of the ability of supporting systems to maintain the critical functions. For some critical functions, the overall status of the critical function can be assessed by a most representative controlled parameter(s). For these critical functions, the process parameter's relationship to the control setpoint(s) and indication of improving or degrading trends is represented on IPSO to the right of the parameter's descriptor. An arrowhead as explained in FIG. 26 is used if the integral of the parameter's value is greater than an acceptable narrow band control value, indicating that the parameter is moving toward or away from the control setpoint. The arrowhead's direction, up or down, indicates the direction of change of the process parameter. If these parameters deviate beyond normal control bounds, a plus or minus sign is placed above or below the control setpoint representation. The following bases were used for the selection of parameters or other indications that are used on IPSO to provide the monitoring of the overall status of the critical functions. 1. Reactivity Control Reactor power is the only parameter displayed on the IPSO as a means of monitoring reactivity. Using Reactor Power, the operator can quickly determine if the rods have inserted. He can also use Reactor Power to determine the general rate and direction of reactivity change after shutdown. Reactor Power, is displayed on IPSO with a digital representation 166 because a discrete value of this parameter is most meaningful to both operators and administrative personnel. The IPSO also provides an alarm representation on the reactor vessel if there is a priority 1 alarm condition associated with the Core Operating Limit Supervisory System. 2. Core Heat Removal A representative Core Exit Temperature 168 and Subcooled Margin 170 are the parameters presented on IPSO for determining if Core Heat removal is adequate. If Core Exit Temperature is within limits, then the operator can be assured of maintaining fuel integrity. The Subcooling Margin is used because it gives the operator the temperature margin to bulk boiling. Core Exit Temperature is represented on IPSO by using a dynamic representation (i.e., trending format), since there is a distinct upper bound that defines a limit to core exit temperature, and setpoints for representational characteristics can be easily defined. Subcooled Margin is also represented on IPSO using a dynamic representation since there is a lower bound which defines an operational limit for maintaing subcooling. 3. RCS Heat Removal T.sub.H, T.sub.C, S/G Level 172 and T.sub.ave 174 are used on IPSO to provide the operator the ability to quickly assess the effectiveness of the RCS Heat Removal Function. In order to remove heat from the Reactor Coolant, S/G Level must be sufficiently maintained so that the necessary heat transfer can take place from the RCS to the steam plant. A dynamic representation is used so the operator can observe degradiations or improvements in deviant condition at a glance. T.sub.H and T.sub.C are used on IPSO because they are needed by the operator to determine how much heat is being transferred from the reactor coolant to the secondary system. A digital value of these parameters is used since a quick comparison of these parameters is desired for observing the delta T. In addition, an indication of their actual values are used often and would be helpful to an operator in locations where the discrete indicator displaying T.sub.h and T.sub.c is not easily visible. T.sub.ave is presented on IPSO using a dynamic representation to allow quick operator assessment of whether this controlled parameter is within acceptable operating bounds. 4. RCS Inventory Control Pressurizer Level 176 is presented on the IPSO using a dynamic representational indication to allow the operator to quickly access if the RCS has the proper quantity of coolant and observe deviations in level indicative of improving or degrading conditions. 5. RCS Pressure Control Pressurizer Pressure 178 and Subcooled Margin is used as the indications on IPSO to determine the RCS Pressure Control. A dynamic representation is used on IPSO to notify the operator of changing pressure conditions that may indicate RCS depressurization or over pressurization. A dynamic representation is used on IPSO for saturation margin. A saturation condition in the RCS can adversely affect the ability to control pressure by the pressurizer. Also, if pressure is dropping, the subcooled margin monitor representation on IPSO depicts a decrease in the margin to saturation. 6. Steam/Feed Conversion The processes associated with Steam/Feed Conversion can be quickly assessed by providing the following information on IPSO: 7. Electric Generation The processes associated with Electric Generation can be quickly assessed by providing the following information on IPSO: 8. Heat Rejection The processes associated with heat rejection can be quickly assessed by providing the following information on IPSO: 9. Containment Environment Control Containment Pressure and Containment Temperature are the parameters which are used on the IPSO to monitor the control of the Containment Environment. These are presented on IPSO using a dynamic representation to allow assessment of trending and relative values. The Containment Pressure variable is used on the IPSO to warn the operator about an adverse overpressure situation which could be the result of a break in the Reactor Coolant System. The Containment Temperature also helps indicate a possible break in the Reactor Coolant System; it also can indicate a combustion in the Containment Building. 10. Containment Isolation The Containment Isolation Safety function is monitored on the IPSO with a Containment Isolation system symbol representation. This symbol will be driven by an algorithm which presents the effectiveness of the following containment isolation situations when the associated conditions warrant containment isolation: 11. Radiological Emissions Control Radiation symbols exist on IPSO which presents notification of high radioactivity levels such as inside containment, and (2) radiation associated with radioactivity release paths to the environment. these symbols will only be presented on IPSO when high radiation levels exist. These indications are presented in the alarm color in a location relative to the sensor in any of the following situations occurs: 12. Vital Auxiliaries Vital Auxiliaries are monitored on IPSO by providing the following information: The systems represented on IPSO are the major heat transport path systems and systems that are required to support the major heat transport process, either power or safety related. These systems include systems that require availability monitoring per Reg. Guide 1.47, and all major success paths that support the plant Critical Functions. The following systems have dynamic representations on IPSO: System Information presented on IPSO includes systems operational status, change in operational status (i.e., active to inactive, or inactive to active) and the existence of a priority one alarm(s) associated with the system. Alarm information on systems can also help inform an operator about success path related Critical Function alarms. Priority 1 alarm information is also presented on IPSO by alarm coding the descriptors of the representative features on IPSO as described above. V. INTEGRATION OF CONTROL ROOM FIG. 27 presents an overview of the integrated information presentation available to the operator in accordance with the invention. From the integrated process status overview or board, the operator may observe the high priority alarms. If the operator is concerned with parameter trends, he may view the discrete indicators. If he is interested in the system and component status, he may view the settings on the system controls. Thus, the IPSO information is displayed either on the board or at the panel CRT, and the other information from the operator's panel or any other panel, is available to the operator on his CRT. From the IPSO overview, the operator may navigate through the CRT or DIAS display pages. Moreover, the operator has direct access to either of these types of information from any of the control panels and when a system control is adjusted or set, the results are incorporated into the other alarm and display generators in the other panels. As shown in FIGS. 2 and 28-31, in general overview, the integration of the system means that each panel including the main console, the safety console, and the auxiliary console, includes a CRT 84 which driven by the data processing system 70 . The data processing system utilizes the plant main computer and, although being, more powerful, it is not as reliable as the DIAS 72 (which may be distributed microprocessors-based or mini-computer based). Also, it is slower because it is menu driven and performs many more computations. It is used primarily for conveying the most important information to the operator and thus important alarm tiles can be viewed on each CRT and acknowledged from any CRT. Any information available on one CRT is available at every other CRT. The indicator and alarm system 72 for a given panel is related to the controls, but the discrete (i.e., quick and accurate) aspects of the alarms and indicator displays 78, 82, and controls of that panel are not available at any other panel. Basically, information is categorized in three ways. Category 1 information must be continuously displayed at all times and this is accomplished in DIAS 72. Category 2 information need not be continuously available, but it must nevertheless be available periodically and this is also the responsibility of DIAS 72. Category 3 information is not needed rapidly and is informational only, and that is provided by the DPS 70. In the event of the failure of DPS, some essential information is provided by DIAS. The DPS and DIAS are connected to the IPSO board by a display generator 180. From the IPSO, the operator can obtain detailed information either by going to the panel of concern, or paging through the CRT displays. It should be appreciated that DIAS and DPS do not necessarily receive inputs for the same parameters, but, to the extent they do receive information from common parameters, the sensors for these parameters are the same. Moreover, the validation algorithms used in DIAS and DPS are the same. Furthermore, the algorithms used for the discrete alarm tiles and the discrete indicators include as part of the computation of the "representative" value, a comparison of the DIAS and DPS validated values. FIG. 29 is a block diagram representing the discrete indicator and alarm system in relation to other parts of the control room signal processing. The DIAS system preferably is segmented so that, for example, all of the required discrete indicator and discrete alarm information for a given panel n is processed in only one segment. Each segment, however, includes a redundant processor. The information and processing in DIAS 1 is for category 1 and 2 information which is not normally displayed directly on IPSO. IPSO normally receives its input from the DPS. However, in the event of a failure of DPS, certain of the DIAS information is then sent to the IPSO display generator for presentation on the IPSO board. It should also be appreciated that both DIAS and the DPS utilize sensor output from all sensors in the plant for measuring a given parameter, but that the number of sensors in the plant for a given parameter may differ from parameter to parameter. For example, the pressurizer pressure is obtained from 12 sensors, whereas another parameter, for example, from the balance of plant, may only be measured by two or three sensors. Some systems, such as the plant protection system, do not employ validation because they must perform their function as quickly as possible and employ, for example, a 2 out of 4 actuation logic from 4 independent channels. In the event the validation for a given parameter differs as determined within two or more systems, an alarm or other cue will be provided to the operator through the CRT. One of the significant advantages of the present invention is that the DPS need not be nuclear qualified, yet it can be confidently used because it obtains parameter values from the same sensors as the nuclear qualified DIAS. These are validated in the same manner and a comparison is made between the validated DPS parameters and the validated DIAS parameters, before the DPS information is displayed on the CRTs or the IPSO. The nuclear qualification of the alarm tiles and windows, and the discrete indicator displays in the DIAS are preferably implemented using a 512.times.256 electroluminescent display panel, power conversion circuitry, and graphics drawing controller with VT text terminal emulation, such as the M3 electroluminescent display module available from the Digital Electronics Corporation, Hayward, California. The control function of each panel is preferably implemented using discrete, distributed programmable controllers of the type available under the trademark "MODICON 984" from the AEG Modicon Corporation, North Andover, Mass., U.S.A. Thus, the computational basis of the DIAS is with either distributed, discrete programmable microprocessors or mini computers, whereas the computational basis of the DPS is a dedicated main frame computer. The ESF control system and the process component control system are shown schematically in FIG. 31, whereas the plant protection system is preferably of the type based on the "Core Protection Calculator" system such as described in U.S. Pat. No. 4,330,367, "System and Process for the Control of a Nuclear Power System", issued on May 18, 1982, to Combustion Engineering, Inc., the disclosure of which is hereby incorporated by reference. Another aspect of integration is the capability to display the critical functions and success path in IPSO as described above. Since the major safety and power generating signal and status generators are connected to both DIAS and DPS, the operator may page through the critical functions in accordance with the display page hierarchy shown in FIGS. 32 through 35. In FIG. 33, the operator is informed that the emergency feed is unavailable in the reactant coolant system. In FIG. 34, the operator is informed that the emergency feed is unavailable and the reactor is in a trip condition. Under these circumstances, the operator must determine an alternative for removing heat from the reactor core and by paging to the second level of the critical function display page which, although shown for inventory control (FIG. 35), would have a comparable level of detail for heat removal. This type of information with this level of detail and integration is available for all critical functions under substantially all operating conditions, not only during accidents. VI. PANEL MODULARITY It should be appreciated that, as mentioned above, the discrete tile and message technique significantly reduces the surface area required on the panel to perform that particular monitoring function. Similarly, the discrete display portion of the monitoring function, including the hierarchical pages, is condensed relative to conventional nuclear control room systems. The control function on a given panel can be consolidated in a similar fashion. Thus, a feature of the present invention is the physical modularity of each panel constituting the master control console, and more generally, of each panel in the main control room. In essence, the space required for effective interface with the operator for a given panel, becomes independent of the number of alarms or displays or controls that are to be accessed by the operator. For example, as shown in FIG. 3, six locations on each side of the CRT may be allocated for alarm and indicator display purposes. Preferably, the top two on each side are dedicated to alarms 78 and the other, four on each side dedicated to the indicator display 82. An identical layout is provided for each panel in the control room. This permits significant flexibility and cost savings during the construction phase of the plant because the hardware can be installed and the terminals connected early in the construction schedule, even before all system functional requirements have been finalized. The software based systems are shipped early with representative software installed to allow preliminary checking of the control room operations. Final software installation and functional testing are conducted at a more convenient point in the construction schedule. This method can accelerate plant construction schedules for the instrumentation and control systems significantly. Since the instrumentation and control requirements for a given plant are often not finalized until late in the plant design schedule, the present invention will in almost every case significantly reduce costly delays during construction. This is in addition to the obvious cost savings in the ability to fabricate uniform panels, both in the engineering phase normally required to select the locations of and lay out the alarms and displays, and in the material savings in fabricating more compact panels. Furthermore, such modularity in the plant facilitates the training of operators and, when operators are under stress during emergencies, should reduce operator error because the functionality of each panel is spatially consistent. Thus, each modular control panel has spatially dedicated discrete indicators and alarms, preferably at least one spatially dedicated discrete controller at 88, a CRT 84, and interconnections with at least one other modular control panel or computer for communication therewith. For example, communication via the DPS includes, among other things, the ability to acknowledge an alarm at one panel while the operator is located at another panel, and the automatic availability at every other panel of information concerning the system controlled at one panel. FIG. 36 (a) illustrates the conventional sequence for furnishing instrumentation and control to a nuclear power plant and 36(b) the sequence in accordance with the invention. Conventionally, the input and outputs are defined, the necessary algorithms are then defined, and these specify the man machine interface. Fabrication of all equipment then begins and all equipment is installed in the plant at substantially the same time before system testing can begin. In contrast, the modularity of the present invention permits fabrication of hardware to begin immediately in parallel with the definition of the input/output. Likewise, the hardware can be installed and generically tested in parallel with the definition of the man machine interface and the definition of the algorithms that are plant specific. The hardware and software are then integrated before final testing. In a conventional nuclear installation, the equipment is installed during the fourth year of the entire instrumentation and control activity, whereas with the present invention, equipment can be installed during the second or third year. With further reference to FIG. 2, the process component control system and the engineered safety features component control system 56 use programmable logic controllers similar to the Modicon equipment mentioned above including input and output multiplexors and associated wires and cabling, all of which can be shipped to the plant before the plant specific logic and algorithms have been developed. This equipment is fault tolerant. The data processing system 70 uses redundant plant main frame computers, along with modular software and hardware and associated data links. Such hardware can be delivered and the modular software that is specific to the plant installed, just prior to integration and system testing. The DIAS 72 also uses input/output multiplexors and a fault tolerant arrangement, with programmable logic processors or mini-computers, with the same advantages as described with respect to the process control and engineered safety features control systems. ##SPC1## |
050323512 | abstract | An improved spacer and method of making a spacer is disclosed for use in a nuclear fuel bundle wherein a plurality of fuel rods enclosed within a channel are maintained in parallel side-by-side relation by a plurality of the spacers. Each spacer is placed within the fuel bundle at selected elevations between upper and lower tie plates. The improved spacer is a member of the class of spacers wherein solid strips of material are welded at interstitially placed tube members between the fuel rods to form the continuous spacer grid. The improvement constitutes forming separate upper and lower reduced section grids from separate, normally aligned, first and second parallel sets of grid members. One grid is formed for the top of the spacer; the remaining grid is formed for the bottom of the spacer. Tube members placed interstitially between the fuel rods are used to interconnect the grids. The tube members themselves are in turn notched; the notches are at the upper portion of the tube members to receive the upper grid and at the lower portion of the tube members to receive the lower grid. Grids are placed within the notched tube members and fastened, typically by welding to the top and bottom of the tubes to form a unitary spacer structure. Thereafter, the excess material of the grid crossing the interior of the tube members is drilled out of the tube members to eliminate excess neutron absorbing material. There results in the disclosed spacer, two interconnected grid members extending at the top and bottom of the spacer having less material than the single and continuous grid of the prior art. At the same time, the assembly of what is otherwise a difficult member to construct is simplified. Provision is made for applying the improved spacer to spacers having differing pitch between the separated fuel rods. Additionally the incorporation of so-called swirl vanes in some grid locations is disclosed in substitution for the tube members. |
047284803 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also, in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. IN GENERAL Referring now to the drawings, and particularly to FIG. 1, there is shown an elevational view of a fuel assembly, represented in vertically foreshortened form and being generally designated by the numeral 10. The fuel assembly 10 is the type used in a pressurized water reactor (PWR), being illustrated diagrammatically in a simplified form in FIG. 10 and designated by the numeral 12. The fuel assembly 10 basically includes a lower end structure or bottom nozzle 14 for supporting the assembly 10 on the lower core plate (not shown) in the core 16 of the PWR 12, and a number of longitudinally extending guide tubes or thimbles 18 which project upwardly from the bottom nozzle 14. The assembly 10 further includes a plurality of transverse grids 20 axially spaced along the guide thimbles 18 and an organized array of elongated fuel rods 22 transversely spaced and supported by the grids 20. Also, the assembly 10 has an instrumentation tube 24 located in the center thereof and an upper end structure or top nozzle 26 attached to the upper ends of the guide thimbles 18. With such an arrangement of parts, the fuel assembly 10 forms an integral unit capable of being conventionally handled without damaging the assembly parts. As mentioned above, the fuel rods 22 in the array thereof in the assembly 10 are held in spaced relationship with one another by the grids 20 spaced along the fuel assembly length. Each fuel rod 22 includes nuclear fuel pellets (not shown) and the opposite ends of the rod 22 are closed by upper and lower end plugs 28,30. The fuel pellets composed of fissile material are responsible for creating the reactive power of the PWR 12. A primary liquid moderator/coolant such as water, being held under high pressure such as 2250 psi so that it remains in liquid form, is pumped through the fuel assemblies 10 of the core 16 in order to extract heat generated therein. The primary coolant is circulated by a primary pump 31 around a closed primary path 32 which leads through a first chamber 33 of a heat exchanger 34 where the extracted heat is given up to a secondary flow of water circulated by a secondary pump 35 around a closed secondary path 36 which leads through a second chamber 37 of the heat exchanger 34. The secondary water turns to steam in the second heat exchanger chamber 37 and drives a turbine 38 which generates electricity for the production of useful work. In the operation of the PWR 12 it is desirable to prolong the life of the reactor core 16 as long as feasible to better utilize the uranium fuel and thereby reduce fuel costs. To attain this objective, it is common practice to provide an excess of reactivity initially in the reactor core 16 and, at the same time, provide means to reduce excess reactivity at the early stage of the core operating cycle and then to increase reactivity later on. To partially control the fission process to attain this objective, a number of control rods (not shown) are reciprocally movable in a well-known manner in some of the guide thimbles 18 located at predetermined positions in the fuel assembly 10. Also, as mentioned earlier, the displacement of moderator water to reduce excess reactivity at the early stage of the core operating cycle and then removal of the moderator displacement later to increase reactivity has been a common technique used to assist in obtaining the above-cited objective. The apparatuses of the invention claimed in the third application cross-referenced above and of the present invention, which will be described in detail shortly hereafter, are directed toward implementing a spectral shift on a reusable basis to control reactivity. The apparatuses are associated with the top nozzle 26 and certain other of the guide thimbles 18 not having control rods installed in them. More particularly, the top nozzle 26 includes a transversely extending adapter plate 40 having upstanding sidewalls 42 (the front wall being partially broken away in FIG. 1) secured to the peripheral edge thereof in defining an enclosure or housing 44. An annular flange 46 is secured to the top of the sidewalls 42. Suitably clamped to the annular flange 46 are leaf springs (not shown) which cooperate with the upper core plate (not shown) of the reactor core 16 in a conventional manner to prevent hydraulic lifting of the fuel assembly 10 caused by upward primary coolant flow, while allowing for changes in fuel assembly length due to core induced thermal expansion and the like. SPECTRAL SHIFT APPARATUS AND METHOD As seen in FIG. 1, the spectral shift apparatus, constituting the invention of the third cross-referenced application and being generally designated 48, is disposed within the space defined by the housing 44 of the top nozzle 26 and extends into certain of the guide thimbles 18. Referring also to FIG. 2, the spectral shift apparatus 48 includes a plurality of water displacer rods 50 adapted to be inserted into respective ones of the guide thimbles 18 of the fuel assembly 10 for displacement of a predetermined volume of the moderator/coolant liquid associated with the fuel rods 22. The displaced volume of the moderator/coolant liquid decreases the reactivity (i.e., the hydrogen/uranium ratio) of the reactor core 16 from a given normal level. Each of the displacer rods 50 is composed of an elongated hollow tubular body 51 being sealed at its lower end by an end plug 52 and having a flow opening 54 defined at its upper end. Also, a quantity of water W equivalent to a predetermined small fraction, for instance fifteen percent, of the volume of the rod 50 is contained within each rod. The spectral shift apparatus 48 also includes a manifold, generally designated by the numeral 56, which interconnects the displacer rods 50. The manifold 56 is located on the top of the fuel assembly 10, being disposed within the top nozzle 26 and resting on its adapter plate 40 (see FIG. 1). Referring to FIGS. 2 and 3, the manifold 56 is in the form of a central cylindrical member 58 which defines a central plenum 60 and has a top central inlet opening 62 in flow communication with the plenum. The manifold 56 also has a plurality of hollow tubelike vanes 64 which are mounted on a hub 66 attached to the cylindrical member 58 so as to close the bottom thereof. The vanes 64 extend radially outwardly from the member 58 in flow communication with the central plenum 60 via radial flow channels 67 formed in the hub 66. The vanes 64 have outward ends, defining a plurality of outlet openings 68 of the manifold 56, which are connected to the respective upper ends of the displacer rods 50 so as to dispose the manifold outlet openings 68 in flow communication with the respective flow openings 54 of the displacer rods. Further, in the preferred form, the spectral shift apparatus 48 includes means, generally designated 70, connected to the cylindrical member 58 of the manifold 56 and disposed across its central inlet opening 62 for sealing the same. In particular, the sealing means 70 includes a rupturable disk 72, a baffle screen 74, and an annular cap 76, commercially available per se as standard off-the-shelf items. The disk 72 is rupturable at a given pressure differential across it. For example, in order to avoid rupturing due to pressure variations caused by application of design duty cycle transients which at a maximum are not expected to be above 1600 psi, the disk 72 is designed to fail or rupture only if the pressure differential exceeds 1800 psi. The disk 72, as depicted in FIGS. 2, 5 and 6, is concave shaped and may beprescored to promote rupture in a particular direction, namely, inwardly into the central plenum 60 of the cylindrical member 58. The ruptured state of the disk 72 is shown in FIG. 7. Additionally, to ensure that the disk 72 ruptures inwardly, the baffle screen 74 (see also FIG. 4) is disposed adjacent to it on its outer or upper side. The disk and screen 72,74 are removably and replaceably disposed on the upper end of the cylindrical member 58 in the sealing position seen in FIG. 2 by the annular cap 76 which is internally threaded to receive the external threads on the upper end of the member 58 for removably attaching the cap thereon. Thus, a ruptured disk 72 can at some subsequent time be readily removed and then replaced by a new disk simply by removing the cap 76. Thus, in the preferred form of the spectral shift apparatus 48 as described above, a common sealed plenum approach is adopted wherein all of the displacer rods 50 share a common sealing disk 72. One disadvantage of this approach is that if the single seal should fail prematurely, the entire apparatus, i.e., all of the displacer rods 50, would be flooded. An alternate approach is depicted in FIG. 9 wherein each of the displacer rods 50' is sealed separately. In the latter approach, each of the rods employes its own rupturable disk 72', baffle screen 74' and annular cap 76' which together function the same as before but now only with respect to the one displacer rod 50'. The advantage of this arrangement is that if one of many seals in the apparatus fails prematurely, its effect will be far less noticeable from a power peaking and reactivity insertion viewpoint. However, the common sealed plenum approach is appealing because of the reduced number of parts and simplicity from a refurbishment point of view. The ultimate choice between which of these two different approaches to use should be based on trade offs of seal reliability, power peaking factors, and other licensing issues. Regardless of the approach chosen (i.e., single or multiple seals), the seal is designed to rupture on demand and to provide the desired flooding of the displacer rods 50. The method of carrying out the spectral shift for controlling nuclear reactivity in the reactor core 16, utilizing the above-described apparatus 48, will now be discussed, particularly with reference to FIG. 10. Once an amount of water equivalent to approximately fifteen percent of the rod volume is put into each rod 50 of the apparatus 48 and it is sealed by attaching the rupturable disk 72, the desired number of spectral shift apparatuses 48 are placed on the respective fuel assemblies 10 with their displacer rods 50 inserted within the guide thimbles 18 of the fuel assemblies 10 in the reactor core 16. In such positions, the displacer rods 50 are disposed in the closed primary flow path 32 of the primary moderator/coolant being pumped by the primary pump 31 so as to displace a predetermined volume of the primary moderator/coolant dependent upon the size of the rods 50. In such manner, the spectrum of nuclear reactivity produced by the fuel rods 22 in the fuel assemblies 10 in the core 16 is shifted down or decreased from an excessively high, initial level. Next, the power of the reactor core 16 is elevated to its normal operating level, ordinary by manipulation of control rods (not shown) which are moved relative to certain of the guide thimbles 18. Also, the operation of the primary pump 31 increases the pressure of the primary moderator/coolant in the flow path 32 to the normal operating level. At this level of reactor operation, the pressure of the water within the displacer rods 50 is also elevated. During normal operation the pressure differential across the rupturable disk 72, based on the pressure of the water contained in the sealed displacer rods 50 relative to the pressure of the primary moderator/coolant, will be 1150 psi or less. As mentioned earlier, the disk is designed to withstand pressure variations due to operating transients up to approximately 1800 psi. When rupture of the disk 72, flooding of the rods 50 and, thus, implementation of the reverse spectral shift is desired, the following unique operational transient is imposed, while maintaining the pressure of the primary moderator/coolant at the normal operating level of approximately 2250 psi. The reactor power is decreased to about two percent of the normal operating level, with the turbine 38 being taken off-line and the reactor core 16 being made subcritical by an amount of reactivity equivalent to the maximum reactivity that could be added by rupture of the disk 72 and flooding of the rods 50. The primary coolant temperature is reduced to 375-400 degrees F. by dropping steam pressure. At 400 degrees F., the saturation pressure inside each of the rods 50 is approximately 250 psi and the differential pressure across the rupturable disk 72 is approximately 2000 psi, above the predetermined differential of 1800 psi at which the disk is designed to rupture. Thus, the disk 72 ruptures inward and the displacer rods 50 are flooded, completing the reverse spectral shift which returns the reactivity back to the desired higher level. Means in the form of a flow restricter 78 is disposed in the upper portion tubular body 51 of each displacer rod 50 for restricting the rate of moderator/coolant flow into the body 51 upon rupture of the disk 72. Finally, the reactor core 16 is returned to normal power. To obtain maximum "spectral shift" benefit, as many fuel assemblies as possible should be equipped with the apparatus 48, for instance, all fuel assemblies without installed control rod assemblies. To minimize or eliminate extension to the refueling down-time, refurbishment of the apparatus 48, consisting of draining the rods 50 and replacing the ruptured disks 72, should probably be done most efficiently off-line. For instance, after installing a refurbished apparatus 48 into the reactor core 16, the removed apparatus can be taken to the fuel storage area and made ready for installation during the next refueling stop. Temporary storage of the apparatuses can be made in the stored fuel. An apparatus lifetime of 3-5 cycles appears to be feasible. REFURBISHING APPARATUS AND METHOD Also, in FIGS. 1 and 8, the refurbishing apparatus of the present invention, being generally designated by the numeral 80, is shown associated in operative relationship with the spectral shift apparatus 48 for ensuring its continued use in the fuel assembly 10 or reuse in other fuel assemblies. The refurbishing apparatus 80 used to remove or drain the moderator/coolant from each of the displacer rods 50 basically includes a plurality of first and second ports 82,84 defined on the central cylindrical member 58, a pair of first and second conduits 86,88 extending between the cylindrical member 58 and each respective displacer rod 50 and a generally cylindrical charging tool 90 insertable within the central plenum 60 of the cylindrical member 58. While only two pairs of the first and second conduits 86,88 are seen in FIGS. 1, 2 and 8, it should be understood that there will be a pair of conduits associated with each of the displacer rods. More particularly, the first and second ports 82,84 are vertically displaced from one another and both communicate with the central plenum 60. Each pair of first and second conduits 86,88 in the form of thin elongated capillary tubes interconnect in flow communication a respective pair of the first and second ports 82,84 with respective upper and lower portions 92,94 of a respective one of the displacer rods 50. As seen in FIG. 8, the first and second conduits 86,88 lead from the first and second ports 82,84 in the form of holes in the cylindrical member 58, radially outward along a respective vane 64 to locations aligned above the one displacer rod 50. Then, the conduits are bent ninety degrees to lead downward through the vane 64 and through the upper flow opening 54 of the rod 50. The first conduit 86 terminates contiguous with the upper portion 92 of the rod 50 a short distance below the vane 64, whereas the second conduit 88 leads down through the flow restricter 78, terminating contiguous with the lower portion 94 of the rod 50 a short distance above the lower end plug 52 of the rod 50. The charging tool 90 has first and second flow passageways 96,98 defined therein with corresponding first and second flow openings 100,102. Means in the form of a plurality of O-rings 104,106,107 are provided to adapt the cylindrical member 58 and the charging tool 90 to seal with one another so as to place the first and second flow openings 100,102 in separate flow communication respectively with the first and second ports 92,94 of the cylindrical member 58. Preferably, the O-rings 104,106,107 are disposed on the interior wall of cylindrical member 58 within the central plenum 60 thereof. The one O-ring 104 is located above the first port 92, whereas the other O-ring 106 is located below the first port 92 and above the second port 94 and the third O-ring 107 is located below the second port 94. It will be seen that when the ruptured disk 72 is removed from the cylindrical member 58 and the charging tool 60 is inserted therein, the first conduit 86 and first flow passageway 96 together provide an entrance path for air under pressure to the upper portion 92 of the displacer rod 50. On the other hand, the second conduit 88 and second flow passageway 98 together provide an exit path for moderator/coolant, such as water, in the rod 50 from the lower portion 94 thereof which is separate from the entrance path. Therefore, the water within the rod 50 can be removed by introducing air under pressure along the entrance path into the upper rod portion 92. Such air under pressure forces the water to flow along the exit path up the second conduit 88 from the lower rod portion 94 until the level of the liquid within the hollow rod 50 lowers to the level of the lower end 108 of the second conduit 88. The second conduit end 108 is located a distance from the bottom of the rod 50 which is proportioned relative to the length of the rod to leave a quantity of water W within the rod equivalent to approximately fifteen percent of the rod volume. When the water level in the rod 50 has lowered to the point where air exits the second conduit 88 and the second passageway 98, this is a signal to the operator that all of the water intended to be drained from the rod 50 has been drained. Then, the charging tool 90 can be removed from the cylindrical member 58 and a new unruptured disk attached on its upper end to seal the evacuated displacer rods 50 and the common central plenum 60. It should be readily apparent to those skilled in the art that within the purview of the present invention the hollow tubelike vanes 64 with their radial channels 67 could be substituted for the air conduits 86. The recharging tool 90 would then be similarly modified by locating the air passageway 96 in the lower end thereof for flow communication with the channels 67 of the vanes 64 and with one O-ring being disposed above and another O-ring being disposed below the inlet of channels 67 for separate flow communication of the air and the water whereby air under pressure enters through the channels 67 and the water exits through the conduits 88. Such modified arrangement would eliminate the air conduits 86 and the need for a third O-ring. It is thought that the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
H00009202 | description | A clearer understanding of the present invention will be obtained from the disclosure which follows. DESCRIPTION OF THE PREFERRED EMBODIMENTS Safety experiments were conducted in which there was a need to prevent fission-product cesium from spreading into unshielded but confined portions of the experimental hardware. The experiments were water-reactor safety experiments in which preirradiated, zircaloy-clad, uranium oxide fuel elements were intentionally caused to fail in a steam environment at 750.degree. F. in order to study the migration of fission-product aerosols following cladding failure. Only part of the experimental apparatus was intended to become contaminated, and the remainder was to remain clean and be reusable. After cladding failure induced in Tests No. 1 and 2, the radioactive cesium thus released to the high temperature steam was carried into the reusable section of the system, thereby severely complicating the post-test handling of the experimental hardware. In these first two experiments, stainless steel wire mesh filters (demisters) were provided to hold back the fission-product aerosols. However, they failed to retain all of the fission products carried by the high temperature steam, particularly the cesium isotopes. Thus, post-test disassembly had to be conducted from behind heavy lead shielding. Accordingly, the wire mesh demisters alone did not do the job, even though the cesium which was most likely present in the steam was CsOH, which is a liquid at the operating conditions of the experiment (750.degree. F. steam). Following the first experiment with its extensive spread of radioactive cesium within the closed experimental system, it was recognized that a material must be discovered which would be capable of removing the cesium isotopes from the steam system during test work either by chemical combination or physical combination, or both. When found, it was intended to insert this material, in suitable form, within the circulating system where it could make contact with the 750.degree. F. steam before it entered the reusable components. Accordingly, it was necessary to examine experimentally a number of materials which on theoretical considerations might meet certain necessary criteria, as defined by the objects set forth hereinabove. Testing was done in an apparatus which was believed to be fairly prototypic of a conventional nuclear reactor system. A number of materials were tested, including stainless steel demisting cloth, silicon carbide, crushed quartz, vitreous carbon, crushed borosilicate glass (Pyrex), borosilicate glass beads, glass wool, and a zeolite. All of these materials were found to be capable of removing some of the cesium from the steam. Of these materials, borosilicate glass in the form of solid spherical beads best met the requirements, but only when the beads were heated to about 700.degree. F., and preferably to about 800.degree. F. and higher. The excellent behavior of the borosilicate glass is believed to arise because the cesium forms a stable silicate by combining with the elements in the glass under the test conditions, and also because the cesium and/or its compounds condenses on the exposed glass surfaces. Based upon these results, it was decided to deploy glass beads in the main filter of stainless steel wire mesh (the demister) in the primary steam exit line in the hardware which was utilized in the Tests No. 1 and 2. To make the demister more effective, it was modified to include a bed of 3 mm diameter Pyrex balls or beads which surrounded the stainless steel wire mesh demister. The steel demister element was exactly the same as that used in the two earlier tests. The borosilicate balls were installed so that the effluent steam passed first through the balls and then through the wire mesh demister element. This test work indicated that the removal of the radioactive cesium from the hot steam approaches 90 to 100%. It is believed that this method of stripping fission products from a high temperature vapor medium, such as steam, is unique. Certainly, the use of borosilicate glass, a dense, solid substance, as an agent that acts to combine chemically is unique and not obvious, since glass is normally used because of its inertness to the materials with which it comes in contact. Thus, the practice of the present invention exploits a rare instance where the glass is used deliberately as a chemical reactant. Although the chemical reaction between the radioactive cesium and the silica in the glass is fairly rapid at 700.degree. F., it is preferable that the temperature be at least about 800.degree. F., or even higher, in order to provide for enhanced penetration of the cesium into the borosilicate glass spherical beads in order to achieve a more complete removal of the cesium from the high temperature steam. Clearly, however, the temperature must not be allowed to reach the melting point of the glass beads. It is felt that the method of this invention has application to the steam systems of commercial light water cooled power reactors, particularly those in which the system steam can become contaminated with radioactive cesium due to the breaching of the fuel element cladding. It is contemplated that the inventive method can be utilized as a means for continually cleaning the circulating high temperature steam of radioactive cesium by locating a filter apparatus containing the silica glass spheres at one or more locations in the plant piping of the circulating high temperature steam. Thus, the use of this invention can maintain or improve safety at a nuclear reactor plant by keeping the radioactivity within the circulating high temperature steam at a low level. In light of the foregoing disclosure, further alternative embodiments of the inventive method for the removal of radioactive cesium from high temperature vapors will undoubtedly suggest themselves to those skilled in the art. It is thus intended that the disclosure be taken as illustrative only, and that it not be construed in any limiting sense. Modifications and variations may be resorted to without departing from the spirit and the scope of this invention, and such modifications and variations are considered to be within the purview and the scope of the appended claims. |
047028793 | description | DETAILED DESCRIPTION The present invention relates to an improved pressurized water nuclear reactor having a passive system for in-core spraying of coolant and a passive safety system incorporating the present reactor. Referring now to the drawings, FIG. 1 illustrates an embodiment of the nuclear reactor 1 of the present invention. As illustrated, a substantially cylindrical flow liner 3 has an open top 5, a cylindrical wall section 7 and a bottom 9. The cylindrical wall section 7 has an outwardly directed support flange 11 about the top wall 13 thereof, and at least one inlet port, 15 preferably 2 to 4 thereof, and at least one outlet port 17, preferably 2 to 4 thereof, therethrough. The flow liner 3 is vertically disposed in, and enclosed within a pressure vessel 19, which has an upper, removable, pressure resistant top 21, an intermediate pressure resistant cylindrical wall section 23 having at least one inlet nozzle 25, preferably 2 to 4 thereof, and at least one outlet nozzle 27, preferably 2 to 4 thereof, therein, which are adapted to communicate with the inlet ports 15 and outlet ports 17 of the flow liner 3, and a pressure resistant spherical lower wall section 29 which encloses the cylindrical wall section 7 and bottom 9 of the flow liner 3. The intermediate pressure resistant cylindrical wall section 23 has a ledge 31 about the inner surface 33 thereof, upon which the outwardly directed support flange 11 of the flow liner rests and is supported thereby. The cylindrical wall section 7 of the flow liner 3 forms therein a riser chamber 35 for positioning of the lower reactor internals. Contained within, and spaced from, the cylindrical wall section 7 of the flow liner 3 is an upright cylindrical barrel 37, the barrel supported within the flow liner 3, such as by outwardly directed flange 39 resting on the top wall 13 of the flow liner. The upright cylindrical barrel 37 has a bottom support plate 41 which is spaced from the bottom wall 9 of the flow liner 3. The spaced upright cylindrical barrel 37 and cylindrical wall section 7 of the flow liner 3 form an annular coolant downcomer annulus or passage 43, while the barrel forms the upright riser chamber 35 therein. The lower reactor internals portion 45 contains a nuclear core 47 having a plurality of fuel assemblies 49. Each fuel assembly 49, as is conventional, contains a plurality of elongated fuel rods 51 containing a nuclear fuel that provides a fission-type chain reaction, and a plurality of elongated control rod assemblies 53 contained within elongated thimbles 55 disposed between the fuel rods 51, all of which are located within the riser chamber 35. Mechanisms (not shown) for control of the placement of the fuel rods and control rods are provided in the upper section of the flow liner, as in conventional reactor systems. Pump means 57 are provided to circulate primary coolant which circulates hot primary coolant through port 17 and from the outlet nozzle 27 through line 59, and a steam generator 61, and then returns the same to the inlet nozzle 25 by means of line 63. In operation, hot primary coolant, after heating by passage through the core 47, is discharged through outlet port 17 and outlet nozzle 27, cooled, and returned through inlet port 15 and inlet nozzle 25. From inlet nozzle 25, the cool primary coolant, at an elevated pressure, flows downwardly through the downcomer or annular passage 43, and thence upwardly through the nuclear core 47 and riser chamber 35 where the same is heated, and is then directed to the outlet nozzle 27 for discharge. As illustrated in FIG. 1, the spherical lower wall section 29 of the pressure vessel 19 is spaced from the cylindrical wall section 7 and bottom 9 to form a second annular chamber 65. In effect, the pressure resistant top 21, intermediate pressure resistant cylindrical wall section 23 and pressure resistant spherical lower wall section 29 form a pressure resistant boundary for the nuclear core and primary coolant circulation system. In order to provide for fluid communication between the second annular chamber 65 and the riser 35, a plurality of fluid communication means 67 are provided, which include at least one opening 69 through the flow liner bottom wall 9, an axially aligned opening 71 through the spaced bottom support plate 41 and means for providing flow communication between the openings 69 and 71, such as hollow tubular members /3, one end 75 of which is connected to the bottom wall 9 and surrounds opening 69, and the other end 77 of which is connected to the bottom support plate 41 and surrounds opening 71. At the other end 79 of opening 71, and connected to bottom support plate 41, is an elongated thimble 81 which extends upwardly into the riser 35 within and part of the fuel assembly 49, the thimble 81 being closed at its upper end 83 and having a plurality of spaced apertures 85 therealong, from the bottom of the fuel assembly to its top end 83. A supply of supplementary liquid coolant 87 is contained in the second annular chamber 65 between flow liner cylindrical wall section 7, bottom 9, and pressure resistant wall 29, with flow communication between the second annular chamber 65 and the riser 35 provided through the flow communication means 67. Means 89 for cooling the major portion of the supplementary liquid coolant 87 in the annular chamber 65, to a first elevated temperature, comprises thermal insulation means 89 in the cylindrical wall section 7 and bottom 9 within the pressure resistant spherical lower wall section 29. In FIG. 1, such insulation means comprises forming the wall section 7 and bottom 9 of closely spaced planar sheets 91 (FIG. 3) of material such as stainless steel, with compartments 93 between adjacent spaced sheets, the compartments filled with a liquid 95, such as water. To further cool the major portion of the supplementary liquid coolant 87 in the annular chamber 65, the pressure resistant spherical lower wall section 29 is disposed in an enclosure, such as a container 97 with walls 99 and bottom 101, with a pool of further liquid 103 disposed in the container 97 between the pressure resistant wall 29 and the walls 99 and bottom 101 of the container. While a major portion of the supplementary liquid coolant 87 in the annular chamber 65 is cooled, means 105 are provided to maintain a localized minor portion thereof, indicated at 107, at a second elevated temperature, which is a temperature in excess of that of the first elevated temperature. Such means can comprise exposing an upper region 109 of the pressure resistant spherical lower wall section 29 to the atmosphere without submergence in the pool 103 of liquid coolant, and a layer 111 of a heat insulative material disposed on the outer surface 113 of the exposed upper region 109 of the pressure resistant spherical lower wall section 29. A means 115 is attached to pressure resistant spherical lower wall section 29 to enable injection into, and removal from, the supplementary liquid coolant 87, in second annular chamber 65, of a borated or other chemistry control solution. As illustrated in FIG. 4, means 115 includes an enclosure 117 attached to the pressure resistant spherical lower wall section 29, with a conduit 119 extending through the wall. A pipe 121, connected to a source of chemistry control solution (not shown) extends through the wall 123 of the enclosure and is connected, such as by tube connector 125 to an intermediate pipe 127, with the intermediate pipe 127 connected at its other end by a tube connector 129 to the conduit 119, all within the enclosure 117. A solenoid block valve 131 (fail close) is provided in intermediate pipe 127. A valve 133 is also provided in line 121. The enclosure may be formed of upper section 135 and lower section 137 connected together at flanged portions 139 and 141, respectively, by bolts 143, and with a gasket 145 provided, to permit inspection and repair. The present reactor contains a supply of supplementary liquid coolant in the second annular chamber 65, with the pressure resistant spherical lower wall section 29 comprising the lower primary coolant system boundary. The supplementary liquid coolant 87 is thus at the same pressure, such as about 2250 pounds per square inch absolute (psia) or 155.1 bar, as the primary coolant passing through the flow liner 3, because of the fluid communication means 67. Energy for forcing supplementary coolant into the core, through fluid communication means 67 is provided by allowing a minor portion (less than 10 percent) of the supply of liquid supplementary coolant to be contained at the second elevated temperature, above the first elevated temperature of the major portion of the supplementary coolant supply. The second elevated temperature of a localized minor portion of the supplementary coolant is achieved by means 105, by conduction, and natural convection, with the primary system cold leg temperature (about 288.degree.-293.degree. C.). In the circumstance of a loss of coolant event, the primary reactor coolant system is rapidly depressurized, causing a resultant reduction in pressure in the second annular chamber 65 containing the supplementary liquid coolant 87. When the pressure in the second annular chamber 65 reaches the saturation pressure in the hot localized minor portion 107 of the liquid supplementary coolant 87 (about 1050 psia at 550.degree. F. or 72.4 bar at 288.degree. C.), that localized minor portion will begin to flash into a water/steam mixture. This flashing, with a resultant large volume expansion at the localized portion of the supplementary liquid coolant, drives the coolant major portion thereof downwardly and then into the openings 69 in the flow liner bottom 9, and then upwardly through hollow tubular members 73 and into the elongated thimbles 81, from which supplementary liquid coolant is sprayed through apertures 85 directly into the core 47. Thus, liquid supplementary coolant is charged directly to the core heat generation source rather than puddling it into the downcomer or the bottom of the vessel where much can be lost to a break without any cooling of the core. The present system relies, for practicality, on the ability to maintain a major portion of the supplementary liquid coolant (about 90 percent or more) at the first elevated temperature (about 149.degree. C.) while a minor localized portion (about 10 percent or less) of the supplementary liquid coolant is maintained at the second elevated temperature close to the cold leg temperature (about 288.degree. C.). The insulation means 89 of the cylindrical wall section 7 and bottom wall 9 limits heat loss from the reactor primary coolant system. The external pool of water 103 transfers the thermal and radiation energy lost from the reactor primary coolant system into the supply of supplementary liquid coolant 87 by cooling the pressure resistant spherical lower wall section 29 to about the boiling point of water (about 93.degree.-121.degree. C.). The localized minor portion of the supplementary liquid coolant 107, is above the level of the pool of water 103 and is insulated by layer 111 of heat insulative material. Therefore, the heat loss from the minor localized portion 107 is minimal and the liquid supplementary coolant therein is heated by the losses through the cylindrical wall section 7 of flow liner 3, and by wall conduction from the hot nozzle belt in the intermediate section 23 of the pressure vessel. Since the supplementary liquid coolant is essentially stagnant during normal operation of the reactor 1, stable stratification will occur due to density difference, with hot fluid (about 288.degree. C.) at the top and cold fluid (about 149.degree. C.) at the bottom in the supply of supplementary liquid coolant 87. The supplementary liquid coolant 87 can contain boron as dissolved boric acid to enable a chemical shimming and shutdown of the nuclear reaction. There is little or no mixing between the supplementary liquid coolant and the primary coolant passing through the riser because the supplementary liquid coolant is stagnant water, the interconnecting hollow tube cross-sectional area is small, and temperature/density difference between systems inhibit mixing at the interface. As an indication of the features and design of the reactor vessel, for a 600 MWe pressurized water reactor, the following would be descriptive. The design would employ 145 standard Westinghouse 17.times.17 fuel assemblies with an active core length of 10 feet to produce 1800 MW of thermal energy and about 600 MW of electricity. The reactor coolant system employs soluble boron for burnup compensation and cold shutdown reactivity control. The reactor vessel upper and lower internals is typical of Westinghouse standard design except that the upper internals package will be modified to incorporate a top insertion incore instrumentation system. The core 47 and internals fit inside a 3.4 meter outer diameter core barrel 37 which, in turn, fits inside a 3.7 meter inner diameter cylindrical wall section 7 of a flow liner 3, which is supported from a flange 11 near the top of the vessel 19. The wall section 7 and bottom 9 are made of austenitic stainless steel and are about 8.9 cm. thick. This wall thickness is sufficient to prevent collapsing of the wall due to the external pressure differential which develops during a loss of coolant accident. The wall 7 and bottom 9 are covered by a 7.6 cm. thick layer of thermal insulation. The pressure resistant spherical lower wall section 29 is a 7.3 meter inner diameter spherical wall, about 18.3 cm. thick, made of a low alloy carbon steel designed for 2500 psig (pounds per square inch gauge; or 172.4 bar) at 343.degree. C. The sperical shape reduces wall thickness and weight compared to a cylindrical shape but incurs the disadvantage of increasing the diameter of the reactor. The wall section 29 is clad on the inside with 304 stainless steel. The second annular chamber 65 provides approximately 119 m.sup.3 of supplementary liquid coolant whose temperature is controlled as hereinbefore described. There are 145 tubular connectors 73 penetrating the bottom wall 9 and leading to the thimbles 81 of each fuel assembly. The fuel assembly zircaloy thimbles 81 are 1.1 cm. inner diameter and are preforated over the entire length with 240 evenly distributed 0.13 cm. diameter apertures 85. When a pressure differential between the riser chamber 35 and second annular chamber 65 of about 130 pounds per square inch or 9 bar is developed, 94.6 liters per minute per thimble, or a total of 197.3 kg. per second of supplementary liquid coolant will be sprayed through the apertures 85 in thimbles 81 into the core 47. This will adequately cool the core 47 during a loss of coolant accident blowdown phase. The aforedescribed 600 MWe reactor illustrated in FIG. 1 is inspectable. The flow liner 3 with its integral insulation can be removed and placed in a refueling cavity for inspection and repair. The inside and outside of the pressure resistant vessel 19 are then fully accessible for visual and ultrasonic testing inspection. Another embodiment of the present reactor, using a cylindrically shaped pressure resistant lower wall section 29' is illustrated in FIG. 5. As illustrated therein, the pressure resistant lower wall section 29' is cylindrical in shape and spaced from cylindrical wall section 7 and bottom 9 of the flow liner 3. In this embodiment also, the insulating means 89 is illustrated as a layer of insulating material 147 on the outer surface 149 of the wall 7 and bottom 9. The nuclear reactor of the present invention is readily incorporated into a passive safety system. For example, the reactor with its supplementary liquid coolant supply may be incorporated into a passive safety system such as that described in co-pending application "Passive Safety System For A Pressurized Water Nuclear Reactor" filed Feb. 7, 1986, as Ser. No. 827,115 in the names of L. E. Conway and T. L. Schulz and assigned to the assignee of the present invention, the contents of said application incorporated by reference herein. In incorporating the present reactor into the passive system described in said co-pending application, the two spherical core make-up tanks and their associated piping may be deleted and replaced by the reactor of the present invention, containing supplementary liquid coolant. An embodiment of the passive safety system 151 incorporating the nuclear reactor of the present invention is schematically illustrated in FIG. 6. As illustrated, the nuclear reactor 1 with the components thereof indicated in the drawing with the same numerals as in the previous description of the reactor 1, wherein two outlet nozzles 27, lines 59, steam generators 61, reactor coolant pumps 57, and inlet nozzles 25 are provided. The nuclear core 47 heats primary coolant, or water circulating in the primary coolant system, with heated water supplied through lines 59, or hot legs, to a pair (or more) of steam generators 61. After heat exchange in the steam generators 61, the cooled primary coolant is returned, by means of reactor coolant pumps 57 to the reactor vessel 19 by means of cold legs or lines 63, where primary coolant is directed through downcomer 43 and thence upwardly through the core 47. A pressurizer 153 communicates with hot leg line 59 by means of a conduit 155 and maintains the required pressure in the primary coolant circuit. An in-containment storage tank 157, with a substantial portion of its volume situated above the level of the reactor coolant piping is connected through line 159 to the downcomer 43 of the flow liner 3, which line 159 contains a check valve 161. The check valve 161 is maintained closed so long as smaller pressure prevails at the tank side of the valve 161 relative to the downcomer side. The storage tank 157, contains a passive residual heat removal heat exchanger 163, such as that described in copending application Ser. No. 827,115, the heat exchanger normally submerged in water stored in the tank 157 and having a horizontal intake manifold 165, outlet manifold 167 and interconnecting plurality of heat exchange tubes 169. The inlet manifold 165 of heat exchanger 163 is connected with the hot leg or discharge line 59 by means of line 171, while the outlet manifold 167 is connected to the cold leg or inlet line 63 of the reactor by means of line 173, which line 173 contains a normally closed, fail open throttle valve 175. A line 177 extends from the steam space of the pressurizer 153, line 177 comprising a depressurizing line which opens into the storage tank 157, which line 177 contains a normally closed power operated pressure relief valve 179. Line 177 discharges into the water storage tank 157 through a sprayer 181. A further line 183 is provided to connect hot leg or discharge line 59 to the containment, line 183 containing a normally closed power operated valve 185. A coolant accumulator tank 187 is provided, which is partially filled with water 189, while a space above the water level contains a pressurized gas 191, such as nitrogen. The tank 187 communicates with line 159, leading to the downcomer 43, through line 193 containing a normally closed valve 195. A second such coolant accumulator tank 197, partially filled with water 199, and having a pressurized gas 201, such as nitrogen, communicates through line 203 containing a normally closed valve 205 to the downcomer 43. All of the above components of the passive safety system are located within the containment shield 207. An inlet line 209 may be provided on the in-containment storage tank 157, which is normally maintained closed by a check valve 211, which prevents outflow of water from the in-containment water storage tank 157 but permits water flow, when necessary from the flooded containment into the storage tank 157. The line 183, previously described, prevents long term concentration of boric acid in the reactor following a break of the cold leg by allowing a circulation of water from the containment through the tank 157 (via line 209 and valve 211) into the reactor coolant system through line 159 containing valve 161; the water is heated in the core 47 and exits the reactor coolant system through nozzles 27, the hot leg 59, and then through line 183 containing valve 185. There may be provided, outside the containment shield 207, a steam generator make-up water tank 213 which connects with steam generator 61 through a line 215 containing a check valve 217, and a containment cooling pool 219 containing water 221 to cool the containment shield 207. Steam from the steam generator 61 is directed outside the containment shield 207 through line 223. This whole system is then enclosed in a shield building 225. |
summary | ||
062122509 | claims | 1. A method for providing a leak-tight metal enclosure to a fuel matrix penetrated by coolant channels, wherein the mutually contacting surfaces of said metal enclosure and said fuel matrix are metallurgically bonded, said method comprising the steps of: (a) placing a metal cladding about the lateral surface of said fuel matrix; (b) disposing metal coolant tubes which have been sealed at one end within said coolant channels; (c) placing a first perforated header plate having tubular extensions at that end of the fuel matrix from which the open coolant tube ends protrude, said coolant tubes passing through said first perforated header plate and said tubular extensions and terminating even with the ends of said extensions; (d) placing a second perforated header plate at the other end of said fuel matrix, said sealed coolant tube ends terminating within the perforations substantially even with the outer end thereof; (e) welding, under vacuum, said cladding to said first and second header plates, the open ends of said coolant tubes to the ends of said tubular extensions, and a cover plate over said second header plate; (f) exposing the assembly comprising the fuel matrix and enclosure to a gas at high temperature and pressure; and (g) machining said first and second header plates to provide a finished fuel element. (a) placing a thin continuous sheet of metal cladding about the lateral surface of said fuel matrix; (b) disposing metal coolant tubes which have been sealed at one end within said coolant channels; (c) placing a first perforated header plate at that end of said fuel matrix from which the open ends of said coolant tubes protrude, said header plate having tubular extensions on its face disposed away from said fuel matrix, said tubular extensions being coaxial with the perforations through said header plate and spaced to receive said coolant tubes which terminate substantially even with the ends of said tubular extensions, said metal cladding partially overlapping said first header plate; (d) placing a second perforated header plate at that end of said fuel matrix from which the sealed ends of said coolant tubes protrude, said sealed coolant tube ends terminating within the perforations through said second header plate substantially even with the outer end thereof, said metal cladding partially overlapping said second header plate; (e) welding the ends of said metal cladding to said first and second header plates while under vacuum; (f) welding a metal cover plate over said second header plate while under vacuum so as to seal the end thereof; (g) welding the open ends of said coolant tubes to the ends of said tubular extensions while under vacuum; (h) exposing said fuel matrix and metal enclosure to a gas at high temperature and pressure to effect a diffusion bond between the mutually contacting surfaces of said cladding, coolant tubes, header plates, and fuel matrix; and (i) machining said header plates so as to open said sealed coolant tube ends and achieve a finished fuel element. (a) placing a thin continuous sheet of tantalum cladding about the lateral surface of said tungsten matrix; (b) depositing tantalum coolant tubes which have been sealed at one end within said coolant channels; (c) placing a first perforated tantalum header plate at that end of said tungsten matrix from which the open ends of said tantalum coolant tubes protrude, said header plate having integral tubular extensions on its face disposed away from said tungsten matrix, said tubular extensions being coaxial with the perforations through said header plate and spaced to receive said coolant tubes, said coolant tubes terminating substantially even with the ends of said tubular extensions, said tantalum cladding partially overlapping said first header plate; (d) placing a second perforated tantalum header plate at that end of said tungsten matrix from which the sealed ends of said coolant tubes protrude, said sealed coolant tube ends terminating within the perforations through said second header plate substantially even with the outer end thereof, said tantalum cladding partially overlapping said second header plate; (e) electron beam welding the ends of said tantalum cladding to said first and second header plates while under vacuum; (f) electron beam welding a metal cover plate over said second header plate while under vacuum so as to seal the end thereof; (g) electron beam welding the open ends of said coolant tubes to the ends of said tubular extensions while under vacuum; (h) exposing said tantalum-enclosed tungsten matrix to helium at high temperature and pressure for a time sufficient for a diffusion bond to develop between the mutually contacting surfaces of said tantalum cladding, coolant tubes, header plates, and the uranium fueled tungsten matrix; (i) machining away said tubular extensions together with the coolant tube portions contained therein; and (j) removing said cover plate and machining away a sufficient portion of said second perforated header plate so as to open said coolant tube ends bonded therein. (a) placing a metal cladding about the lateral surface of said fuel matrix; (b) disposing metal coolant tubes within said coolant channels; (c) placing a perforated header plate having tubular extensions at each end of the fuel matrix from which the coolant tube ends protrude, said coolant tubes passing through said perforated header plate and said tubular extensions and terminating even with the ends of said extensions; (d) welding, under vacuum, said cladding to said header plates, and the ends of said coolant tubes to the ends of said tubular extensions; (e) exposing the assembly comprising the fuel matrix and enclosure to a gas at high temperature and pressure; and (f) machining said header plates to provide a finished fuel element. 2. A method for providing a leak-tight metal enclosure to a fuel matrix penetrated longitudinally by a multiplicity of coolant channels, wherein the mutually contacting surfaces of said metal enclosure and said fuel matrix are metallurgically bonded, said method comprising the steps of: 3. A method for providing a leak-tight tantalum enclosure to a uranium fueled tungsten matrix penetrated longitudinally by a multiplicity of coolant channels, wherein said tantalum enclosure is diffusion bonded to said tungsten matrix, said method comprising the steps of: 4. A method for providing a leak-tight metal enclosure to a fuel matrix penetrated by coolant channels, wherein the mutually contacting surfaces of said metal enclosure and said fuel matrix are metallurgically bonded, said method comprising the steps of: |
055984495 | claims | 1. A process for the synthesis of .sup.13 N-ammonia in a target system, in a synthesis apparatus comprising: circulating target water in said target system; charging hydrogen into said synthesis apparatus to create a pressurized condition of 0.5 to 2 kg/cm.sup.2 to maintain the target water in a reducing atmosphere; wherein said circulating is conducted continuously at a speed to prevent the target water in an irradiation cell from changing into an oxidizing atmosphere upon irradiation of a proton beam; and irradiating said target water with a proton beam to produce .sup.13 N-ammonia having a radiochemical purity of 95% or more. purifying target water containing .sup.13 N-ammonia by contacting said target water containing said .sup.13 N-ammonia with a Na-type cation-exchange resin to collect the .sup.13 N-ammonia, and charging a saline solution to elute the .sup.13 N-ammonia collected. 2. The process for the synthesis of .sup.13 N-ammonia in a target system, in a synthesis apparatus according to claim 1 further comprising: 3. The process of claim 1 wherein the amount of target water used provides a hydrogen gas space which maintains the target water in a reducing atmosphere. |
description | This application claims priority under 35 U.S.C. §119 (e) to, and hereby incorporates by reference, U.S. Provisional Application No. 61/237,455, filed 27 Aug. 2009, U.S. Provisional Application No. 61/267,021, filed 5 Dec. 2009, and U.S. Provisional Application No. 61,237,436, filed 27 Aug. 2009. 1. Field of the Invention The present invention relates to LED-UV lamps. More particularly, the invention is suitably used in the application of UV-curing of inks, coatings, and adhesives having UV photo initiators therein. 2. Background UV LED lamps are permanently mounted within the UV-curing process. Depending upon the optics used, the UV LED lamps can be required to be located at a specific distance from the substrate so that uniformity and intensity are optimized. Some UV LED lamps are scalable in length with coarse resolution. UV LED lamps are mounted into a UV-curing process in a manner that makes them difficult and time consuming to remove for cleaning, maintenance, or the like. UV LED lamps are mounted in fixed positions within the UV-curing process where the location within the position is often determined by the process machinery into which the UV curing LED lamps are being integrated. Different positions within the UV-curing process could require the UV LED lamps to be at different locations with respect to the substrate. A conflict could arise between the location required by the optics of the lamp and the location determined by the machinery of a UV-curing process which scenario could render the UV LED lamp unsuitable for placement in particular positions within a UV-curing process. If a lamp is required at an alternate location, either an existing lamp must be uninstalled from an existing location and reinstalled at the desired location which option would only be suitable if the location required by the optics of the UV LED lamp is compatible with the location available in the desired position, or a new lamp must be purchased possibly with redesigned optics. UV LED lamps of different wavelength would also not be easily interchangeable. The coarse resolution in length scalability could result in the scenario where the lamp length options that are available are either too short or too long for a particular application which may make the UV LED lamps difficult or impossible to install into some UV-curing applications. For example, if the length of a UV LED lamp was scalable in 3 inch increments, and a 40 inch lamp was required, the options would either be 39 inches (13×3 inches), or 42 inches (14×3 inches). The 39 inch lamp would be too short and could result in uncured product at the ends of the lamp. The 42 inch lamp could be too long to fit into the envelope that is available within the UV-curing process. LEDs are mounted onto short subassembly segments that may be produced in assorted lengths which segments are then easily mounted into the LED-UV module in a row running along the length of the module. Assembling the LEDs in segments that are easily mounted into the LED-UV module would simplify the process of LED replacement and possibly make the process less expensive. If an LED fails, the segment whereon the failed LED had been assembled can be disconnected, removed, and then a new segment can be installed in its place. The LEDs may degrade as they get older and their output power may decrease below an acceptable level for their application. In this case the owner of the LED-UV lamp would have the option of replacing the segments with new ones as opposed to replacing the whole module. LEDs are solid state semiconductor devices. The efficiency and power output of LEDs can increase from one generation to the next as scientific breakthroughs are made and manufacturing processes improve. The owner of the LED-UV module would have the option to easily upgrade the module by swapping out old segments for new ones with improved operating characteristics. Providing the segments in an assortment of lengths could enable the length of the row of segments to be scalable with a finer resolution than what may be possible if all of the segments where the same length, while at the same time the total number of parts required to assemble the row of LEDs could be reduced. For example, the segments could be configured in a 3 inch version, a 4 inch version, and a 6 inch version. A 12 inch row of segments could then be assembled by connecting 2 of the 6 inch segments. A 13 inch row of segments could be assembled by connecting a 6 inch segment, a 4 inch segment, and a 3 inch segment. A 14 inch row of segments could be assembled by connecting a 6 inch segment and two 4 inch segments. The row of LEDs could be assembled in a variety of lengths with a 1 inch resolution. On the other hand, if only one segment was made, in a 3 inch version for example, the resolution of the possible LED row lengths would be 3 inches, resulting in fewer length options available for customizability. The segment could be made in a 1 inch version to achieve a 1 inch resolution, but doing so could increase the complexity of the assembly by increasing the number of parts required to construct a row. The main module body contains a surface extending the length of the module, whereon the LED segments can be mounted. This surface provides correct positioning and easy mounting of the LED segments. The main module body contains an integral heat sink feature with coolant passages that run the length of the module and are positioned such that they pass near the surface whereon the LED segments mount. The heat sink feature provides a simple means of effectively extracting heat from the LEDs. This maintains the LED junction temperature at an acceptably low level thus maximizing the life of the LEDs. The module is designed so that it is interchangeable and can therefore be quickly and easily installed into or removed from docking ports that are rigidly mounted into a UV-curing process without the use of tools. Interchangeability allows the modules to be easily removed from the UV-curing process for cleaning, repair, maintenance, or the like. LED-UV modules of different wavelengths can be installed into the UV-curing process and the modules can be moved between different locations within the UV-curing process as long as there is a docking port available. Removal and installation of the LED-UV modules from the associated docking ports within a UV-curing process is a tool-less procedure and can be done by a person of no extraordinary skill. All necessary connections (e.g. power, communication, liquid cooling) are made automatically upon installation of the LED-UV module into a docking port, and then disconnected automatically during the removal of the LED-UV module from a docking port. Automatic engaging and disengaging of the connections between the LED-UV module and the docking port upon insertion and removal of the LED-UV module ensure that the connections are made properly, save time, and make the overall operation of the UV-curing process more convenient for the user. All connection devices (e.g. electrical pins, coolant valves) are positioned such that they do not protrude beyond the outer surfaces of the LED-UV module. Designing the connections such that they do not protrude beyond the outer surfaces of the module protects them from damage. With the LED-UV module being designed in a manner that the module is easily removable from the UV-curing process, the possibility of damage to the connections that could result from handling the module will be significantly reduced. The LED-UV module can incorporate a common optical design using a parabolic or elliptical trough reflector that allows for varying distances and mounting locations with respect to the substrate being cured without a significant loss of uniformity or optical (irradiant) intensity. LEDs, by themselves, typically exhibit a Lambertian radiation pattern in which the intensity of the light output by the LED chip is directly proportional to the cosine of the angle between the point of observation and the surface normal of the LED chip. An elliptical or parabolic trough reflector can effectively gather the light and project it onto a substrate that is positioned at varying distances (i.e. from fractions of an inch to several inches) from the base of the LED-UV module with a minimal loss in intensity and in a very uniform manner. Without the use of a reflector, the LED-UV module may need to be placed at either a fixed, optics-dependent distance from the substrate or much closer to the substrate than would be allowable by some UV-curing processes or some positions within a UV-curing process. One example could be in a sheet-fed printing press. In sheet-fed printing, it is typically desirable to locate one or more LED-UV modules immediately following the application of one or more UV-curable inks following the inking units of the printing machine in order to “pin” or “dry” the UV-curable inks or spot varnishes prior to the application of a UV-curable coating at the end of the press prior to the delivery of sheets onto a pile. For inking unit curing locations (immediately following the inking units), it would be desirable to locate the LED-UV modules closer (typically 1 to 3 inches) to the substrate for the benefit of easier mechanical mounting in order to fit within the space constraints provided by various makes and models of printing machines. However, at end-of-press curing locations, the method of sheet transfer provided by most printing machines prohibits closer mechanical location through the end-of-press sheet delivery area and would require the LED-UV module to be mounted as far as 3 to 5 inches away from the substrate. If the LED-UV module were placed too close to the substrate it would collide with the moving machinery of the printing press. The use of reflector style optics enables a single, interchangeable design of the LED-UV module of the invention to be placed in multiple docking or mounting positions at differing distances to the substrate without significant loss of optical uniformity or radiant intensity within a UV-curing process that would otherwise have inaccessible or impractical mounting locations and/or require multiple, non-interchangeable optical designs of the LED-UV modules between the various positions of the process. The LED-UV modules would be available in a variety of UV wavelengths and each wavelength module would be interchangeable with the others and could therefore be applied to any docking port within the UV-curing process. Different types of UV curable products can cure most effectively when irradiated by different wavelengths of UV light. For example, clear products may cure most effectively with longer wavelength UV light, while darker, more heavily pigmented products may cure more effectively with shorter wavelength UV light. Overall system performance may be maximized by the ability to interchange LED-UV modules of different wavelength within the UV-curing process depending upon the preferences of the UV-curable product that is being cured. The LED-UV module could incorporate multiple, adjacent, parallel rows of LEDs where each row shines into a corresponding trough reflector. Incorporating multiple, adjacent, parallel rows of LEDs where each row shines into a corresponding trough reflector would increase the radiant power output by the LED-UV module by a factor equal to the number of rows of LEDs. A single lamp of this embodiment could have the same radiant power output as multiple lamps of the single row embodiment with the added advantages of lower cost and smaller form factor compared to multiple lamps of the single row embodiment. It is understood that the above-described figures are only illustrative of the present invention and are not contemplated to limit the scope thereof. The following is a description of possible embodiments of the LED-UV module of the invention. The examples and figures that follow are intended to teach a person skilled in the art how to effectively design and implement the present invention, but are not intended to limit the scope of the invention. The features and methods disclosed in the detailed description may be used separately or in conjunction with other features and methods to provide improved devices of the invention and methods for making the same. The features and methods disclosed in this detailed description may not be necessary to practice the invention in the broadest sense, but are provided so that a person of skill in the art may further understand the details of the invention. Another description of the LED-UV lamp of this invention, as well as a docking system accommodating such lamp, is present in U.S. patent application Ser. No. 12/868,827, entitled Interchangeable UV LED Curing System, and filed concurrently with this application, the entire disclosure of such application hereby incorporated by reference. Referring to FIGS. 1, 2a-c, and 3, an LED-UV module 100 is shown having electrical connections 102, coolant valves 104, a module body 106, a module cover 108, a connection end cap 110, a cross-over end cap 112, alignment pins 114, a transparent cover 116, a trough reflector 118, coolant passages 120, LED segments 122, and a surface 124 on the module body 106 to mount the LED segments 122. The electrical connections 102 would be located on the connection end 126 of the LED-UV module 100 and may be mounted onto the connection end cap 110. To protect the electrical pins 102 from damage during handling of the LED-UV module 100, the electrical connections 102 could be mounted to the connection end cap 110 in a recessed fashion so that they do not protrude beyond the outer surfaces of the connection end cap 110. The electrical connections 102 would be used to transfer power and possibly communications from the LED-UV module 100 to mating electrical connections that would be present in the docking ports within the UV-curing process. The electrical connections could be pin and socket type connections. The coolant valves 104 would be located on the connection end 126 of the LED-UV module 100 and may be mounted onto the connection end cap 110. To protect the coolant valves 104 from damage during handling of the LED-UV module 100, the coolant valves 104 could be located on the connection end cap 110 such that they do not protrude beyond the outer surfaces of the connection end cap 110. The coolant valves 104 would connect to mating coolant valves that would be present in the docking port and would provide a supply and return for cooling fluid to flow through the LED-UV module 100. The coolant valves 104 and the mating coolant valves in the docking port could be spring actuated poppet style valves that would automatically be pushed open when they are engaged, and automatically spring closed when they are disengaged. The module body 106 would be the main supporting component of the LED-UV module 100. Two significant features on the module body 106 could be the surface 124 that locates the LED segments 122, and the coolant passages 120. The module body 106 could support one edge of the transparent cover 116. The module body 106 could be made of an extrusion out of a material that is a good heat conductor such as aluminum. The module cover 108 would serve as the final component of the LED-UV module 100 assembly and cover all of the internal components. The module cover 108 could contain a feature that would hold the trough reflector 118 in the correct position and shape. The module cover 108 could support one edge of the transparent cover 116. The module cover 108 could be made of an extrusion and the material could suitably be the same as the material of the module body 106. The connection end cap 110 would serve as the mounting structure for the electrical connections 102, coolant water valves 104, and the alignment pins 114. The connection end cap 110 would mount to the appropriate end of the module body 106 forming the connection end 126 of the LED-UV module 100. Power and communications would pass through the connection block 110 into the inside of the LED-UV module 100 through the electrical connections 102. Liquid coolant would flow between the coolant valves 104 and the coolant passages 120 at the interface 128 where the connection block 110 mounts to the module body 106. This interface may be sealed by a gasket such as an o-ring to prevent liquid coolant from leaking at the interface 128. The cross-over end cap 112 would mount to the end of the module body 106 that is opposite the connection block 110 forming the cross-over end 130 of the LED-UV module 100. The cross-over end cap would contain a passage that would connect one of the coolant passages 120 to the other thus forming a circuit for liquid coolant to flow into the LED-UV module 100 through one of the coolant valves 104, through one of the water passages 120, through the passage in the cross-over end cap 112 through the other of the water passages 120, and then out of the LED-UV module 100 through the other of the coolant valves 104. The interface 132 between the cross-over end cap 112 and the module body 106 could be sealed with a gasket such as an o-ring to prevent liquid coolant from leaking at the interface 132. The alignment pins 114 would be located on the connection end of the LED-UV module and may be mounted to the connection block 110. The alignment pins 114 could serve to align the connections 102, 104 prior to their engagement with the mating connections present in the docking port. The transparent cover 116 would most suitably be made of a durable material that would be highly UV transparent such as quartz, glass, acrylic, or the like. The transparent cover 116 would serve as a protective window that would protect the internal components of the LED-UV module while allowing the light generated by the LEDs to pass through the transparent cover. The transparent cover could be supported on one edge by the module body 106 and supported on the opposite edge by the module cover 108. The ends 134 of the transparent cover 116 could be trapped by the connection end cap 110 on one end and the cross-over end cap 112 on the other. The reflector 118 would be made of a highly UV reflective material such as acrylic mirror, polished metal, or the like, and could be formed into shape prior to installation into the LED-UV module 100. The reflector 118 could be held in position and shape by a mating feature in the module cover 108. The reflector could be trough shaped and may incorporate a parabolic or elliptical geometry that would transfer the UV light emitted by the LEDs onto the substrate. The coolant passages 120 would run the length of the module body 106 and be positioned so that they pass near the surface 124 whereon the LED segments 122 mount. The coolant passages 120 facilitate the removal of heat generated by the LEDs and may be dimensioned and located such that the temperature of the module body 106 is essentially uniform over a length of such module. Heat generated at the P/N junctions of the LEDs is conducted from the LED segments 122, into the module body 106 where it is transferred to the liquid coolant by means of convection at the surfaces of the coolant passages 120. The coolant passages could contain fin features 136 that protrude into the liquid coolant. The fin features 136 would serve to increase the convective surface area of the coolant passages 120 as well as generate turbulence in the liquid coolant that would increase the associated convection coefficient. The fin features could also increase the rate of heat conduction through the module body. The presence of fin features 136 in the coolant passages 120 would serve to increase the rate of heat convection from the module body 106 to the liquid coolant, ultimately resulting in lower LED junction temperatures. Lower LED junction temperatures could enable longer LED lifetimes. One embodiment of an LED segment 122 is shown in FIG. 4. The LED segment 122 could consist of a heat transfer plate 138, a plurality of LED packages or segments 140, thermal interface material 142, and fasteners 144 to attach the LED packages 140 to the heat transfer plate 138. The LED packages 140 could be off-the-shelf packages or they could be custom designed. The LED package 140 specifications could suitably be low thermal resistance, high powered UV output, and quick disconnect power terminals 146. The LED segment 122 could contain mounting features such as bolt holes 148 to enable fastening to the module body 106 in a manner that maximizes heat transfer from the LED segment 122 to the module body 106. Multiple LED segments 122 could suitably be mounted to the module body in a lengthwise, end-to-end configuration to form a long row of LEDs. The LED segments 140 would be designed in a manner that maximizes the LED line density (i.e. number of LEDs per inch) and the LED segments 140 could be designed in an assortment of lengths which would enable finer length resolution when assembling the LED segments 140 in a lengthwise, end-to-end configuration to form a long row of LEDs. The finer length resolution would facilitate customizability for a variety of different length UV-curing applications. Providing the segments 140 in an assortment of lengths could enable the length of the row of segments 140 to be scalable with a finer resolution than what may be possible if all of the segments 140 were the same length, while at the same time the total number of parts required to assemble the row of LEDs could be reduced. For example, the segments 140 could be configured in a 3 inch version, a 4 inch version, and a 6 inch version. A 12 inch row of segments 140 could then be assembled by connecting 2 of the 6 inch segments. A 13 inch row of segments 140 could be assembled by connecting a 6 inch segment, a 4 inch segment, and a 3 inch segment. A 14 inch row of segments 140 could be assembled by connecting a 6 inch segment and two 4 inch segments. The row of LED segments 140 could be assembled in a variety of lengths with a 1 inch resolution. On the other hand, if only one segment 140 was made, in a 3 inch version for example, the resolution of the possible LED row lengths would be 3 inches, resulting in fewer length options available for customizability. The segment 140 could be made in a 1 inch version to achieve a 1 inch resolution, but doing so could increase the complexity of the assembly by increasing the number of parts required to construct a row. FIG. 5 illustrates how the implementation of a trough reflector 118 could effectively transfer light 150 from the LEDs onto a substrate 152 at a distance 154 of several inches. This type of optical configuration would be very suitable for UV-curing applications wherein it is not possible to place the LED-UV module in close proximity to the substrate. Some UV-curing applications may require more UV power than an LED-UV module 100 having a single row of LED segments 122 can provide. An alternative embodiment of the LED-UV module 100 of the invention could consist of two or more adjacent, parallel rows of LED segments 122 shining into separate trough reflectors 118. Referring to FIGS. 6, 7a-c, and 8, an LED-UV module 200 is shown having electrical connections 202, coolant valves 204, a first module cover 206, a second module cover 208, a connection end cap 210, a cross-over end cap 212, alignment pins 214, a transparent cover 216, a plurality of trough reflectors 218, coolant passages 220, LED segments 122, a heat sink 224, and surfaces 226 on the heat sink 224 to mount the LED segments 122. The electrical connections 202 would be located on the connection end 228 of the LED-UV module 200 and may be mounted onto the connection end cap 210. The electrical connections 202 would be used to transfer power and possibly communications from the LED-UV module 200 to mating electrical connections that would be present in the docking ports within the UV-curing process. The electrical connections could be pin and socket type connections. The coolant valves 204 would be located on the connection end 228 of the LED-UV module 200 and may be mounted onto the connection end cap 210. To protect the coolant valves 204 from damage during handling of the LED-UV module 200, the coolant valves 204 could be located on the connection end cap 210 such that they do not protrude beyond the outer surfaces of the connection end cap 210. The coolant valves 204 would connect to mating coolant valves that would be present in the docking port and would provide a supply and return for cooling fluid to flow through the LED-UV module 200. The coolant valves 204 and the mating coolant valves in the docking port could be spring actuated poppet style valves that would automatically be pushed open when they are engaged, and automatically spring closed when they are disengaged. The first module cover 206 would cover one side of the LED-UV module 200. The first module cover 206 could contain a feature that would hold one of the trough reflectors 218 in the correct position and shape and the first module cover 206 could support one edge of the transparent cover 216. The first module cover 206 could be made of an extrusion out of a material such as aluminum or plastic. The second module cover 208 would cover the other side of the LED-UV module 200. The second module cover 208 could contain a feature that would hold another of the trough reflectors 218 in the correct position and shape and the second module cover 208 could support the other edge of the transparent cover 216. The second module cover 208 could be made of an extrusion out of a material such as aluminum or plastic. The connection end cap 210 would serve as the mounting structure for the electrical connections 202, coolant water valves 204, and the alignment pins 214. The connection end cap 210 would mount to the appropriate end of the LED-UV module 200 forming the connection end 228 of the LED-UV module 200. Power and communications would pass through the connection block 210 into the inside of the LED-UV module 200 through the electrical connections 202. Liquid coolant would flow between the coolant valves 204 and the coolant passages 220 at the interface 230 where the connection block 210 mounts to the heat sink 224. This interface may be sealed by a gasket such as an o-ring to prevent liquid coolant from leaking at the interface 230. The cross-over end cap 212 would mount to the end of the LED-UV module 200 that is opposite the connection block 210 forming the cross-over end 232 of the LED-UV module 200. The cross-over end cap 212 would contain a passage that would connect one of the coolant passages 220 to the other thus forming a circuit for liquid coolant to flow into the LED-UV module 200 through one of the coolant valves 204, through one of the water passages 220, through the passage in the cross-over end cap 212 through the other of the water passages 220, and then out of the LED-UV module 200 through the other of the coolant valves 204. The interface 234 between the cross-over end cap 212 and the module body 206 could be sealed with a gasket such as an o-ring to prevent liquid coolant from leaking at the interface 234. The alignment pins 214 would be located on the connection end of the LED-UV module and may be mounted to the connection block 210. The alignment pins 214 could serve to align the connections 202, 204 prior to their engagement with the mating connections present in the docking port. The transparent cover 216 would most suitably be made of a durable material that would be highly UV transparent. The transparent cover 216 would serve as a protective window that would protect the internal components of the LED-UV module while allowing the light generated by the LEDs to pass through the transparent cover 216. The transparent cover 216 could be supported on one edge by the first module cover 206 and supported on the opposite edge by the second module cover 208. The ends 234 of the transparent cover 216 could be trapped by the connection end cap 210 on one end and the cross-over end cap 212 on the other. The reflectors 218 would be made of a highly UV reflective material and could be formed into shape prior to installation into the LED-UV module 200. The reflectors 218 could be held in position and shape by mating features in the first and second module covers 206 and 208. The reflectors 218 could be trough shaped and may incorporate a parabolic or elliptical geometry that would transfer the UV light emitted by the LEDs onto the substrate. The coolant passages 220 would run the length of the heat sink 224 and be positioned so that they pass near the surface 226 whereon the LED segments 122 mount. The coolant passages 220 facilitate the removal of heat generated by the LEDs. Heat generated at the P/N junctions of the LEDs is conducted from the LED segments 122, into the heat sink 224 where it is transferred to the liquid coolant by means of convection at the surfaces of the coolant passages 220. The coolant passages could contain fin features 238 that protrude into the liquid coolant. The fin features 238 would serve to increase the convective surface area of the coolant passages 220 as well as generate turbulence in the liquid coolant that would increase the associated convection coefficient. The fin features 238 could also increase the rate of heat conduction through the module body. The presence of fin features 238 in the coolant passages 220 would serve to increase the rate of heat convection from the heat sink 224 to the liquid coolant, ultimately resulting in lower LED junction temperatures. Lower LED junction temperatures could enable longer LED lifetimes. FIG. 9 illustrates how the implementation of a plurality of trough reflectors 218 could effectively transfer light 240 from the multiple, adjacent, parallel rows of LEDs onto a substrate 242 at a distance 244 of several inches. This type of optical configuration would be very suitable for UV-curing applications wherein it is not possible to place the LED-UV module in close proximity to the substrate and where the power of multiple LED-UV lamps 100 is required in a single location. The LED-UV modules 100, 200 of the invention could be produced in an assortment of models where each model would have a different peak wavelength, or could have a plurality of peak wavelengths, in its spectral output depending on the LEDs used in the LED segments 122. To achieve a plurality of peaks in the spectral output of the LED-UV modules 100,200, a mixture of LEDs of different UV wavelength could be used, in an alternating pattern, within each LED segment 122. A single LED-UV module 100,200 with a single peak wavelength in its spectral output is contemplated to be within the spirit and scope of this invention. Additionally, a single LED-UV module 100,200 with a plurality of peak wavelengths in its spectral output is contemplated to be within the spirit and scope of this invention. Having different models of LED-UV modules 100,200 available, each with a different peak wavelength output, or emitting a plurality of peak wavelengths, and where the LED-UV modules 100,200 are interchangeable within a UV-curing process would increase the flexibility of the UV-curing system. Many LED-UV lamps are available in an assortment of UV wavelengths and some with the option of multiple peaks in their spectral output. The LED-UV modules 100,200 of this invention would be designed such that they can be quickly inserted into and removed from a UV-curing process without the use of tools provided that the associated docking ports are mounted into the UV-curing process. A model of an LED-UV module 100,200 of one UV spectral output can be removed and a model of a different UV spectral output can be inserted in a matter of minutes by a person of no extraordinary skill. A person of ordinary skill in the art will readily appreciate that individual components shown on various embodiments of the present invention are interchangeable to some extent and may be added or interchanged on other embodiments without departing from the spirit and scope of this invention. Because numerous modifications of this invention may be made without departing from the spirit thereof, the scope of the invention is not to be limited to the embodiments illustrated and described. Rather, the scope of the invention is to be determined by the appended claims and their equivalents. |
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047088222 | summary | Background of the Invention The present invention relates to a method of solidifying radioactive waste, and more specifically to a method of solidifying radioactive solid waste having a predetermined shape such as that of a pellet. Radioactive waste has heretofore been solidified by mixing dried and granulated radioactive waste into a solidifying material such as a plastic material or concrete. In this case, the solidifying material such as plastic or concrete admixed with the granulated waste could be regarded as a homogeneous material, and the strength of the solidifying material had to be increased simply to increase the strength of the solidified package. In recent years, a method has been proposed in which the granulated waste is pelletized and is then embedded in the solidifying material (Japanese Patent Laid-Open No. 34,200/1977), in order to increase the ratio of waste material embedded, or to reduce its volume. To increase the strength of the material which is solidified by the above method, however, can not be accomplished simply by increasing the strength of the solidifying material. For example, when the solidified package is disposed at sea and is subjected to high pressures, cracks often develop at the boundaries between the solidifying material and the solidified waste embedded therein. SUMMARY OF THE INVENTION An object of the present invention is to provide a method of solidifying radioactive solid waste which is durable and which maintains a sufficiently large safety factor, i.e., which is not destroyed even under increased pressure conditions. Another object of the present invention is to provide a method of solidifying radioactive solid waste so that it is suitable for sea disposal or ground disposal. The method of solidifying radioactive waste of the present invention was achieved by studying the relationship of the modulus of elasticity of the solidifying material and the waste. According to the present invention, the modulus of elasticity of the solidifying material is adjusted to be equal to, or smaller than, that of the radioactive solid waste, in order to prevent stress concentrations at the boundaries between the solidifying material and the radioactive solid waste, particularly on the solidifying material side thereof. Thus the invention makes it possible to prepare a solidified package with a desired durability and safety factor. If a plastic solidifying material is used, the objects of the invention can be accomplished by using a resin with a large distance between crosslinking points. If cement or any other inorganic solidifying material is used, the objects of the invention can be accomplished by adding a rubber-like binder or the like. According to the present invention, solidified radioactive waste is obtained which does not develop stress concentrations within the solidified package even when high pressures are applied thereto, and which does not develop cracks which would lead to destruction, even under highpressure conditions such as on the seabed. |
062333003 | claims | 1. A shroud head to top guide interface in a nuclear reactor comprising a top guide flange, and a shroud head flange configured to couple to said top guide flange, said top guide flange comprising a plurality of frusto-conical shaped guide pins extending from a top surface of said top guide flange, each said frusto-conical guide pin comprising a first base and a second base, said first base having a larger diameter than said second base, said first base of said frusto-conical guide pins located adjacent said top surface of said top guide flange, said shroud head flange comprising a plurality of guide pin openings, each said guide pin opening comprising a frusto-conical shaped portion configured to receive a corresponding frusto-conical shaped guide pin. 2. An interface in accordance with claim 1 wherein a bottom surface of said shroud head flange is in surface to surface contact with said top surface of said top guide flange. 3. An interface in accordance with claim 2 wherein each guide pin opening is defined by an inside surface of said shroud head flange, each said inside surface defining said guide pin opening comprises a frusto-conical portion, said frusto-conical portion opening includes a large base and a small base, and extends from said bottom surface of said shroud head flange. 4. An interface in accordance with claim 3 wherein each guide pin opening further comprises a cylindrical portion, said cylindrical portion extending from said small base of said frusto-conical portion to a top surface of said shroud head flange. 5. An interface in accordance with claim 3 wherein a slope of said inside surface of said shroud head flange defining each said guide pin opening is configured to be equal to a slope of said frusto-conical guide pins. 6. An interface in accordance with claim 5 wherein a diameter of said large base of each said guide pin opening is configured to be larger than a diameter of said large base of each said frusto-conical guide pin so that there is a clearance of less than about 1.0 millimeter between each said inside surface of said shroud head defining said guide pin opening and each said guide pin when the shroud head flange is positioned in surface to surface contact with said top guide flange. 7. An interface in accordance with claim 5 wherein a diameter of said large base of each said guide pin opening is configured to be larger than a diameter of said large base of each said frusto-conical guide pin so that there is a clearance of less than about 0.8 millimeter between each said inside surface of said shroud head defining said guide pin opening and each said guide pin when the shroud head flange is positioned in surface to surface contact with said top guide flange. 8. An interface in accordance with claim 5 wherein each said frusto-conical guide pin comprises a cone angle of between about 20 to 80 degrees. 9. An interface in accordance with claim 8 wherein each said frusto-conical guide pin comprises a cone angle of between about 55 to 65 degrees. 10. A shroud for a nuclear reactor comprising a top guide and a shroud head, said top guide comprising a flange, said shroud head comprising a flange configured to couple to said top guide flange, said top guide flange comprising a plurality of frusto-conical shaped guide pins extending from a top surface of said top guide flange, each said frusto-conical guide pin comprising a first base and a second base, said first base having a larger diameter than said second base, said first base of said frusto-conical guide pins located adjacent said top surface of said top guide flange, said shroud head flange comprising a plurality of guide pin openings, each said guide pin opening comprising a frusto-conical shaped portion configured to receive a corresponding frusto-conical shaped guide pin. 11. A shroud in accordance with claim 10 wherein a bottom surface of said shroud head flange is in surface to surface contact with said top surface of said top guide flange. 12. A shroud in accordance with claim 11 wherein each guide pin opening is defined by an inside surface of said shroud head flange, each said inside surface defining said guide pin opening comprises a frusto-conical portion, said frusto-conical portion includes a large base and a small base, and extends from said bottom surface of said shroud head flange. 13. A shroud in accordance with claim 12 wherein each guide pin opening further comprises a cylindrical portion, said cylindrical portion extending from said small base of said frusto-conical portion to a top surface of said shroud head flange. 14. A shroud in accordance with claim 12 wherein a slope of said inside surface of said shroud head flange defining each said guide pin opening is configured to be equal to a slope of said frusto-conical guide pins. 15. A shroud in accordance with claim 14 wherein a diameter of said large base of each said guide pin opening is configured to be larger than a diameter of said large base of each said frusto-conical guide pin so that there is a clearance of less than about 1.0 millimeter between each said inside surface of said shroud head defining said guide pin opening and each said guide pin when the shroud head flange is positioned in surface to surface contact with said top guide flange. 16. A shroud in accordance with claim 14 wherein a diameter of said large base of each said guide pin opening is configured to be larger than a diameter of said large base of each said frusto-conical guide pin so that there is a clearance of less than about 0.8 millimeter between each said inside surface of said shroud head defining said guide pin opening and each said guide pin when the shroud head flange is positioned in surface to surface contact with said top guide flange. 17. A shroud in accordance with claim 14 wherein each said frusto-conical guide pin comprises a cone angle of between about 20 to 80 degrees. 18. A shroud in accordance with claim 17 wherein each said frusto-conical guide pin comprises a cone angle of between about 55 to 65 degrees. 19. A shroud for a nuclear reactor comprising a top guide and a shroud head, said top guide comprising a flange, said shroud head comprising a flange configured to couple to said top guide flange, said shroud head flange comprising a plurality of frusto-conical shaped guide pins extending from a bottom surface of said shroud head flange, each said frusto-conical guide pin comprising a first base and a second base, said first base having a larger diameter than said second base, said first base of said frusto-conical guide pins located adjacent said bottom surface of said shroud head flange, said top guide flange comprising a plurality of guide pin openings, each said guide pin opening comprising a frusto-conical shaped portion configured to receive a corresponding frusto-conical shaped guide pin. 20. A shroud in accordance with claim 19 wherein said bottom surface of said shroud head flange is in surface to surface contact with a top surface of said top guide flange. 21. A shroud for a nuclear reactor comprising a top guide and a shroud head, said top guide comprising a flange, said shroud head comprising a flange configured to couple to said top guide flange, said top guide flange comprising a plurality of frusto-conical shaped guide pins extending from a top surface of said top guide flange, and said shroud head flange comprising a plurality of frusto-conical shaped guide pins extending from a bottom surface of said shroud head flange, each said frusto-conical guide pin comprising a first base and a second base, said first base having a larger diameter than said second base, said first base of said frusto-conical guide pins extending from said shroud head flange located adjacent said bottom surface of said shroud head flange, said first base of said frusto-conical guide pins extending from said top guide flange located adjacent said top surface of said top guide flange, and said shroud head flange comprising a plurality of guide pin openings, each said guide pin opening comprising a frusto-conical shaped portion configured to receive a corresponding frusto-conical shaped guide pin, and said top guide flange comprising a plurality of guide pin openings, each said guide pin opening comprising a frusto-conical shaped portion configured to receive a corresponding frusto-conical shaped guide pin. 22. A shroud in accordance with claim 21 wherein said bottom surface of said shroud head flange is in surface to surface contact with said top surface of said top guide flange. |
050874101 | claims | 1. A method for avoiding locallized hydrogen buildups in the atmosphere of a safety tank of a reactor, said method including the step of: after a break-down accompanied by loss of coolant has occurred, by utilizing the secondary heat of said reactor adjusting the cooling water temperature in the sump of said safety tank in the phase of the long term cooling to a temperature that is higher than the air temperature in the dome of said safety tank so that said secondary heat used to adjust the cooling water temperature is continuously dissipated to the outside via said atmosphere in said safety tank thereby inducing a convection that effects an adequate intermixing of said atmosphere in said safety tank. 2. A method according to claim 1, in which said adjusting step includes additionally adjusting the temperature in said safety tank sump by regulating a reheating cooler. 3. A method according to claim 1, in which said adjusting step comprises adjusting water escaping from a break in a primary circuit in the phase of the long term cooling such that a water temperature in said safety tank sump of at most approximately 70.degree. C. holds the atmosphere in a lower portion of said safety tank to temperatures that are higher than temperatures of, for example, about 40.degree. C. that exist at the same time in an upper portion of said safety tank. 4. A method according to claim 1, wherein said adjusting step includes operating redundant aftercooler systems. |
042736161 | summary | BACKGROUND OF THE INVENTION This invention relates generally to nuclear reactors, and more particularly to nuclear fuel rods for water-moderated commercial reactors. Current practice in water reactors is to use fuel rods having Zircaloy cladding encapsulating a column of uranium dioxide (UO.sub.2) pellets in fuel assemblies where the fuel rods are spaced apart and supported by a plurality of grids. Typically, several of these grids are positioned axially along the full length of the fuel rods, which is typically 10 to 15 feet. The conventional UO.sub.2 pellet is a solid cylinder with a variety of end shapes, usually dished and chamfered. Recently, commercial nuclear power plant owners have found it advantageous to burn the nuclear fuel in longer cycles to improve the economics in view of the inability to reprocess spent fuel. One fuel pellet design which is known to permit extended burnup is in the form of a hollow cylinder. When stacked in columns of the type required for commercial water reactors, these hollow pellets form a long and continuous central passage within the fuel rod. During operation, the fuel pellets can crack, forming pellet debris which can enter the passageway and accumulate at the bottom of the fuel rod. If this debris deposit is sufficiently large, the redistribution of the UO.sub.2 in the fuel rod could significantly affect the power distribution generated by the fuel rod. This redistribution is not desirable and may in some instances limit the operating level of the reactor. SUMMARY OF THE INVENTION The present invention solves this problem associated with using hollow fuel pellets in modern water reactors by locating spacer plugs at selected elevations along the pellet column to trap pellet debris that may fall through the passage. This prevents accumulation of debris at the bottom of the column, and reduces the shift in power distribution of the fuel rod. The plugs are preferably located at the same elevations as the fuel assembly grids. Because the grids are typically parasitic thermal neutron absorbers, the grid locations usually operate at a depressed power level when compared to areas away from the grids. Thus, the perturbation in the physics characteristics of the fuel rod caused by the presence of the plug is minimized. In another embodiment, the plugs are made of graphite, which is a good neutron moderator. The addition of moderator helps restore some of the thermal neutron flux which has been depressed by the grids. |
056278659 | claims | 1. A nuclear fuel assembly for boiling water reactors, the assembly having a plurality of elongated parallel fuel rods supported between a lower tie plate positioned toward the bottom of the assembly and an upper tie plate positioned toward the top of the assembly; an outer channel surrounding the plurality of fuel rods and having a substantially square cross-sectional area for conducting coolant/moderator about the fuel rods from the bottom of the assembly toward the top of assembly; at least one spacer for positioning and retaining the fuel rods in a predetermined configuration; the fuel rods being arranged with a predetermined pitch in an array of 10.times.12 where the centers of the fuel rods are located at the vertices of equilateral triangles. a plurality of spaced apart part length fuel rods supported at their lowermost ends to the lower tie plate, and extending upwards therefrom into an upper portion of the fuel assembly, a plurality of water rods for conducting coolant/moderator therethrough from the bottom of the assembly toward the top of the assembly; at least one spacer for positioning and retaining fuel rods; an outer channel surrounding the plurality of full length fuel rods and part length fuel rods and having a substantially square cross-sectional area for conducting coolant/moderator about the full length fuel rods and part length fuel rods from the bottom of the assembly toward the top of assembly; the full length fuel rods, part length fuel rods and water rods being arranged with a predetermined pitch in an array of 10.times.12 where the centers of the full length fuel rods, the part length fuel rods, and the water rods are located at the vertices of equilateral triangles. an outer channel surrounding the plurality of fuel rods and having a substantially square cross-sectional area for conducting coolant/moderator about the fuel rods from the bottom of the assembly toward the top of assembly; a center water channel disposed toward the center of the cross-sectional area of the assembly and having at least one wall for conducting coolant/moderator therethrough from the bottom of the assembly toward the top of the assembly; a plurality of water rods for conducting coolant moderator therethrough from the bottom of the assembly toward the top of the assembly; at least one spacer for positioning and retaining the fuel rods in a predetermined configuration; the fuel rods, water rods, and center water channel being arranged with a predetermined pitch in an array of 10.times.12 where the centers of the fuel rods are located at the vertices of equilateral triangles. 2. A nuclear fuel assembly for boiling water reactors, the assembly having: a plurality of elongated parallel full length fuel rods supported between a lower tie plate positioned toward the bottom of the assembly and an upper tie plate positioned toward the top of the assembly; 3. A nuclear fuel assembly for boiling water reactors, the assembly having a plurality of elongated parallel fuel rods supported between a lower tie plate positioned toward the bottom of the assembly and an upper tie plate positioned toward the top of the assembly; 4. The fuel assembly as in claim 3 wherein the center water channel comprises a tube having a diameter greater than the diameter of the fuel rod and each one of the plurality of water rods has a diameter substantially the same as the diameter of the fuel rod. 5. The assembly as in claim 4 wherein the tube which comprises the center water channel has a diameter comprising of two rod pitches and a fuel rod diameter. |
description | This is a continuation, under 35 U.S.C. §120, of copending international application No. PCT/EP2006/007225, filed Jul. 22, 2006, which designated the United States; this application also claims the priority, under 35 U.S.C. §119, of German patent applications DE 10 2005 036 367.9, filed Jul. 29, 2005 and DE 10 2005 037 966.4, filed Aug. 11, 2005; the prior applications are herewith incorporated by reference in their entirety. The invention relates to a control rod for a pressurized water nuclear reactor. The control rod of a pressurized water nuclear reactor essentially consists of a cylindrical sheathing tube in which a cylindrical absorber rod is placed. The sheathing tube is sealed tight toward the outside and filled with a gas such as a noble gas, the filling pressure of which is a maximum of 1.5 bar at room temperature. In the process of operating, in areas with high neutron flux density, i.e. especially in a lower area of the control rod, a volume enlargement of the absorber rod takes place, caused by neutron absorption. This volume enlargement, designated as swelling, and increasing with operational duration, can lead to a radial stretching of the sheathing tube in this area, and in an unfavorable instance to damage to it, so that the control rod has to be replaced well before the end of its computed service life, i.e. at a time when it still has a sufficient neutron-absorbing effect. For reasons having to do with manufacturing techniques, control rods are produced with a diametric gap of about 100 μm between the absorber rods and the sheathing tube, which is the reference design. During operation of control rods, the existing gap is reduced by the sheathing tube creeping downwards, i.e., by a reduction of the diameter of the sheathing tube caused by neutron irradiation and excess pressure, and by swelling of the absorber rod. To reduce the problems mentioned initially, that go along with swelling of the absorber rod that occurs in pronounced fashion in the lower area, and delay closing of the gap, it is known in the state of the art to reduce the diameter of the absorber rod in a lower section to a length of up to about 350 mm, so that there the annular gap increased by an additional 130 μm diametrically. By this means a free space is made available, into which the absorber rod can extend. However, with this measure also, a satisfactory reduction of the stretching of the sheathing tube accompanying the swelling of the absorber rod could not be achieved. It is accordingly an object of the invention to provide a control rod for a pressurized water reactor which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which provides for a control rod in which the degree of sheathing tube stretching caused by swelling of the absorber rod is lessened. With the foregoing and other objects in view there is provided, in accordance with the invention, a control rod for a compressed water nuclear reactor, comprising: an absorber rod disposed in a sheathing tube, said absorber rod having a lower section with a circumferential surface; the absorber rod being formed with a recess, at least in the lower section and on at most one part of the circumferential surface, to form a free space within the sheathing tube surrounding the absorber rod. In other words, the objects of the invention are achieved in that the control rod has an absorber rod that is placed in a sheathing tube and which, at least in a lower section, is formed with at least one recess, that at most occupies one part of the circumferential area of that section. In this the invention is based on the knowledge that the expanded gap that is known in the state of the art, and that extends over the entire circumference and over a larger sectionin the lower part, is linked to significant reduction in heat transmission from the absorber rod to the sheathing tube, and from it to the cooling water, so that in this section the absorber rod becomes considerably more heated. This leads to greater deformation of the absorber rod, which is caused by axial forces acting on the absorber rod when the control rods are in motion, due to great acceleration. This increase in creep deformation is designated as slumping and leads to rapid reduction in the free space obtained, so that it no longer is available, or only partially so, to admit the absorption rod that has swollen due to neutron absorption. According to the invention, not only is a recess made available within the hollow cylindrical sheathing tube that surrounds the absorber rod, into which the swelling absorber can penetrate due to a plastic deformation resulting from the swelling, but also care is taken that in this section the absorber rod is provided with recesses on at most a part of its circumferential surface, i.e., in this section it has no gap that is larger than the reference design. Therefore it is ensured that despite creation of a recess, sufficient heat can be transported outward. In this way, in this area the temperature rises in the absorber only to a negligible degree, so that the increase in creep deformation associated with such a temperature rise plays virtually no role, and the free space produced resulting from the recess is markedly overcompensated. In addition, by creation of a recess, the surface of the absorber rod is increased, and thus its effectiveness is improved. Owing to these measures, the risk of having to replace the control rod well before the end of its calculated service life is reduced. The recess can be formed by a screwlike groove running around on the outer circumference, by an annular groove, by a longitudual groove running along the outer surface of the absorber rod, or by a borehole extending in the axial direction. These design measures can be undertaken either individually or in combination with each other. If, in addition, the control rod is filled with a noble gas, preferably Helium He, the filling pressure at room temperature is measured at greater than 1.5 bar, and especially greater than 10 bar, then owing to the improved heat transmission that accompanies this, the temperature rise in the absorber material in the area of the recess is additionally reduced. Additionally, due to a higher filling pressure, the downward creep, mentioned above, of the sheathing tube is lessened, since the filling pressure acts counter to external pressure. In this it has been shown that even with an increase in filling pressure to about 50 bar, one can expect service life to be increased by an additional 2 to 4 operating cycles. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in control water for a pressurized water nuclear reactor, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, an absorber rod 2 has an essentially cylindrical shape. At its lower and upper ends 4 and 6 it is conically shaped, i.e., provided with a beveled edge 7 and 8, respectively. The absorber rod 2, that can be composed axially of a multiplicity of partial rods, is placed in a sheathing tube 3 that is indicated in the figure by a dash-and-dot line, by which it is surrounded in gas-sealed fashion. By “lower end” what is meant is the end of absorber rod 2 with it, in its installed state and in operation, is inserted, together with the sheathing tube 3 into a control rode guide tube of a burner. In a lower section 9 that adjoins this conical section 7, the absorber rod 2 is formed with a multiplicity of recesses in the form of annular flutes or grooves 10a. In the area of these recesses, the absorber rod 2 thus has a cross sectional surface perpendicular to its long axis that is markedly smaller than the cross sectional surface perpendicular to the long axis of the cylindrical sheathing tube 3 shown with dots and dashes in the figure. In one area 13 between the grooves 10a or adjoining the grooves 10a, the absorber rod 2 has a cylindrical shape and there has a diameter only slightly less than the diameter of the sheathing tube 3, so that in this area 13, only a small gap s, barely visible in the figure, to the sheathing tube 3 exists, on the order of magnitude of about 100 μm. In other words, only in a partial section of the circumferential surface of lower section 9 is the absorber rod 2 provided with recesses. In the exemplary embodiment, a depth d of the grooves is about 1 mm, and their width b is about 2 mm, so that with the seven grooves that are each at a distance of about 1 cm from each other, a free space results with a volume on the order of magnitude of about 270 mm3. Into this free space, the swelling absorber rod 2 can extend without leading to a stretching of the sheathing tube 3. In addition, it can be gleaned from FIG. 1A that the surface on which the absorber rod 2 in section 9 is in contact with the sheathing tube when installed, is only slightly reduced, only by about 10-20% in the exemplary embodiment. Instead of annular grooves 10a, screwlike grooves 10b can also be provided, as is shown in the figure by dashes. FIG. 2 now shows a section from a control rod, in the sheathed tube 3 of which absorber rod 2 according to FIG. 1A, 1B is shown. It can clearly be seen that between the absorber rod 2 and the inner surface of sheathing tube 3, in its areas 13 adjoining the grooves, only a small gap exists, so that heat transport there is good. In the embodiment according to FIGS. 3A and 3B, the recess is formed by a multiplicity of longitudinal grooves 10c running in the axial direction, that extend in the depicted exemplary embodiment over the entire length of the absorber rod 2, so that the absorber rod 2 is symmetrical, and the lower and upper ends 4, 6 can be exchanged. It has been shown in practice that it suffices if the longitudinal grooves 10c extend from the lower end 4 over a length/of about 100-300 mm, since it is only in this area that the neutron loading is very great and results in a pronounced swelling. Between the longitudinal grooves 10c are the areas 13 in which only a small gap exists between the absorber rod 2 and the sheathing tube 3. In the embodiment form according to FIGS. 4A and 4B, the conical areas or the beveled edges 7, 8 are modified by a reduction in the bevel angle α to values between 2° and 30° as well as by an increase in the length h of beveled edges 7, 8, i.e., the height of the conical area is altered to values greater than 1 mm. In this way, an increased free space is created into which the absorber material can swell. The bevel angle α is reduced and at the same time the height h of the conical area is increased. Therefore, the free space produced by these beveled edges 7, 8 can be increased while the front part A of absorber rod 2 remains the same. In this case also, it is sufficient to only modify one of the bevels 7, 8. If only one bevel is modified, for example bevel 7, the lower end 4 of the absorber rod 2 is determined. In other words: the absorber rod can then be placed only in one direction in the sheathing tube of the control rod. Additionally in the figure, at the lower end, a recess is made in the form of a central axial borehole 10d, which likewise serves as a free space and can be implemented in addition to, or alternatively to, the measures explained above and below. Typically such a borehole 10d has a diameter D of about 3 mm and a depth T of about 50-100 mm. If such a borehole 10d is made, care must be taken that the front part A, which simultaneously is the contact surface for absorber rod 2 on an interior surface of the sheathing tube, is not reduced by an appropriate configuration of the bevel 7. The measure depicted in FIG. 4A, 4B can be used also in combination with the grooves 10a, 10b running around as depicted in FIGS. 1A, 1B, or the longitudinal grooves 10d depicted in FIGS. 3A, 3B. Instead of the grooves, flutings or axial borehole shown, recesses can also be provided with other geometric forms, such as trough-shaped indentations or holes. The only thing that is primarily important is that additional free space results, into which the absorber rod can swell, and that they occupy at most a part of the circumferential surface of the lower section. |
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abstract | A shielding assembly for use in a semiconductor manufacturing apparatus, such as an ion implantation apparatus, includes one or more removable shielding members configured to cover inner surfaces of a mass analyzing chamber. The shielding assembly reduces process by-products from accumulating on the inner surfaces. In one embodiment, a shielding assembly includes first and second shielding members, each having a unitary construction and configured to cover a magnetic area in the mass analyzing chamber. The shielding members desirably are made entirely of graphite or impregnated graphite to minimize contamination of the semiconductor device being processed caused by metal particles eroded from the inner surfaces of the mass analyzing chamber. |
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abstract | A method of performing a medical procedure includes providing a radiation-shielding cubicle having an interior defining a medical personnel region and including a first wall having an opening therein, locating the cubicle with respect to an x-ray table so a portion of the x-ray table extends through the opening into the interior of the cubicle, and separating medical personnel from an x-ray emitter disposed outside of the cubicle using the first wall to shield the medical personnel from radiation emitted by the x-ray emitter. |
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description | The invention relates to a method of determining the crystallographic structure of a crystal by electron diffraction tomography using an electron microscope. This method is known from “Towards automated diffraction tomography: Part I-Data acquisition”, U. Kolb et al., Ultramicroscopy 107 (2007) 507-513. There is great interest in determining the structure of macro-molecules, such as catalysts, proteins, viruses, DNA and RNA. The knowledge is of importance for understanding how e.g. proteins operate, how to produce, for example, more effective medicaments, enzymes, etc., and for example to understand why certain illnesses occur. A group of techniques known as crystallography is used to determine the structure of molecules, of which X-ray crystallography is the most well-known. Here a multitude of diffraction patterns is recorded by irradiating a crystal by a beam of X-rays, and a diffraction pattern of said beam is recorded. A disadvantage of X-ray crystallography is that the size of the crystals must be rather large, e.g. 0.1 μm or more, because the interaction between the crystals and the X-ray beam is small. For many inorganic crystals this is not a problem, as these can easily be grown to a size of 0.1 μm or more, but it proves to be extremely difficult to grow crystals of e.g. proteins to such a size. X-ray diffraction is thus less suited for determining the structure of e.g. proteins. The interaction between a beam of accelerated electrons, as used in e.g. an electron microscope, and the atoms of a crystal is much larger than when using X-rays. Therefore diffraction patterns of nano-crystals, with a diameter of less than 1 μm down to several nm, can be recorded with, for example, a transmission electron microscope (TEM). In the known method described by U. Kolb, three-dimensional (3D) diffraction data are collected by manually tilting a crystal around a selected crystallographic axis and recording a set of diffraction patterns (a tilt series) at various crystallographic zones. In a second step, diffraction data from these zones are combined into a 3D data set and analyzed to yield the desired structure information. It is noted that data collection can be performed automatically. This involves a software module for a TEM enabling automated diffraction pattern collection while tilting around the goniometer axis. Kolb then proceeds to describe such a software module for a TEM, combining Scanning Transmission Electron Microscopy (STEM) imaging with diffraction pattern acquisition in nanodiffraction mode. It allows automated recording of diffraction tilt series from nanoparticles with a size down to 5 nm. In the introduction Kolb teaches that the diffraction patterns can be recorded by illuminating the crystal with area selecting, the so-named Selected Area Electron Diffraction (SAED) technique, in which an aperture downstream of the diffraction plane is used to limit the part (the area) of the sample contributing to the diffraction pattern. The beam can be a convergent, focused beam (CBED), a substantially parallel beam, or any convergence angle therein between. Parallel illumination can be obtained by Köhler illumination. Alternatively a small aperture, known as the C2 aperture, can be used to decrease the beam diameter to a few nanometers while keeping the beam almost parallel. Kolb proceeds to describe that working in TEM mode with a small beam of typically 50 nm diameter makes it nearly impossible to position the beam with any degree of accuracy on a crystal that is larger than the beam. Therefore the position of the crystal is determined in STEM mode. It is noted that Kolb mentions that in principle the diffraction patterns can be recorded using a more or less parallel beam, but fails to give an example of this. On the contrary, she proceeds showing Convergent Beam Electron Diffraction, and e.g. at page 509 of her article, lower right corner, says that the diffraction pattern is not focused in the back-focal plane. A disadvantage of said method is that not all TEM's are equipped with a scanning unit, as a result of which not all TEM's can operate in STEM mode. There is a need for a method that can be performed on an instrument that is not equipped with a scanning unit in order to operate in STEM mode. The invention describes a method for electron diffraction tomography in a Transmission Electron Microscope. Known methods involve using Scanning Transmission Electron Microscope, and use the scanned beam for STEM diffraction. The invention proposes to form the diffraction patterns with a stationary beam (200) with a diameter slightly larger than the crystal, as a result of which a TEM without STEM unit can be used. Finding the crystal is done in TEM mode. Advantages of the method according to the invention are: a TEM without scanning unit can be used, and the diffraction volume is not depending on the orientation of the crystal, as the whole crystal is illuminated while obtaining the diffraction pattern. The foregoing has outlined rather broadly the features and technical advantages of the present invention in order that the detailed description of the invention that follows may be better understood. Additional features and advantages of the invention will be described hereinafter. It should be appreciated by those skilled in the art that the conception and specific embodiments disclosed may be readily utilized as a basis for modifying or designing other structures for carrying out the same purposes of the present invention. It should also be realized by those skilled in the art that such equivalent constructions do not depart from the spirit and scope of the invention as set forth in the appended claims. The method according to the invention is characterized in that the beam used for recording the diffraction patterns is a substantially parallel beam having a diameter larger than the size of the crystal. By using a beam with a diameter larger than the crystal, and keeping this beam stationary with respect to the crystal while recording the diffraction pattern, the TEM used does not need to be equipped with a scanning unit for scanning the beam over the crystal. As the diameter of the beam is larger than the diameter of the crystal, the interaction volume, or scattering volume, also known as diffraction volume is the volume of the crystal itself, and thus for all tilt angles the same. This eases the normalization and post-processing otherwise needed when recording the diffraction patterns and/or analyzing the recorded diffraction patterns. In an embodiment of the method according to the invention the centering of the crystal with respect to the beam involves shifting the beam and/or mechanically moving the crystal. By moving the beam over the sample, e.g. by applying a magnetic or electrostatic field, the beam can be positioned over the crystal. For large displacements also a mechanical movement of the sample can be performed. In a preferred embodiment of the method according to the invention the beam used for recording the diffraction pattern is a substantially parallel beam. When tilting the sample the position of the sample will often change, which can be corrected by shifting the beam. When using a parallel beam tilting of the sample and/or shifting the beam does not shift the origin of the diffraction pattern, nor will it distort the diffraction pattern. This is advantageous when compared to the work described by U. Kolb, where the diffraction pattern is formed with a convergent beam, resulting in disks rather than spots in the back focal plane of the objective lens. For both the convenience of later stage data processing and more importantly to separate each reflections clearly (for macro-molecular crystals, the reflections can be very dense and disks are easily overlapped with each other), an extra step has to be taken to re-focus the disks into spots by any of the three methods: 1). vary the current of the objective lens; 2). vary the current of the projection system; 3). vary the current of a third lens, called a diffraction lens on FEI microscopes. The consequence is that the diffraction pattern will shift with the beam shift. This is also acknowledged in the work by U. Kolb in page 509 and 510. It is noted that other methods of crystallography are known, such as, but not necessarily limited to, X-ray single crystal diffraction, X-ray powder diffraction, 2D electron diffraction, and Precession Electron Diffraction. X-ray single crystal diffraction methods have as draw-back the large crystals needed, as discussed previously. X-ray powder diffraction method has the difficulty of offering a unique or reliable solution, that is: it is difficult or impossible to determine a unique crystallographic structure. This is due to the fact that rotational information in the diffraction pattern is lost as a result of the diffraction patterns of a multitude of crystals. 2D electron diffraction demands a very special type of crystal (a 2D monolayer) which is very difficult to produce. Precession Electron Diffraction demands a TEM capable of precession (rotating the beam in a cone, the center of the cone on the crystal) and cancelling the rotation of the beam again after passing the crystal, so that a stationary diffraction pattern is formed. This demands special TEM's.Therefore all these other methods have drawbacks with regard to the limitations of the method itself, the crystals, or the instruments used. In an embodiment of the method according to the invention the crystal has a largest diameter less than 10 μm, more specifically less than 1 μm, most specifically less than 100 nm. Making such small crystals is much easier than making the large crystals needed in, for example, X-ray diffraction, typically having sizes of 100 μm or more. This especially holds for proteins. In yet another embodiment of the method according to the invention the crystal is a crystal of macro-molecules from the group of catalysts, proteins, viruses, DNA and RNA. Identification of the crystallographic structure of e.g. proteins is very much in the interest, both for industrial processes (e.g. for synthesizing enzymes) and in healthcare (e.g. for synthesizing drugs) In still another embodiment of the method according to the invention a multitude of crystals is identified, and for each tilt angle a multitude of diffraction patterns is recorded, each diffraction pattern associated with one of the crystals. Recording diffraction patterns for a number of crystal using the above mentioned method with identical crystallographic structure results in better results due to a better signal-to-noise ratio, and because more crystallographic directions are probed, as each of the crystals is likely to have a different orientation with respect to the beam. This can be realized with one tilt series, and therefore with a limited amount of (mechanical) steps for tilting the crystals. It is noted that in this method for each of the crystals a series of diffraction patterns is made and analyzed. Therefore this method differs from analyzing poly-crystalline material. It is further noted that this method can be used to acquire data from two or more crystals with different crystallographic structure and/or composition as well, and analyze the two or more sets of data separately. This results in a higher throughput, as only one mechanical tilt series is made, and also the centering can be realized more efficiently by determining the position of one crystal or one set of features (the mutual positions known), thus saving acquisition time. In still another embodiment of the method according to the invention the electron microscope is a cryo electron microscope and the diffraction patterns are recorded while the sample is at a cryogenic temperature. The environment within an electron microscope is a harsh environment, with high levels of radiation and vacuum. As known to the person skilled in the art the “lifetime” of molecules at cryogenic temperatures is much larger than when the molecules are studied at room temperature. It is noted that TEM's equipped to operate at liquid nitrogen temperature and/or liquid helium temperature are readily available. In still another embodiment the sample is mounted on a tilt holder from the group of single-tilt holders and double-tilt holders, and the tilting is the result of tilting the tilt holder. To position a sample in a TEM the sample is normally mounted on a grid, for example a copper grid with a diameter of 3.05 mm, and said grid is in turn mounted on a holder. The holder is then inserted in a so-named goniometer, which seals against the holder, while simultaneously enabling movement of the sample at the sample position. Some holders enable (in cooperation with a goniometer) tilt, some enable tilt in two directions. Also holders enabling cryogenic use, and/or enabling heating, etc. are known. For the analysis needed here a single tilt holder suffices. It is noted that the holder used is typically a side-entry rigid holder that is tilted by the so-named goniometer. The goniometer is typically capable to shift the holder in the x, y and z direction and rotate the holder in one direction, It is further noted that often a so-called tomo-holder is used, that is essentially a single tilt holder equipped for large tilt angles without touching the pole pieces of the (magnetic) lenses of the microscope and having provisions for not intercepting the incoming and outgoing beam of electrons even at a high tilt angle (typically 60 to 80 degrees). It is mentioned that holders of other types, such as top loading holders, or double tilt holders, are known, as well as holders where the rotation/translation of the sample is realized by, for example, piezomotors at the tip of the holder. In still another embodiment the diffraction pattern is formed using an objective lens and during at least part of the tilt positions the position of the crystal is centered with respect to the beam and the objective lens using a model of the movement of the tilt holder with respect to the beam and said objective lens, as a result of which the crystal is not exposed to electrons during the centering of the crystal. If the accuracy/reproducibility of the position of the holder and the goniometer are sufficient, the crystal can be positioned at least part of the time by ‘dead-reckoning’. This minimizes the exposure of the crystal to electrons, and thus minimizes damage during centering of the crystal. In still another embodiment centering the crystal involves imaging at least part of the sample using a beam of electrons. The TEM is capable of imaging the sample with great accuracy, and thus the position of the crystal can be determined with high precision with respect to the beam. In a further embodiment the imaging of the sample involves imaging the crystal. By imaging the crystal, usually at a relatively low magnification which limits the dose to the crystal, its position is recorded with high accuracy. It is noted that the electron dose per unit area need not be large, and thus the amount of electrons to which the crystal is exposed can be low. In another embodiment the position of one or more features in the sample relative to the crystal is determined before making the tilt series, and the imaging of the sample involves imaging the one or more features, and the position of the features is used to center the crystal. Here first the position of the crystal with respect to one or more features is determined. During the tilt series the position of the crystal can now be derived by determining the position of the one or more features. In this way the crystal is not exposed to electrons while centering, It is noted that when one feature is used to determine the position of the crystal preferably both the crystal of interest and the feature are located at or near the tilt axis, but that when two or more features are used, neither the features nor the crystals need to be located on the tilt axis (although the position with respect to the tilt axis should be known to make a model describing how the features and crystal move due to the rotation. In still another embodiment the diameter of the beam used for centering the crystal differs from the diameter of the beam used for recording a diffraction pattern. Preferably the beam diameter for the centering is larger than the beam diameter used for diffraction, so that during centering a large field of view is available, while the beam used for recording the diffraction pattern is only slightly larger than the crystal. In still another embodiment the crystal is during the recording of the tilt series exposed to a dose of less than 105 electrons/nm2 By exposing the crystal to a dose less than 105 electrons/nm2 for the sum of all the tilt positions in the tilt series (the accumulated dose during the whole series), the damage to the crystal is limited. See also the publication by Kolb. In still another embodiment the crystal is during the recording of the tilt series exposed to a dose rate of less than 300 electrons/(nm2 s). As mentioned by Kolb, also the dose rate needs to be controlled to a low value of, for example, less than 300 electrons/(nm2 s). FIG. 1 depicts schematically optical elements for a TEM performing the method according to the invention. FIG. 1 shows an electron source 100 for producing a beam 102 of energetic electrons with an energy of, e.g. between 50 and 400 keV, along electron- optical axis 104. It is noted that in reality the position where the beam is focused (shows cross-over's) differs from the sketched positions—and thus angular and linear magnifications differ—, but these cross-over's are used to limit the beam diameter. It is further noted that electron microscopes using lower and higher beam energies are known. It is noted that one or more lenses between the electron source and the aperture may be present, as well as alignment coils to center the beam on the axis. Condenser lenses 108 and 110 are used to form a beam on sample position 112. The diameter of the beam at the sample position is governed by aperture 106. A sample mounted on sample positioning unit, the so-named goniometer 114, is placed on said sample position, the goniometer enabling positioning the sample on the sample position along any of axis x, y, z and rotating the sample along the x-axis. Objective lens 116 with a back-focal plane 118 images the sample, and projection lenses 120 and 122 form an enlarged image on imaging plane 124, which may be a fluorescent screen, or the plane where a camera system resides. It is noted that the sample may be immersed in the (magnetic) field of the objective lens 116. In that case the objective lens can be thought to be split in two parts, one cooperating with the condenser lenses 106 and 108 to illuminate the sample and a second part cooperating with the projection lenses 118 and 120 forming an image. A TEM can image a sample in different ways. Two important modes are: Diffraction mode: in diffraction mode the sample is illuminated with a, preferably parallel, beam of electrons, as a result of which a diffraction pattern is formed in the back-focal plane of the objective lens (all parallel rays are focused in this plane, the position where the focus is formed dependent only on the angle with which the electrons leave the sample plane), and the projection lenses form an enlarged image of this back-focal plane on the image plane. TEM imaging mode: in TEM imaging mode the sample is illuminated with a beam of electrons (that may be a parallel beam). The projection lenses do not image the back-focal plane of the objective lens, but the sample plane on the imaging plane (for example a florescent screen or a camera). The image is formed by intensity variation resulting from a part of the electrons being absorbed in the sample, and electrons diffracted (scattered) in the sample interfering with electrons that pass the sample unhindered. It is noted that a Scanning Transmission Electron Microscope resembles a TEM, but is additionally equipped with deflection coils between lens 108 and the sample, and focuses the beam on the sample. By then scanning the focused beam over the sample with these deflection coils, a scanning image is made, using detectors placed under the sample (at the side removed from the electron source). FIG. 2 schematically shows the ray diagram near the sample in diffraction mode. FIG. 2 shows a parallel beam of electrons 200 impinging on a sample 204. The sample comprises a crystal 202, causing the beam 200 to split in an undiffracted beam and a diffracted beam 206. Objective lens 116 focuses both the undiffracted and the diffracted beam in diffraction plane 118, as both beams are parallel beams. It is noted that, for perfectly parallel beams and an objective lens without lens aberrations, the foci formed in diffraction plane 118 are points. In reality the spots have a small but finite diameter due to beam convergence/divergence, mainly as a result of the finite diameter of the source. As is clear from FIG. 2, a beam with a diameter larger than the crystal results in a constant diffraction volume: the volume of the complete crystal. It is noted that for the method of the invention normally some adjustments are made before recording the tilt series. These are: the so-named camera-length is determined. This parameter describes the magnification of the diffraction plane to the image plane (camera or screen). alignment of the beam to illuminate the same area on the sample both when recording a diffraction pattern and when imaging the Image/beam shift has to be calibrated at each Magnification. the position of the tilt axis of the stage has to be determined FIG. 3 shows a diffraction pattern. Clearly a strong central peak is visible, as well as a multitude of sub-peaks. The central peak is the result of the focusing of the electrons that pass through the sample unhindered, and each of the sub-peaks corresponds with electrons that are scattered under a specific angle with respect to the incoming beam. The more complex a crystal is (that is: the more atoms there are in a unit cell), the more complex the diffraction pattern is: the more spots are visible. Also: the more complex a crystal is, the more weak spots are present. It is noted that this diffraction pattern shows symmetry around the central spot, but that for most tilt angles the diffraction pattern does not show symmetry. It is noted that, for example, “Collaborative Computational Project No. 14” (CCP14) resulted in a suite of software packages to analyze diffraction patterns, see the CCP14 website http://www.ccp14.ac.uk/about.htm. This and other packages are well known to the person skilled in the art. Although the present invention and its advantages have been described in detail, it should be understood that various changes, substitutions, and alterations can be made herein without departing from the spirit and scope of the invention as defined by the appended claims. Moreover, the scope of the present application is not intended to be limited to the particular embodiments of the process, machine, manufacture, composition of matter, means, methods and steps described in the specification. As one of ordinary skill in the art will readily appreciate from the disclosure of the present invention, processes, machines, manufacture, compositions of matter, means, methods, or steps, presently existing or later to be developed that perform substantially the same function or achieve substantially the same result as the corresponding embodiments described herein may be utilized according to the present invention. Accordingly, the appended claims are intended to include within their scope such processes, machines, manufacture, compositions of matter, means, methods, or steps. |
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047298693 | summary | BACKGROUND OF THE INVENTION 1. FIELD OF THE INVENTION The present invention relates to facilitating the maintenance of nuclear power systems and particularly to minimizing the exposure to radioactivity of service personnel working within the steam generator of a nuclear power system during the performance of routine periodic maintenance thereon. More specifically, this invention is directed to a radiation shielding system employing shielding modules or tiles which may be rapidly and easily installed in steam generators. 2. DESCRIPTION OF THE PRIOR ART While not limited thereto in its utility, the present inventive system has been designed for use in, and has special utility for, the servicing of the steam generators of nuclear power systems, particularly such systems which employ pressurized water type reactors. Such steam generators comprise a pressure vessel having a lower plenum area which is a hemisphere of five to seven feet radius divided into two halves. A coolant, which has been heated in the reactor, is delivered to one of the plenum halves and is then circulated, via a bundle of tubes, which may contain from 3,000 to 11,000 tubes supported by a horizontal tube sheet, through the steam generator vessel. The coolant is subsequently discharged from the other plenum half and returned to the reactor. During passage through the steam generator tube bundle, heat from the reactor coolant is transferred to water under pressure which subsequently flashes to steam for driving a turbine. Nuclear Regulatory Commission rules require periodic inspection of the steam generators of nuclear power systems. To this end, the steam generator vessel is provided with access openings, known in the art as "manways", in the lower plenum area. During normal operation these access openings are sealed by means of covers. In order to perform inspections, after the vessel has cooled and the "primary loop" has been drained of reactor coolant, the manway covers are removed. Once access to the interior of the steam generator pressure vessel is possible, a number of different procedures may have to be performed within the vessel. However, since the interior of the steam generator vessel is classified as a highly radioactive environment, maintenance personnel may work in the lower plenum area for only short periods of time. The types of procedures which may have to be performed from within the lower plenum area of a stem generator vessel include non-destructive testing, steam generator tube pulling, steam generator tube plugging, installation of sleeves in steam generator tubes and the installation of nozzle dams to prevent backflow of coolant from the reactor pressure vessel should it be necessary to flood the reactor in order to perform separate operations thereon. The non-destructive testing will typically comprise ultrasonic and/or eddy current examination of the interior of a preselected percentage of the steam generator tubes. For example, 3% of the tubes will be tested to determine if there is any reduction in effective wall thickness, i.e., cracks, pits, or corrosion, of greater than 20% or any growth in effective wall thickness, i.e., scaling or other deposits, of more than 20%. If a preselected number of the tested tubes are found to exceed the set limits for increased or reduced wall thickness, an additional percent of the tubes will be tested. Steps must be taken to minimize the radiation exposure of service personnel who are working under the tube sheet in the lower plenum area. Prior attempts to provide the requisite radiation shielding in the lower plenum area below the tube sheet have largely been limited to hanging lead blankets under the open tube ends and tube sheet on scaffold-like racks. This has not proven to be a satisfactory procedure since the racks and blankets take a long time to install, and the installers are subjected to radiation during the installation. The recent development of a new "Rapid Installation Tube Gripper" invention has provided a fastening device which made the modular radiation shielding system of the present invention practical. The gripper was invented by Glenn E. Schukei, one of the joint inventors of the instant invention, and Robert J. Schukei, and is the subject of U.S. patent application Ser. No. 686,114, filed Dec. 24, 1984. The gripper device can be rapidly attached to the inside of a steam generator tube to secure a modular lead shielding tile. The device uses hard balls which are held in position by a sleeve so that the balls can be wedged between a tapered shaft and the wall of the tube when an attempt is made to withdraw the tapered shaft. The harder the shaft is pulled, the tighter the balls are wedged, thus insuring that they cannot be inadvertently withdrawn. To release the device, the tapered shaft is inserted further into the tube than the sleeve, so that the balls are relocated relative to the tapered surface. The released position of the balls and tapered shaft is maintained by the sleeve until the device is removed from the tube. A single worker is permitted to stay in the lower plenum without shielding for only two or three minutes, during which time he may be exposed to his three month radiation dose limit. From this it can be seen that any system that can save even a few seconds or can shield any significant radiation is quite valuable. SUMMARY OF THE INVENTION The present invention overcomes the above-discussed and other deficiencies and disadvantages of the prior art and in so doing comprises a novel modular radiation shielding system for providing extended periods of worker access to a radioactive environment. Specifically, the present invention comprises a novel radiation-shielding modular sheathed lead tile assembly which, through the use of rapid installation tube gripper fastening components, may be employed to allow extended periods of access by service personnel to the radioactive environment in the plenum under the tube sheet and tube bundle it supports. The tubes' inside surface contribute the major portion of radiation in the primary head or plenum. This area of the steam generator is highly contaminated and frequently has radiation levels of 20 to 30 rem at the tubesheet with perhaps a general field of 10 rem in the head area. As can be seen, a high percentage of the incident radiation comes from the steam generator tubes and tubesheet. This is caused by the large surface area inside tubes which have a thin layer of "crud" that causes radiation to beam down to the head work area. In fact, the contaminated surface of only 6 inches of tube length is nearly 12.5 in.sup.2 as opposed to the tube cross section area of slightly less than 0.3 in.sup.2. Obviously, the tubes hold a large percentage of available radioactive material. The 1/10 value layer (thickness of material to reduce the radiation by 1/10 for gammas) is about 1 inch of lead. If we could put 1 inch of lead under the tubesheet, we would therefore reduce the radiation contribution from the tubesheet area by a factor of 10 (reduction of from 30 to 3 rem). This reduction in exposure would increase the amount of time a worker could spend in the head which would result in significant savings in: 1. Exposure (due to the increased amount of time for each worker to remain in the plenum); 2. Manpower (less workers needed for a specific job); and 3. Job efficiency (more time allowed, so workers are not pressured into mistakes, due to hurrying). The invention, then, is a modular radiation shielding system which consists of lead tiles one inch thick being mounted rigidly under the tubesheet. This is accomplished using a rapid tube gripper which enters and locks itself in a tube with one single shove, therefore decreasing installation time. Each gripper can hold loads in excess of 200 lbs with no slippage, giving the workers below confidence in its integrity. Moreover, it cannot be bumped free or accidentally released but when purposely disengaged will exit the tube quickly and easily. The individual lead shielding tiles are approximately 71/2".times.83/4".times.1" with a transverse gripper hole and weigh approximately 30 lbs. This size and shape can be easily handled by one person and is light enough to be remotely installed. The tile size and weight may vary depending on the particular steam generator and the amount of shielding desired. Each tile is contained in a stainless steel sheath which preserves the tile's geometry and prohibits undesirable lead deposits in the head. The sheath also allows for easy decontamination of the tiles and is compatible with steam generator materials. The tiles are symmetrical about one axis and asymmetrical about a second axis perpendicular thereto, so that if a previously plugged tube is encountered, a quick 180.degree. rotation aligns the gripper hole with another tube. When the sheathed modular tile is pushed to the tubesheet, the gripper holds it in position. It's release is accomplished by pulling on the projecting handle to release the cams and, thus, the assembly. It is estimated that 6 to 10 assemblies could be installed in one minute by one jumper. This would result in an installation time of about 10 minutes for an entire 8000 tubes. Since nearly 1200 (1/7) of the tubes are in the outer 5 rows which do not significantly contribute to worker exposure due to their location, it is obvious that the radiation level would be greatly reduced before all the tubes are blocked. The installation exposure is, therefore, appreciably less than that which would be received in a field of 30 rem for 10 minutes (5 man rem). The individual sheathed tiles can also be independently removed to allow access to local tubes while still providing area shielding. |
description | This application claims priority to and the benefit of Korean Patent Application No. 10-2014-0083596 filed in the Korean Intellectual Property Office on Jul. 4, 2014, the entire contents of which are incorporated herein by reference. The present invention disclosed herein relates to a porous cooling block for cooling corium and corium cooling apparatus including the same and corium cooling method using the same, and more particularly, to a porous cooling block for cooling corium and corium cooling apparatus including the same and corium cooling method using the same capable of cooling corium with safety and rapidity with being applied to a variety of nuclear reactor facilities. Generally, a nuclear power plant (NPP) carries out a producing function of electric energy to be able to be used in practical living using nuclear energy by controlling a plenty of energy occurred by nuclear fission to be released slowly. Here, though the possibility of occurrence is very low, corium of ultra-high temperature with radioactive feature may be released into a cavity of primary reactor containment under a nuclear reactor container by nuclear fuel of reactor core being molten and breakdown of nuclear reactor container when a severe accident occurs at a nuclear power plant. At this situation, the corium released is a radioactive material with ultra-high temperature higher than 2000K, and has a feature that heat is occurred endlessly. And when fail to cooling the released corium properly, the primary reactor containment of nuclear reactor constructed as a concrete structure may be damaged by the corium of ultra-high temperature and radioactive material may leak outside. In case that radioactive material being leaked from the primary reactor containment of nuclear reactor like this is released to soil or atmosphere, it may not only become a threat to a stability of the nuclear power plant facilities, of course become a pollution to the surrounding environment of the nuclear power plant, but also may cause a critical harmful influence to the health of ordinary general public. Thus, recent nuclear power plant applies or develops a method for cooling and controlling properly the corium in the primary reactor containment of nuclear power plant in order not to leak the corium outside of the primary reactor containment of nuclear power plant. That is, a method of capable of cooling the corium with ease is required since the corium of high temperature released from a nuclear reactor container to a cavity pore of the primary reactor containment of nuclear power plant may melt and erode the floor concrete of the primary reactor containment of nuclear power plant if not cooled properly. At this time, the cooling of released corium is divided upper portion cooling and lower portion cooling of the corium. Since a damage of floor concrete occurs by contact of lower portion of corium, a lower portion cooling of the corium is required. For this lower portion cooling of the corium, an indirect cooling method in which cooling is performed at a state a cooling-water and the corium are not in contact using a cooling container, a direct cooling method in which cooling is performed at a state a cooling-water and the corium are directly in contact, etc. may be applied. Among this, the direct cooling method can get a more increased cooling efficiency than that of indirect cooling method since it cools directly the corium by contacting the corium and cooling-water, and has a merit of requiring less installation space. Thus, a development of a corium cooling apparatus is wanted with easy installation for application not only to a new nuclear power plant but also to a running nuclear power plant and capable of maximization of cooling efficiency. The present invention is to provide a porous cooling block for cooling corium and a corium cooling apparatus having the same and a corium cooling method using the same capable of increasing the safety of nuclear power plant by cooling the corium of high temperature released from the container of nuclear power plant with ease at an occurrence of significant accident of the nuclear power plant. The present invention is to provide a porous cooling block for cooling corium and a corium cooling apparatus having the same and a corium cooling method using the same capable of application not only to a new nuclear power plant but also to a running nuclear power plant where the installation space is limited since any extension for installation space is required and with easy installation of cooling apparatus. The present invention is to provide a porous cooling block for cooling corium and a corium cooling apparatus having the same and a corium cooling method using the same with increased easiness of production and construction by forming a natural circulation in a stacked structure of block type and with easy maintenance. A porous cooling block for cooling corium according to an embodiment of the present invention comprises a base part comprising one surface, an other surface facing the one surface, and a side surface connecting the one surface and the other surface each other, and including a plurality of pores, and a channel part formed open on at least one surface of the one surface, the other surface and the side surface and communicating with the plurality of pores. The channel part comprises a first channel formed to extend inward direction from one surface of either the one surface or the other surface of the base part, and a second channel passing through the side surface of the base part and communicating with the first channel. The second channel is provided in plurality in a way to pass through the side surfaces in one direction and another direction, and the plurality of the second channels may communicate with one another with being formed to cross in the base part. The first channel may comprise a first hole formed open on a surface of either the one surface and the other surface, and a first flow path connected to the first hole to form a path toward inside of the base part. The second channel may comprise a pair of second hole formed open to the side surface, and a second flow path connecting the second holes to form a path inside of the base part and communicating with the first channel. An average cross-sectional area of the first channel may be smaller than that of the second channel and larger than that of the plurality of pores. A corium cooling apparatus according to an embodiment of the present invention comprises a plurality of porous cooling blocks arranged to align with one direction and another direction crossing the one direction to form a plane, a sacrificial part received safely on the plurality of porous cooling block and covering an exposed upper surface of the porous cooling blocks, and a cooling-water supply unit for supplying cooling-water to the porous cooling blocks. Each of the plurality of porous cooling blocks may be arranged to be able to separate from an aligned and arranged region. The porous cooling block may include any one of above mentioned features. The sacrificial part may comprise a separation member covering the porous cooling block, and a sacrificial member received safely on the separation member, and at least one member of the separation member and the sacrificial member may be provided being divided in plurality and may be aligned and stacked in order on the plane. On a side surface of the porous cooling block, a side surface separation member may be arranged which is arranged on the most outside edge of a width formed by the plane. The separation member may provide a sealed space for sealing the porous cooling block. The cooling-water supply unit may comprise a cooling-water storage storing cooling-water supplied to the porous cooling block, and a cooling-water passing pipe whose one end is connected to the cooling-water storage and another end communicates with the porous cooling block. A corium cooling method according to an embodiment of the present invention comprises detecting an occurrence of disorder by release of corium, supplying cooling-water to a plurality of porous cooling block at the same time when the corium melts a sacrificial part, and cooling the corium by discharging the cooling-water from the porous cooling block. The sacrificial part may lower a thermal load per unit volume of the corium and may have the corium be distributed over an upper surface of the sacrificial part. The cooling the corium may comprise generating steam by heat transfer of the corium and the cooling-water, and cooling the corium to the shape of a porous form by jetting the steam or the cooling-water toward the corium. The porous cooling block may include any one of above mentioned features. According to a porous cooling block for cooling corium and a corium cooling apparatus having the same and a corium cooling method using the same of the present invention, corium of high temperature can be cooled with safety and ease, thereby corium of high temperature is prevented from being released outside of the primary reaction containment of nuclear power plant. That is, porous cooling block is provided in plurality which can endure the temperature delivered from the corium released from a container of nuclear power plant due to an occurrence of significant accident. And cooling-water flow path formed on each of the plurality of porous cooling blocks is fabricated to communicate with each other. Thus, the cooling-water can be supplied to the one surface forming the cooling block uniformly on the whole, which can cool the corium with ease to solidify it in the shape of a porous form. And, the porous cooling block is fabricated by the unit of a plurality of brick structure and arranged to align with a position at which the corium may be released. Thus, it can be fabricated in a factory and a block of uniform quality may be obtained. And it may be assembled at a field which enhances the easiness of construction. Thus, fabrication and installation is easy to be applied and used to a running nuclear power plant facility with ease. And when maintenance is needed, a cooling block of needed region can be replaced which facilitates easy maintenance and short required time, which leads to a reduction in maintenance cost. This short installation and maintenance time due to easiness of construction and maintenance may minimize the exposure of a worker to radioactive ray and may reduce the cost of operator of nuclear power plant. And, a small vertical channel is formed on the one surface in the porous cooling block facing the corium. Thus, the steam jetted through the channel pushes the corium and penetrate into the corium to cool it rapidly and solidify it in the shape of a porous form. Here, the small vertical channel can secure an enough cooling flow path even when any floating matter like dregs capable of occurring with the corium being cooled and solidified may stop up a pore formed at the cooling block. And, a supply unit of cooling-water is fabricated to be connected to a porous cooling block directly when installed to a new nuclear power plant facility. Thus, the porous cooling block can be installed as a complete sealing structure in a accommodation space. Thus, an installation as a high level passive facility may be possible which does not require any active facilities such as outer power supply, action of operator, valve and pump etc. even at the time of severe accident. Hereinafter, a desirable embodiment of a porous cooling block for cooling corium and a corium cooling apparatus having the same and a corium cooling method using the same of the present invention may be described in reference to accompanying drawings. A melt cooling apparatus according to an embodiment of the present invention is an apparatus of blocking the discharge of the melt outward from the accommodating space by cooling the melt within the accommodating space with ease when a melt of high temperature is released from a container arranged in facilities forming an accommodating space of a predetermined size. It includes a cooling block of porous and can reduce the temperature of the melt by supplying a cooling medium to the melt contacting to the cooling block with ease. Thus, the present invention may be used to cool the corium with equipped at a position where the corium is released from the nuclear reactor container in a nuclear reactor facility having a primary reactor containment of nuclear reactor forming a space arranged for nuclear reactor container. However, the usage of the porous cooling block and the melt cooling apparatus having the same may not be limited this, and it may be used to a variety of facilities for suppressing or preventing a thermal damage of the facilities by a melt of high temperature. That is, the melt in the present invention may be corium, and the container accommodating the melt may be a nuclear reactor container. And, the thing forming an accommodating space of a predetermined size may be a primary reactor containment of nuclear reactor. Thus, the terminologies described above may be used in a mixed sense which means the same meaning. Hereinafter, a porous cooling block and a corium cooling apparatus having the same according to an embodiment of the present invention will be described in reference to FIG. 1 to FIG. 6. FIG. 1 is a schematic diagram showing a nuclear reactor facility equipped with a corium cooling apparatus according to an embodiment of the present invention. FIG. 2 is a perspective view illustrated selectively of a corium cooling apparatus according to an embodiment of the present invention. FIG. 3 is a drawing for description of a porous cooling block according to an embodiment of the present invention. Here, FIG. 3(a) is a perspective view showing the porous cooling block and inner waterway. FIG. 3(b) to (d) are multi-side view of the porous cooling block. FIG. 4 is a drawing for description of a porous cooling block according to a varied embodiment of the present invention. FIG. 5 is a drawing for description of charging-water status of the porous cooling block and a circulation status of cooling-water illustrated in FIG. 3. FIG. 6 is a drawing for description of installation status of the corium cooling apparatus according to a varied embodiment of the present invention. Referring to FIG. 1 and FIG. 2, a melt cooling apparatus 1000 according to the present invention may be installed at a lower portion cavity under a nuclear reactor container 10, for cooling the corium M released through a damaged part of the nuclear reactor container and preventing a mutual reaction of corium M and primary reactor containment of nuclear reactor at an occurrence of severe accident at a nuclear power plant. That is, at least a portion of the configuration of the melt cooling apparatus 1000 may be arranged at a structure partition of accommodating space R isolated from lower part of nuclear reactor container on the accommodating space R forming the primary reactor containment. At this time, the melt cooling apparatus 1000 forms a plane of predetermined size by being arranged to align to one direction, X axis direction and another direction, Y axis direction crossing the one direction, X axis direction at an isolated position from the lower part of the nuclear reactor container 10. It comprises a plurality of porous cooling block 100 including a channel part 130, 150 communicating with at least d portion of a plurality of pore P, a sacrificial part 300 received safely on the plurality of porous cooling block 100 and covering an exposed upper surface of the porous cooling blocks 100, and a cooling-water supply unit 500 for supplying cooling-water W to the porous cooling block 100. The accommodating space R provides a space in which the nuclear reactor container 10 and provides a region to which a cooling-water W is supplied. For example, the accommodating space R is made of a plurality of partitions arranged to be isolated to outside of the nuclear reactor container 10 and may be a primary reactor containment 20. This is a space formed to prevent a radioactive material exceeding a permitted amount from being discharged over a managed region to an environment such as soil or atmosphere at the time of accident of nuclear reactor. The nuclear reactor container 10 and some elements of the melt cooling apparatus 1000 may be arranged inside. Such an accommodating space R may be divided into a cooling part Rb at which some configurations of cooling apparatus 1000 for cool the corium M released from a damaged part of the nuclear reactor container 10 and dropped downward, and a storage Ra accommodating a predetermined amount of cooling-water W for injection of cooling-water W to the porous cooling block 100 described later from outside region of the cooling part Rb. However, the accommodating space R is not limited to the separation region, and may provide a sealing part additionally according to a varied embodiment described later. Referring to FIG. 3, the porous cooling block 100 is arranged at a position isolated from a lower part of the nuclear reactor container 10. More particularly, it is arranged to align on a floor structure facing the lower part of the nuclear reactor container 10 among partitions configuring the accommodating space R. The porous cooling block 100 is provided to cool the corium M by supplying the cooling-water to the corium M released from the nuclear reactor container 10 at the time of occurrence of severe accident at a nuclear reactor facility. Such a porous cooling block 100 comprises a base part 110 comprising one surface 112, an other surface 114 facing the one surface 112, and a side surface 116 connecting the one surface 112 and the other surface 114 each other, and including a plurality of pores P, and a channel part 130, 150 formed open to at least any one surface of the one surface 112, the other surface 114 and the side surface 116 and communicating with at least some of the plurality of pores P. Here, as shown in FIG. 2, porous cooling block 100 is a block structure of predetermined size. And by a plurality of it being arranged to align to one direction, X axis direction and another direction, Y axis direction, it may form a plane having a predetermined size, that is area, in the one direction, X axis direction and the other direction, Y axis direction and it may be arranged to be able to be separated on the floor structure facing the lower part of nuclear reactor container 10. The base part 110 is a basic structure to form a body of the porous cooling block 100, and comprises one surface 112, an other surface 114 facing the one surface 112, and a side surface 116 connecting the one surface 112 and the other surface 114 each other and a plurality of pores P. That is, the base part 110 is fabricated as a brick of a predetermined size, and the plurality of pores P are formed thereon to deliver the cooling-water with ease inside by communicating various path each other. As described above, the base part 110 may be made of porous concrete or a ceramic material of high-temperature. Here, in case the base part 110 is fabricated with the former porous concrete, a base part 110 of block shape can be formed by pouring concrete into a mold of predetermined size and solidifying it. Thus, the base part 110 can be formed in a simple way. However, high-temperature strength of concrete is weak, and damage may occur at concrete by release weight of the corium M or falling weight of damaged part of nuclear reactor container 20. And a problem of being molten by the corium M during cooling process of the corium M may occur. On the other hand, in case the base part 110 is fabricated with a latter ceramic material of high-temperature, a material having excellent high-temperature strength such as high-purity aluminum oxide (Al2O3) of melting point of 2072° C., silicon carbide (SiC) of melting point of 2730° C., silicon nitride (Si3N4) of melting point of 1900° C. and zirconium oxide (ZrO2) of melting point of 2715° C. may be used as a ceramic material of high-temperature. During fabrication of the base part 110, by applying compressing pressure to the base part 110 in a strength to form pore P, the base part 110 can be fabricated to have the plurality of pores P. Here, in case of too low pressure, the base part 110 cannot show enough mechanical strength. Thus, the compressing power for fabrication of the base part 110 should be a value to endure the corium M of high-temperature and to have strength not to be broken when contacting each other at the time of arrangement to align the plurality of base parts 110. By fabrication of the base part 110 using ceramic material of high-temperature like this, more increased thermal stability and structural strength can be realized than by the case of the base part 110 using porous concrete. As described above, the base part 110 may be fabricated with selection from the above-mentioned materials, and more particularly, may be applied and used according to temperature and design weight of the corium M. The channel part 130, 150 is formed open to any one surface of the one surface 112, the other surface 114 and the side surface 116 forming the base part 110, and it communicates with at least a portion of a plurality of pores P formed on the base part 110. Thus, the channel part 130, 150 is provided to discharge the cooking-water W from the base part 110 with ease. That is, the channel part 130, 150 communicates with some of pores P of the base part 110, and is provided to discharge the cooling-water W with ease than in the case of the cooling-water being discharged through pore P from the base part 110. Thus, the channel part 130, 150 comprises a first channel 130 formed to extend inward from any one surface of the one surface 112 and the other surface 114 of the base part 110, and a second channel 150 passing through the side surface of the base part 110 and communicating with the first channel 130. The first channel 130 is provided to discharge the cooling-water W to an upper surface of the base part 110, and comprises a first hole 132 formed open to any one surface of the one surface 112 and the other surface 114 of the base part 110 facing the nuclear reactor container 10, and a first flow path 134 forming a path inward direction of the base part 110 with being connected to the first hole 132. That is, the first channel 130 is formed to discharge the cooling-water W to upward with ease than in the case discharging the cooling-water W through pore P. And as shown in FIG. 3, it can be formed to have average diameter of h larger than the average diameter p of one pore P in case of forming the path in circular shape, and it can be formed to have average cross-sectional area larger than that for the plurality of pores P in case of forming the path in non-circular shape. Here, the average cross-sectional area means an average cross-sectional area for one pore P, and it means that an average cross-sectional area of each of the plurality of pores P have smaller average cross-sectional area to the average cross-sectional area of the first channel 130. And, the first channel 130 can perform a function for minimizing the flow path loss of the cooling-water W, even in case some pore P are stopped up by dregs like a floating matter which can be occurred at the time of melting sacrificial member 330 described later and cooling the corium M. The direction of extension formation of the first channel 130 like this is inward direction of the base part 110 to form to extend to the depth communicating with the second channel 150 flowing in the cooling-water W. The shape is shown as circular holes from which the flow path is formed, but the shape of the first channel 130 is not limited to this and the size can be formed variously. The second channel 150 is formed to passing through the side surface of the base part 110 to flow in the cooling-water W into the porous cooling block 100, and may comprise a pair of second holes 152a and 152b, or collectively 152, communicating with at least some pores P and the first channel 130 formed on the base part 110, being provided to deliver the cooling-water W to the pore P and the first channel 130 and formed open to the side surface, and a pair 154 of second flow paths 154a and 154b, or collectively 154, forming path inside of the base part 110 by connecting the second hole and communicating with the first flow path 134. Here, in an embodiment of the present invention, to facilitate an easy supply of the cooling-water W to the plurality of porous cooling blocks 100 arranged to align in one direction and in the other direction, a pair of second channel is provided on the base part 110. The second channel 150 may comprise a pair of second channels 150a and 150b, which can cross each other in the base part 110 to form to communicate each other. That is, while one of the second channel 150 is formed to pass through two side surfaces facing each other among the side surfaces 116 of the base part 110, the other of the second channels 150a and 150b may be formed to pass through two side surfaces facing each other among the rest of the side surfaces 116 of the base part 110 and communicate each other. In case a pair of the second channels 150a and 150b are provided and formed per the base part 110 like this, the second channels 150a and 150b formed respectively on the porous cooling block 100 arranged to align to the one direction and the other direction, may communicate each other in the one direction and the other direction, by which the cooling-water W can move along the second channel 150 uniformly with ease to enable a uniform supply of the cooling-water W. Here, in case the second channel 150 is formed in a circular form to supply the cooling-water W to the plurality of base part 110, the average diameter H can be formed as larger size than the average diameter h of the first channel 130 described above, and in case in non-circular form, it can be formed to have an average cross-sectional area increased than that of the first channel 130. As described above, referring to FIG. 3(d) of plane view shown from a lower direction showing A-A′ cross-section of the first channel 130 and the second channel 150, a moving path of the cooling-water W is formed to communicate with each other in the base part 110, thus the cooling-water W can move with ease in the base part 110 to the side surface side and the upper surface side. Moreover, in case the plurality of porous cooling blocks 100 are arranged to align each other and the second channels 150 communicate each other formed on each of the porous cooling block 100, the cooling-water W can be delivered with ease via the flow path to enable an even supply of the cooling-water W. On the other hand, porous cooling block 100 may be fabricated to be varied like the one shown in FIG. 4. In a porous cooling block 100 according to a varied embodiment of the present invention, the second channel 150′ formed on the base part 110 is formed open to any one surface of the side surface facing each other, and the one surface 112 and the other surface 114 on which the first channel 130 is formed. Referring to FIG. 4, the second channel 150′ according to a varied embodiment is formed open to the other surface 114 of the base part 110 and at least a portion of the side surface 116 and the other surface 114 of the base part 110 may be formed open as shown hatched region of B in FIG. 4(b). In case the second channel 150′ is formed as such, supplying the cooling-water W from a lower part is easy. And since the cooling-water W can be supplied through an increased flow path area than a case in which the cooling-water W is supplied only through the side surface, occurrence of a problem by a flowage resistance can be suppressed even when plenty of cooling-water W is flowed in. A detailed description according to a porous cooling block 100′ of a varied embodiment as such can be described in relation to FIG. 6. The porous cooling block 100, 100′ of the present invention as such has an advantage that it can be installed to facilities with ease by making plane by being aligned to the one direction and the other direction in plurality. This has a merit that the structure of facilities and the formation method is simple and easy compared to a conventional one in which a concrete composite is constructed and cured to a predetermined height on a installation space, and then the cooling-water is supplied to the corium by flowing the cooling-water to the concrete composite to produce a structure for supplying the cooling-water. The sacrificial part 300 is received safely on the plane configured by the porous cooling blocks 100, and may be provided to increase the required time of getting contact of the porous cooling block 100 and the corium M. More specifically, the sacrificial part 300 can react with the corium M firstly to secure a time to charge the cooling-water W to the porous cooling block 100 by the cooling-water supply unit 500 described later in case corium M is detected to be released from nuclear reactor container 10 by an occurrence of severe accident of nuclear reactor facilities. It comprises a separation member 310 received safely on the top part of the porous cooling block 100, and a sacrificial member 330 received safely on the top part of the separation member 310. The separation member 310 is arranged on the porous cooling block 100 and is for separating the sacrificial member 330 and the porous cooling block 100. It can be provided as single or divided plurality, and can cover an exposed upper surface of the porous cooling block 100 by being aligned on the plane which the porous cooling block 100 forms. The separation member 310 may be constructed as a metal plate, and is provided to suppress or prevent the sacrificial member 330 of not solidified from entering into the pore P or the first channel 130 of the porous cooling block 100 to block up at the time of field construction of the sacrificial member 330 described later. Here, the separation member 310 has an advantage of simple installation since there is no need of sealing of the porous cooling block 100 with welding or any fixing member not shown. That is, the separation member 310 can cover the porous cooling block 100 in single configuration or divided configuration with being received safely on the porous cooling block 100 without any fixation. In case the separation member 310 has a divided configuration into plurality, separate maintenance of separation member 310 of desired region is possible at the time of maintenance, and in case it is installed in a running nuclear reactor facility, it has an advantage of easy installation compared to a case it is provided as single. On the other hand, a side surface separation member 320 may be provided on the side surface of the porous cooling block 100 in a direction crossing the extension direction of the separation member 310. The side surface separation member 320 covers at least a partial region of the side surface of the porous cooling block 100 and may be formed to extend to a higher position than the porous cooling block 100. More specifically, the side surface separation member 320 may be provided to cover at least a part of opened one side of the porous cooling block 100 and to be formed to extend in a predetermined length toward the upper direction. The side surface separation member 320 may perform a function of pushing the porous cooling blocks 100 to reduce a separation distance among the plurality of porous cooling blocks, when the porous cooling block 100 is arranged on an opened side surface and is received safely on the floor part of the cooling part Rb. That is, it is arranged to contact to the most outer edge of the plane formed by the porous cooling block 100, and may perform a role of restricting the space on the partition occupied by the porous cooling block 100. The sacrificial member 330 may be provided to be received safely on the separation member 310 and to be separated from the porous cooling block 100. It is provided to reduce the temperature firstly and to facilitate a feature of spread to the corium M on the plane when corium M is released from the upper part on occurrence of severe accident. That is, it reduces a thermal output power per unit volume of the corium M by the reaction with the corium M and can reduce the thermal load per unit volume of the corium M. Thus, it facilitates cooling by the cooling-water W discharged from the porous cooling block 100. And it reduces the viscosity of the corium by the reaction with the corium M and can enhance the spreading of the corium M. Thus, it can suppress or prevent a local thermal load from rising on the porous cooling block 100. And, the sacrificial member 330 can perform a role to secure the time needed to charge the cooling-water W in the porous cooling block 100 during reaction with the corium M. The sacrificial member 330 may be provided on the separation member 310 with configured as single like the separation member 310, or may be provided in the same or similar number to the separation member 310 and arranged to be stacked on the top part of the separation member 310. On the other hand, a side surface sacrificial member 340 may be provided on an inner surface of the side surface separation member 320 in a direction crossing the extending direction of the sacrificial member 330 and on a partition of primary reactor containment of nuclear reactor. The side surface sacrificial member 340 is provided at inner side of the side surface separation member 320. It may be provided to protect a part of region of side surface separation member 320 even though it does not react with the corium M. That is, a side surface sacrificial member 340 performs a role of pressing a pressure to the side surface of the porous cooling block 100. Separation among the plurality of the porous cooling blocks 100 may occur in case melting starts by high temperature before supply of the cooling-water W through the porous cooling block 100. It can suppress or prevent a problem from occurring that the cooling-water W do not move uniformly by the occurrence of separation at communication paths among the second channels 150 of each of the porous cooling blocks 100. The sacrificial member 330 and the side surface sacrificial member 340 as such may be fabricated using a sacrificial concrete composite, and the composite is not limited but should be able to reduce the thermal output power and viscosity per unit volume of the corium M by reacting with the corium M. And it may be formed of a composite which can prevent recriticality of the corium M and can reduce the production amount or occurrence rate of hydrogen by reaction with the corium M. The cooling-water supply unit 500 is a means for supplying the cooling-water W to the porous cooling block 100 with being connected to an accommodating space R. The top part of the porous cooling block 100 is exposed after melting the sacrificial part 300 with reaction of corium M of high temperature and the sacrificial part 300. That is, the corium M and the porous cooling block 100 contacts each other. At this time, the cooling-water W is supplied to prevent the corium M from being released to outside of the primary reactor containment 20 forming the accommodating space R by lowering the temperature of the corium M. The cooling-water supply unit 500 as such comprises a cooling-water storage 510 accommodating the cooling-water, and a cooling-water passing pipe 530 whose one end is connected to the cooling-water storage and another end communicates with the porous cooling block 100. The cooling-water storage 510 is provided to accommodate and supply the cooling-water W to the porous cooling block 100. An apparatus capable of supplying the cooling-water W to the porous cooling block 100 continuously and repeatedly by a predetermined amount may be used. Here, in an embodiment of the present invention, the cooling-water W may be supplied to the storage Ra communicating with the second channel 150, to supply the cooling-water W through the second channel 150 of the most outer opened surface of the porous cooling block 100 in the accommodating space R. Here, the cooling-water W supplied from the cooling-water storage 510 is supplied in a state set at a predetermined low temperature and can cool the corium M of high temperature in a short time. And the cooling-water W supplied to the porous cooling block 100 and used to cool the corium M is accommodated again to the cooling-water storage 510 and is lowered to a predetermined temperature. And, the circulation rate of the cooling-water W may be increased by resupply the cooling-water W. The cooling-water passing pipe 530 forms a moving path of the cooling-water W to supply the cooling-water W to the porous cooling block 100. Here in an embodiment, a path communicating with the accommodating space R and supplying a cooling medium to storage Ra is formed. Thus, the cooling-water passing pipe 530 is arranged for a predetermined region to be inserted from the cooling-water storage 510 provided at outside of the primary reactor containment 20 into the primary reactor containment 20, by which the cooling-water W can be supplied from the cooling-water storage 510 into the inside of the accommodating space R. Here, a sealing member for sealing not shown is provided between the cooling-water passing pipe 530 and the primary reactor containment 20. The cooling-water passing pipe 530 and the primary reactor containment 20 may be formed not to have a separation space therebetween. Referring to FIG. 5, a brief description on the water charge status of the porous cooling block 100 using the cooling-water supply unit 500 will be given. The cooling-water W discharged from the cooling-water storage 510 to the accommodating space R through the cooling-water passing pipe 530 is charged in the storage Ra which is a space to communicate with the side surface of the porous cooling block 100 within the accommodating space R. That is, the cooling-water W is charged up to a height for the storage Ra and the porous cooling block 100 to communicating with each other, for the cooling-water W to flow in to the porous cooling block 100 arranged at a relatively high position compared to the storage Ra. And when the cooling-water W is charged up to on a similar or equal line or to higher position to the side surface of the porous cooling block 100, the cooling-water W is supplied in the porous cooling blocks 100 forming a plane through the second channels 150 communicating with each other of the plurality of the porous cooling block 100. The cooling-water W supplied as such is charged fully to the first channel 130, the second channel 150 and the plurality of pores P, and then is discharged to the top part of the porous cooling block 100 when the upper surface of the porous cooling block 100 is open. And, the discharged cooling-water W can be flowed in again to the storage Ra according to the discharged amount and recirculated and reused. On the other hand, a method in which the cooling-water supply unit 500 supplies the cooling-water W to the accommodating space R and then the cooling-water W is flowed in the porous cooling block 100, that is, a method in which the cooling-water W being charged up to a predetermined amount in the storage Ra is diffused to be supplied to the porous cooling block 100 is described above. However, the method by which the cooling-water supply unit 500 is provided is not limited to the above method. It may be provided as a melt cooling apparatus 1000 of a varied embodiment described later. Thus, the cooling-water supply unit 500 may be provided at a region not disturbing a major configuration of the nuclear reactor facilities at a running nuclear reactor facility and newly produced nuclear reactor facilities, and may be formed in a variety of structure capable of supplying the cooling-water W to the porous cooling block 100. Hereinafter, an installation status of a corium cooling apparatus 1000′ according to a varied embodiment will be described with reference to FIG. 6. Here, FIG. 6 is a drawing for description on an installation status of a corium cooling apparatus according to a varied embodiment of the present invention. Referring to FIG. 6, the melt cooling apparatus 1000′ according to a varied embodiment of the present invention performs identical or similar role to the structure proposed in the melt cooling apparatus 1000 according to the above-mentioned embodiment, except that the sacrificial part 300′ is positioned so as for the porous cooling block 100′ to be provided to seal in the accommodating space R, and thus, the arranged position of the cooling-water supply unit 500 became different. Thus, the description on the porous cooling block 100 will be omitted hereinafter, but configuration and position of the sacrificial part 300′ and the cooling-water supply unit 500 will be described. The sacrificial part 300′ comprises a separation member 310′ forming a sealed space for covering and sealing the opened upper surface of the plurality of the porous cooling block 100 arranged to align, and a sacrificial member 330 received safely on the separation member 310′. Here, the separation member 310′ seal and cover the opened upper surface of the porous cooling block 100′ without open space aligned and arranged on the partition facing the nuclear reactor container 10, and suppress or prevent the porous cooling block 100′ from communicating with outside. And, in a varied embodiment, by including a side surface separation member 320′ not contacting to the partition of the primary reactor containment of the porous cooling block 100′, and arranged to contact to an opened one side surface to be connected to the separation member 310′, the porous cooling block 100′ can be provided in a sealed status. Here, the cooling-water supply unit 500 to supply the cooling-water to the porous cooling block 100′ supplies the cooling-water to the sealed space in which the porous cooling block 100′ is sealed. For this, the cooling-water passing pipe 530 may be provided to be buried in the partition structure in which the porous cooling block is arranged to align. That is, the cooling-water W can be supplied to the lower part of the porous cooling block 100′ through the partition structure from the lower part of the partition structure for a part to be buried in the partition structure. The cooling-water supply at a sealed space of the porous cooling block 100′ as described above can be applied to a case wherein the burial of the cooling-water passing pipe 530 is easy as in new nuclear power plant. By supplying the cooling-water directly to the sealed space with arrangement of the porous cooling block 100′ as a sealed structure as such, the cooling-water W may be charged to the plurality of the porous cooling blocks 100′ fully always, or the water charging time required to charge may be minimized. That is, the cooling-water W is accommodated in a predetermined space, and then the cooling-water W is charged to the sealed porous cooling block 100. And, the sacrificial member 330′ is eroded by the corium M on occurrence of severe accident of nuclear reactor facilities. Thus, when the separation member 310′ is applied and opened, the cooling-water W is released naturally to the upper part of the porous cooling block 100′ to achieve passive facilities. Here, during a normal operation of nuclear reactor facilities, the cooling-water W is charged in the porous cooling block 100′ for a long time, which leads to decline in the quality of the cooling-water W, occurrence of impurity, erosion of the separation member 310′ and the cooling-water passing pipe 530, and contamination of pores of the porous cooling block 100′. To prevent or suppress them, during normal operation of nuclear reactor facilities, operation may be done in the state of injection of corrosion inhibitor or a predetermined gas in the porous cooling block 100′ and some of cooling-water passing pipe 530. On the other hand, to the corium cooling apparatus 1000, a temperature detector may be provided which can detect the release of the corium M from the accommodating space R. The temperature detector not shown is provided at least any one place of accommodating space R where the corium M is released and measures the temperature of the accommodating space R to be able to detect the occurrence of release of the corium M. More specifically, it is arranged at a position close to a point at which the corium M is released from the nuclear reactor container 10, or to a point at which the sacrificial part 300 and the porous cooling block 100 are arranged, and is able to measure the temperature of the accommodating space R. Here, the temperature detector is non-contact type which can measure the temperature at a separated place from the point where the corium is released, and a pyrometer may be used for it capable of measuring the temperature by detecting thermal energy. And, a controller not shown to control the operation of the cooling-water supply unit 500 according to the result of temperature measurement of the temperature detector with being connected to the temperature detector may be provided to the corium cooling apparatus 1000. The controller controls the operation of the cooling-water supply unit 500 according to the temperature inside the accommodating space measured by the temperature detector. More specifically, in case the temperature of the accommodating space R measured by the temperature detector is a value that is rapidly raised compared to the temperature of normal status, or the temperature when the corium is not released from the nuclear reactor container, the controller determines as there occurred a severe accident with release of corium M, and then have the cooling-water supply unit 500 operate to supply the cooling-water W to the porous cooling block 100. Here, the controller is a device capable of determining the occurrence of accident by being delivered with the temperature value and capable of delivering an operation signal to the cooling-water supply unit 500, and a device such as a PLC panel and PC can be used. However, the controller is not limited such, and various transferring device of measures of operator and various signal may be used for the same. And, in case the temperature detector which is provided as in a varied embodiment 1000′ of FIG. 6 and capable of detecting the occurrence of severe accident is not necessary, or active facilities having outer power supply or operator's action for the cooling-water supply unit 500 is not necessary, it may be used as passive facilities. A melt cooling method using the melt cooling apparatus 1000 fabricated and configured as mentioned above will be described in reference to FIG. 7 and FIG. 8. FIG. 7 is a flow chart showing in order the method of cooling the melt using the porous cooling block and the melt cooling apparatus according to an embodiment of the present invention. FIG. 8 is a process diagram showing a method of cooling melt of FIG. 7. Hereinafter, the corium cooling method will be described on the basis of porous cooling block and the melt cooling apparatus having the same of above mentioned embodiment, where the cooling method thereof will be applied to the varied embodiment the same. Referring to FIG. 7 and FIG. 8, the melt cooling method according to the embodiment of the present invention comprises detecting an occurrence of severe accident by release of corium M, supplying cooling-water W to a plurality of porous cooling blocks 100 comprising a plurality of pores P, a first channel 130 and a second channel 150 each formed in upward direction and planar direction, at the same time when the corium M melts a sacrificial part 300, and cooling the corium M by discharging the cooling-water W through pore P and the first channel 130. Even though not high possibility of occurrence, there is a possibility that corium M is released to lower part of nuclear reactor container 10 through a damaged part of nuclear reactor container 10 in case of occurrence of severe accident at nuclear power plant. The corium M is a melt material of high temperature mixed with enriched uranium of nuclear fuel of core of nuclear reactor installed inside of nuclear reactor container 10, zirconium used as clad material and many materials inside the nuclear reactor container 10. Cooling of the corium M is required since the corium M generates heat by decay of nuclear fission product inside. Thus, the released corium M reacts with floor concrete configuring the primary reactor containment 20 to generate plenty of non-condensable gas to melt and erode the floor. Therefore, as shown in FIG. 8(a), it is detected that the corium M is released from the nuclear reactor container 10 and dropped into a inner space of the primary reactor containment 20 (S1). That is, a rapid increase of temperature in the accommodating space R is known by a temperature detector measuring the temperature of the accommodating space R in the primary reactor containment 20, and the occurrence of severe accident of release of the corium M from the nuclear reactor container 10 is verified. When a release of the corium M is detected, the cooling-water is supplied to charge the cooling-water in the porous cooling block for cooling of the corium M (S2). That is, as shown in FIG. 8(b), to supply the cooling-water W to the second channel 150 passing through the side surface of the porous cooling block 100, the cooling-water is supplied to the storage Ra in the accommodating space R, the cooling-water W charged to a predetermined height in the storage Ra flows in the porous cooling block 100, and the cooling-water W is uniformly supplied to the plurality of blocks through the second channels 150 formed on each of the porous cooling block 100. On the other hand, during supply of cooling-water W, the process of melting and erosion of the corium M and the sacrificial part 300 arranged on the porous cooling block 100 is performed at the same time (S3). That is, at least before opening the porous cooling block 100 to the corium M by melting and erosion of the sacrificial part 300, charge of cooling-water W to upper part of the porous cooling block 100 or to an upper water level can be completed, or the charge completion of cooling-water W in the porous cooling block 100 and the melting and erosion of the sacrificial part 300 can be performed at the same time. In the reaction process of the sacrificial part 300 and the corium M, firstly the sacrificial member 330 and the corium M which are arranged the most upper surface react each other and the sacrificial member 330 is molten. And then the separation member 310 which is arranged under part of the sacrificial member 330 and the corium M react each other and the separation member 310 is molten. Here, by the melting reaction of the sacrificial member 330 and the corium M, the thermal output power per unit volume of the corium M is reduced, and the viscosity of the corium M is reduced. As a result, the load in the process of cooling the corium M by the cooling-water W can be reduced. And with reduced viscosity, the corium M can be distributed on the plane formed by the sacrificial part 300 and the porous cooling block 100, by which the occurrence of local temperature increasing region by agglomerate of the corium M may be suppressed or prevented. After the cooling-water W is charged (S2), the sacrificial part 300 is molten (S3). Then the upper surface of the porous cooling block 100 covered by the sacrificial part 300 is exposed (S4). That is, the corium M moves gradually toward the porous cooling block 100 as the sacrificial part 300 melts, and as shown in FIG. 8(c), the cooling-water W can contact directly to the corium M (S5) through the plurality of pores P and the first channel 130 which are open to the upper surface of the opened porous cooling block 100 to cool the corium M (S6). Here, the cooling method of the corium M will be described in detail. The cooling-water W charged in the plurality of pore P and the first channel 130 which are vacancy in the porous cooling block 100 are boiled by heating of the corium M. That is, the cooling-water W is boiled and the steam pressure becomes larger than the pressure of circumstances. And the boiling occurs not only at the surface of the liquid but also inside to generate the steam rapidly and cools the corium M. After that, the melting and erosion of the sacrificial part 300 by continuous heating of the corium M is repeated, and when the cooling block 100 contacts with the corium M, the steam or cooling-water W in the cooling block 100 penetrates and jets into the corium M to cool the corium M. The cooled corium M is solidified in a shape of porous form, and then long-term cooling can be performed by natural circulation of the cooling-water W. In the present invention, the cooling-water W is boiled and directly contact to the lower part of the corium M to preform cooling. Moreover, in the long-term cooling when the corium M is solidified in a shape of porous form, the cooling-water W or steam is continuously discharged to the upper part through the pores P and the first channel 130, by which the corium M is enclosed and contact and cooling over the whole region, or upper part and lower part of the corium M can be performed. Therefore, the surface area of contact between the cooling-water W and the corium M can be increased and the cooling of the corium M can be performed with more ease. And a part of the cooling-water W discharged from the porous cooling block 100 is moved to the storage Ra again, and then the cooling-water W supplied once is recycled to be used again to cool the corium M. As described above, the porous cooling block and the corium cooling apparatus having the same according to an embodiment of the present invention use a ceramic material of high temperature with increased resistivity on the melting by the temperature of the corium. By forming the porous block using the same and forming the plane with aligned state each other, it can be applied to various facilities including a nuclear reactor facility without limitation of facilities. Here, a channel open to a space between the side surface and upper surface of the porous cooling block is formed, which facilitates easy delivery of the cooling-water among the plurality of the porous cooling block and the cooling-water can be supplied uniformly on the plane of the cooling block contacting the corium. And the first channel formed open to the upper surface maintains the flow of the cooling-water, and even in the case some residue of the corium and the sacrificial member penetrate into the pore during recirculation of the cooling-water, it relaxes or prevents the reduction of cooling performance owing to the reduction of cooling flow path. Although the present invention has been illustrated and described as to a desirable embodiment, the present invention is not limited by the above embodiment, and those ordinary skilled in the art may understand that variety of variation and equivalent other embodiment are possible within the scope of the present invention claimed. Therefore, the range of technical protection of the present invention should be determined by the claim attached. |
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052873927 | claims | 1. In a nuclear reactor comprising a core of nuclear fuel elements, a downcomer annulus between a pressure vessel and said core, means for feeding feedwater into said pressure vessel, means for forcing said feedwater to flow through said core to cool said fuel elements, and means for separating steam and water phase in the mixture exiting said core, said water phase flowing from said steam/water separating means to said downcomer annulus, the improvement wherein catalytic means are arranged downstream of said steam/water separating means, said catalytic means comprising catalytic material arranged in an open structure which allows water phase to flow therethrough, said catalytic means being disposed so that substantially all of the water phase exiting said steam/water separating means flows therethrough, and said catalytic material having a surface area-to-volume ratio sufficiently high enough to ensure that substantially all of the water phase passing through said catalytic means flows close enough to a surface of said catalytic material to enable said catalytic material to catalyze the decomposition of hydrogen peroxide molecules dissolved in said water phase exiting said steam/water separating means into water and oxygen molecules. 2. The nuclear reactor as defined in claim 1, wherein said catalytic material comprises a water recombination catalyst which catalyzes both the decomposition of hydrogen peroxide molecules and the recombination into water of hydrogen and oxygen molecules dissolved in said water phase exiting said steam/water separating means. 3. The nuclear reactor as defined in claim 1, wherein said catalytic means comprise a generally annular containment means in which said catalytic material is packed, said containment means having openings which enable water phase to flow through said containment means, but which are not of a size to allow escape of said catalytic material therefrom. 4. The nuclear reactor as defined in claim 3, further comprising a core shroud and a shroud head, wherein said catalytic means is mounted on said shroud head and circumferentially encompasses said steam/water separating means. 5. The nuclear reactor as defined in claim 3, further comprising a core shroud and a shroud head, wherein said catalytic means is mounted on said core shroud and extends across said downcomer annulus. 6. The nuclear reactor as defined in claim 1, wherein said catalytic material comprises entangled wires or strips or crimped ribbons made from metal. 7. The nuclear reactor as defined in claim 6, wherein said metal is stainless steel plated or alloyed with a noble metal. 8. A nuclear reactor comprising a core of nuclear fuel elements, a downcomer annulus between a pressure vessel and said core, means for feeding feedwater into said pressure vessel, means for forcing said feedwater to flow through said core to cool said fuel elements, means for separating steam and water phase in the mixture exiting said core, said water phase flowing from said steam/water separating means to said downcomer annulus, and means for catalyzing the recombination into water of hydrogen and oxygen molecules dissolved in said water phase exiting the steam/water separating means, wherein said means for catalyzing water recombination are arranged downstream of said steam/water separating means and comprise catalytic material arranged in an open structure which allows water phase to flow therethrough, said means for catalyzing water recombination being disposed so that substantially all of the water phase exiting the steam/water separating means flows therethrough. 9. The nuclear reactor as defined in claim 8, wherein said means for catalyzing water recombination comprise containment means in which said catalytic material is packed, said containment means having openings which enable water phase to flow through said containment means, but which are not of a size to allow escape of said catalytic material therefrom. 10. The nuclear reactor as defined in claim 8, wherein said means for catalyzing water recombination has a generally annular structure. 11. The nuclear reactor as defined in claim 10, further comprising a core shroud and a shroud head, wherein said means for catalyzing water recombination is mounted on said shroud head and circumferentially encompasses said steam/water separating means. 12. The nuclear reactor as defined in claim 10, further comprising a core shroud and a shroud head, wherein said means for catalyzing water recombination is mounted on said shroud and extends across said downcomer annulus. 13. The nuclear reactor as defined in claim 9, wherein said catalytic material comprises entangled or crimped stainless steel plated or alloyed with a noble metal. |
description | Referring now to the drawings in detail and in particular to FIG. 1 there is generally shown a boiling water reactor 10 of a nuclear power plant for commercially generating electricity. The reactor 10 generally includes a reactor pressure vessel 12 and a reactor recirculation system 14. The reactor recirculation system 14 generally comprises a plurality of reactor recirculation loops, illustrated by loops 16 and 18, hydraulically connected in parallel with the reactor pressure vessel 12. When generating power during normal online operations, primary coolant (high purity water containing ppm levels of various ions and, in some cases, dissolved hydrogen gas) is pumped by feedwater pumps (not shown) into the reactor pressure vessel 12 through an inlet nozzle 19 and steam is generated with the reactor pressure vessel 12. The steam flows out of the pressure vessel 12 through an outlet nozzle 20 and then to a turbine (not shown) which generates the electrical power. The reactor recirculation system 14 facilitates the flow of primary coolant to fuel assemblies in the central core regions in the pressure vessel 12. A commercial facility embodying this boiling water reactor design is the Oyster Creek Plant near Forked River, N.J. The reactor pressure vessel 12 includes a bottom head 22 with a sidewall 24 extending vertically to a flange 26. A removable head 28 has a flange 30 that may be bolted to the reactor pressure vessel flange 26. The reactor pressure vessel 12 has a core shroud 32 and a core plate 34, which define a central core region 36 for containing removable fuel assemblies 38. The core shroud 32 has a removable upper end 40 that may be removed in order to remove the fuel assemblies 38. The core shroud 32 (or, equivalently, in a similar reactor design, a supporting skirt [not shown] supporting the core shroud 32) is spaced from the reactor pressure vessel wall 24 by a structural ring member 42. The pressure vessel wall 24, core shroud 32 and ring member 42 define an annulus region 44 surrounding the central core region 36. The annulus region 44 frequently is referred to as a xe2x80x9cdowncomerxe2x80x9d or a xe2x80x9cdowncomer annulusxe2x80x9d. The reactor pressure vessel bottom head 22 and the core plate 34 define a lower internals region 46 which is in fluid flow communication with the central core region 36 via flow holes 48 in the core plate 34. Each reactor circulation loop 16 and 18 of the reactor circulation system 14 shown in FIG. 1 generally includes a centrifugal pump 56 with a pump suction nozzle and a pump discharge nozzle. The pump 56 may have a nominal capacity of up to about 50,000 gallons per minute or more. The pump suction nozzle is connected with piping 58 extending from one or more nozzles, illustrated by nozzle 60 in FIG. 1, in the pressure vessel wall for fluid flow connection with the annulus region 44. The pump discharge nozzle is connected with piping 62 extending to one or more nozzles, illustrated by nozzle 64, in the pressure vessel wall 24 for fluid flow connection with the lower internals region 46 of the reactor pressure vessel 12. When generating power during normal online operations, the primary coolant pumped through the inlet nozzle 19 into the annulus region 44 flows through the recirculation system 14, into the lower internals region 46, up through flow holes 48 in the core plate 34 and fuel assemblies 38 in the central core region 36 (where steam is generated) and up through steam/condensate separators (not shown) supported on the core shroud head 40. The separated condensate drains back to the annulus region 44. The steam flows up into the upper portion of the reactor pressure vessel 12, through steam dryers (not shown) and then out of the pressure vessel 12 through outlet nozzle 20. FIG. 2 shows a different boiling water reactor design which employs internal jet pump assemblies 76 disposed in the annulus region 44 in the reactor circulation loops 16 and 18 for circulating coolant from the annulus region 44 to the lower internals region 46. Each jet pump assembly 76 includes inlet piping 78 with a jet nozzle 80 in fluid flow communication with one of the reactor recirculation loops 16 or 18. A mixing assembly 82 has a suction inlet end 84 spaced from the jet pump nozzle 80 in fluid flow communication with the annulus region 44. The primary coolant around the suction inlet end 84 in the annulus region 44 is entrained by the primary coolant flowing out through the jet pump nozzle 80 and the two fluids are mixed together in the mixing assembly 82. The mixing assembly 82 is connected with a diffuser assembly 86 having an outlet end 88 in fluid flow communication with the lower internals region 46. U.S. Pat. No. 5,515,407 entitled xe2x80x9cJet Pump Assembly For Recirculating Coolant Through a Recirculation Loop Of A Boiling Water Reactor Vesselxe2x80x9d is incorporated by this reference for its detailed description of the structure of such an assembly. As is shown in FIG. 3 and as is described in U.S. Pat. No. 5,515,407, jet pump assemblies 76 are conventionally arranged in pairs with an inlet riser pipe 90 extending directly from an inlet nozzle 64 or via a header (not shown) to a header 92 and then to the inlet piping 78, which may be a piping elbow with a lifting eye 94. The assembly comprising the header 92 and the inlet piping 78 to each jet pump assembly 76 is frequently referred to as a xe2x80x9cramsheadxe2x80x9d. The riser pipe 90 and the ramshead are maintained in place by a holddown assembly 96. The mixing assembly 82 and the diffuser assembly 86 are held against the riser pipe 90 by restrainers 98. Also, the diffuser assembly outlet end 88 may be fit into a fitting 100 extending to or through the ring member 42 and to a connector (not shown) for providing primary coolant to the lower internals region 46. As discussed above, after generating electric power during online operations, it is desirable to decontaminate boiling water reactors and their recirculation systems 14 but not necessarily the central core regions 36. In accepted commercial decontamination processes such as, e.g., the LOMI, CAN-DEREM, CAN-DECON, Citrox and various permanganate processes, low oxidation state metal ions, permanganates, oxalates, citrates, EDTA and other agents are added to the primary coolant to generate a decontamination solution. The decontamination solution is then circulated past the activated surfaces to dissolve and break up the radioactive oxide films that have formed and release the activated metal ions. In accordance with the practice of the present invention, a decontamination solution is circulated through at least one of the reactor recirculation loops 16 or 18 and the annulus region 44 without circulating the decontamination solutions through the central core region 36. Preferably, the decontamination solutions are circulated through all of the recirculation loops 16 and 18. FIGS. 4-7 show various modifications to jet pump assemblies 76 of boiling water reactors, such as the paired jet pump assembly arrangement illustrated in FIG. 3, for at least restricting the flow of the decontamination solution from the annulus region 44 into the lower internals region 46 so that the decontamination solution flowing through the annulus 44 and the recirculation system 14 will not circulate through the central core region 36. Preferably, the turbulence in the lower internals region 46 (if any) is sufficiently low that substantial amounts of the decontamination solution will not splash through the holes 48 in the core plate 34 and into the central core region 36 because this could unnecessarily generate additional activated ions in the decontamination solution which would need to be removed and thereby reduce the efficiency of the process. FIGS. 4 and 5 show modified jet pump assemblies 76 of FIG. 2 fed by a riser pipe 90 from which at least part of each jet pump assembly has been removed. The remaining part of each jet pump assembly 76 in the pressure vessel 12 which is in fluid flow communication with the lower internals region 46 is covered by a cap 110. Each cap 110 may be placed over the remaining part of the jet pump assembly 76 to provide an umbrella type cover that substantially prevents the liquid which may fall or splash from entering into the diffuser assembly 86. A liquid tight cap seal may be employed where the liquid level in the annulus 44 were to be maintained near or above the level of the cap 110 and it is desired to substantially prevent decontamination solution from flowing into the lower internals region 46. FIG. 4 illustrates a jet pump assembly modification wherein the jet pump nozzles 80 and the mixing assemblies 82 have been removed, e.g., by underwater cutting, and the diffuser assemblies 86 covered by caps 110. Also, as shown, the jet pump nozzles 78 may be removed and adapters 112 such as orifice plates or other flow devices may be welded or attached to the ends of the inlet pipes 92. The adapters 112 may be employed to direct decontamination solution downwardly to avoid creating a geyser in the annulus 44. This can be particularly important to reduce splashing over the core shroud 32 in a decontamination practice where the core shroud head 40 and fuel assemblies 38 are removed. FIG. 5 illustrates another jet pump assembly modification wherein only the suction inlets 84 of the mixing assemblies 82 have been removed and the remaining portions of the mixing assemblies 82 capped. The flow of decontamination solution in each recirculation loop 16 and 18 usually may be controlled by variable speed circulation pumps 56 or by flow control valves (not shown) to maintain the required net positive suction head (NPSH) of the pumps and to limit vibrations on the temporary adapters 112. Also, the recirculation pumps 56 may be operated at a rate that will provide sufficient energy input into the decontamination solution in order to heat up and maintain the system at a temperature of about 180xc2x0 F. to about 250xc2x0 F. for the chemical decontamination agents to be effective in an acceptable period of time. An overpressure with a gas such as air or nitrogen may be provided in the reactor vessel 12 to prevent boiling and to provide the required NPSH. Alternatively, where it is not possible or undesirable to operate the recirculation pumps 56, the Residual Heat Removal Pumps (not shown) or external pumps (not shown) may be employed to circulate the decontamination solution. FIG. 6 illustrates a jet pump assembly modification wherein some of the jet pump assemblies 76 are modified and other assemblies 76 are not modified. This modification permits decontamination solution to be pumped into the lower internals region 46 and circulate upwardly through the remaining parts of the modified assemblies 76 without circulating the solution through the central core region 36. It is noted, however, that substantial turbulence in the lower internals region 46 should be avoided because turbulence may cause some decontamination solution to splash through the core plate 34 depending upon the numbers of assemblies 76 which are modified. FIG. 7 illustrates a piping arrangement where one or more jet pump nozzles 80 of modified jet pump assemblies 76 (as shown in FIG. 6) of one recirculation loop are connected by jumper pipes or hoses 120 with jet pump nozzles 80 of modified jet pump assemblies 76 (as shown in FIG. 6) of another recirculation loop. A recirculation pump 56 of one recirculation loop may then be operated to pump decontamination solution up through riser pipes 90 and headers 92 of the one recirculation loop, through jumpers 120, into headers 92 and down riser pipes 90 of the other recirculation loop, through the other recirculation loop including its recirculation pump (which would not be operating), and into the annulus region 44 of the pressure vessel 12. Advantageously, the flow could be reversed by operating the other recirculation pump 56. FIG. 8 illustrates a modification of the reactor pressure vessel 12 of FIG. 1 (which does not employ jet pumps) wherein flow holes in the ring member 42 or in the lower portion of the core shroud 32 below the core plate 34 would permit decontamination solution in the lower internals region 46 to circulate into the annulus region 44 without circulating through the central core region 36. The flow holes could be uncovered annulus manways 130 and/or flow holes cut into the ring member 42 or in the lower part of the core shroud 32 (or, equivalently, its supporting skirt) below the core plate 34. With each of these modifications, additional plant pumps could be operated to circulate the decontamination solution through appurtent systems if needed. For example, Reactor Water Clean-Up, Residual Heat Removal, High Pressure System Injection, Low Pressure System Injection Systems and others could be decontaminated if desired. Entrained particulates and activated ions in the circulating decontamination solutions may be removed in filters (not shown) and on cation resins (not shown), respectively, during the course of the decontamination operations. Then, at the conclusion of the decontamination operations, the primary coolants may be cleaned up on the resins. If plants are to be decommissioned, decommissioning operations may continue without restoring the reactor pressure vessels to their initial conditions. However, if plants are to be returned to online power generating operations, then pressure vessels may need to be repaired or further modified. While a present preferred embodiment of the present invention has been shown and described, it is to be understood that the invention may be otherwise variously embodied within the scope of the following claims of invention. |
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062333003 | description | DETAILED DESCRIPTION OF THE INVENTION FIG. 1 is a sectional view with parts cut away of a nuclear reactor pressure vessel (RPV) 10. RPV 10 includes a generally cylindrical side wall 12. A core shroud 14 of generally cylindrical shape is located within RPV 10 and surrounds the reactor core (not shown). Shroud 14 is supported by a shroud support structure 16. A core plate 18 is spaced below a top guide 20 within RPV 10. Core plate 18 and top guide 20 are coupled to shroud 14. Shroud 14 includes a shroud head 22 coupled to top guide 20. Particularly, top guide 20 includes a flange 24 and shroud head 22 includes a flange 26 configured to engage top guide flange 24. More particularly, top guide flange 24 engages shroud head flange 26 to form a shroud head to top guide interface 28. RPV 10 is shown in FIG. 1 as being shut down with many components removed. For example, and in operation, many fuel bundles and control rods (not shown) are located in the area between top guide 20 and core plate 18. In addition, and in operation, steam dryers and many other components (not shown) are located in the area above top guide 20. Also, steam separators 30 are permanently coupled to shroud head 22. Top guide 20 is a latticed structure including a plurality of top guide beams 32 defining top guide openings 34. Core plate 18 includes a plurality of openings 36 which are substantially aligned with top guide openings 34 to facilitate positioning the fuel bundles between top guide 20 and core plate 18. Fuel bundles are inserted into the area between top guide 20 and core plate 18 by utilizing top guide openings 34 and core plate openings 36. Particularly, four fuel bundles are inserted through a top guide opening 34, and are supported horizontally by an orificed fuel support (not shown) inserted in core plate opening 36, core plate 18, and top guide beams 32. Shroud 14, core plate 18, and top guide 20 limit lateral movement of the core fuel bundles. FIG. 2 is an enlarged exploded view of section A of shroud head to top guide interface 28 shown in FIG. 1. Interface 28 is formed by a top surface 38 of top guide flange 24 and a bottom surface 40 of shroud head flange 26. Top guide flange 24 includes a plurality of frusto-conical shaped guide pins 42 extending from top surface 38. Guide pins 42 are located around the circumference of top guide flange 24. Shroud head flange 26 includes a plurality of corresponding guide pin openings 44 extending from shroud head flange bottom surface 40 through shroud head flange 26. Each guide pin opening 44 is configured to receive a frusto-conical guide pin 42. FIGS. 3, 4, and 5 are enlarged exploded sectional views of section B of the shroud head to top guide interface shown in FIG. 2 and illustrate various positions of shroud head flange 26 in relation to top guide flange 24 during assembly. As described above, shroud head flange 26 includes a plurality of guide pin openings 44 configured to align with guide pins 42 located on top guide flange 24. Each guide pin opening 44 includes a frusto-conical portion 46, defined by an inside surface 48 of shroud head flange 26, that extends through shroud head flange 26 from bottom surface 40 and has a slope equal to the slope of frusto-conical guide pins 42. Each guide pin opening 44 also includes a cylindrical portion 50 that extends from the small base 52 of frusto-conical portion 46 of guide pin opening 44 to a top surface 54 of shroud head flange 26. The diameter of frusto-conical guide pin opening 44 at bottom surface 40 of shroud head flange 26 is configured to be larger than the diameter of frusto-conical guide pin 42 immediately adjacent top surface 38 of top guide flange 24. Because of the frusto-conical shape, guide pins 42 include a first base 56 and a second base 58, with first base 56 having a larger diameter than second base 58. Second base 58 is located immediately adjacent top surface 38 of top flange 24. During assembly, shroud head 22 is suspended from an overhead crane and lowered into engagement with top guide flange 24. Particularly, shroud head 22 is lowered so that each frusto-conical guide pin 42 extending from top guide flange 24 aligns with a corresponding guide pin opening 44 in shroud head flange 26. Shroud head 22 is lowered until bottom surface 40 of shroud head flange 26 is in surface to surface contact with top surface 38 of top guide flange 24. The conical shape of guide pins 42 and the conical shape of guide pin openings 44 provide greater clearance between guide pin 42 and guide pin opening 44 as shroud head flange 26 and top guide flange 24 approach engagement than the clearance when the guide pins are cylindrically shaped. In the assembled condition, a distance C between guide pin 42 and guide pin opening 44 is less than about 1.0 millimeters (see FIG. 5). At a position where the distance between the two flanges 24 and 26 is equal to the height of guide pins 44, distance C between each guide pin 42 and each guide pin opening 44 is about 4.0 millimeters when a cone angle D of guide pin 42 is 60 degrees (see FIG. 3). Cone angle D is measured in reference to first base 56 of guide pin 42. Of course, distance C is dependent on the value of cone angle D and the relative position of shroud head flange 26 and top guide flange 24. For example, at an intermediate position shown in FIG. 4, distance C is about 2.3 millimeters for a cone angle D of 60 degrees. Cone angle D may vary over a wide range, for example from about 20 to about 80 degrees. Preferably, cone angle D is about 40 to about 75 degrees, more preferably about 55 to about 65 degrees. If cone angle D is too high the conical shape of guide pin 42 approaches that of a cylinder and may not over come the inherent alignment problems of a cylindrical guide pin. If cone angle D is too low, guide pin 42 may not provide sufficient restriction of horizontal movement during a seismic event. The above described top guide to shroud head interface 28 includes frusto-conical guide pins 42 and guide pin openings 44 that provide for suitable clearances between guide pins 42 and guide pin openings 44 to accommodate flexing of shroud head flange 26 during installation. Additionally, frusto-conical guide pins 42 and guide pin openings 44 provide less than 1.0 millimeter of clearance in the installed position to minimize the impact loading on guide pins 42 and openings 44 caused by horizontal seismic accelerations during a seismic event. In an alternate embodiment, guide pin openings 44 do not extend through shroud head flange 26. In this embodiment, each guide pin opening 44 includes frusto-conical portion 46 configured to receive a guide pin 42, but does not include cylindrical portion 50. In another embodiment, guide pins 42 extend from shroud head flange 26 instead of top guide flange 24. In this embodiment, corresponding guide pin openings 44 are located in top guide flange 34. In still another embodiment, some guide pins 42 extend from top guide flange 24 and some guide pins 42 extend from shroud head flange 26. Additionally, each guide pin 42 has a corresponding guide pin opening 44 located in the opposing flange. Particularly, each guide pin 42 extending from top guide flange 26 has a corresponding guide pin opening located in shroud head flange 26, and each guide pin 42 extending from shroud head flange 26 has a corresponding guide pin opening 44 located in top guide 24. From the preceding description of various embodiments of the present invention, it is evident that the objects of the invention are attained. Although the invention has been described and illustrated in detail, it is to be clearly understood that the same is intended by way of illustration and example only and is not to be taken by way of limitation. Accordingly, the spirit and scope of the invention are to be limited only by the terms of the appended claims. |
description | This application claims priority to Korean Patent Application No. 2003-3793, filed on Jan. 20, 2003, in the Korean Intellectual Property Office, the disclosure of which is incorporated herein in its entirety by reference. The present invention relates to a mobile electronic unit, and more particularly, to a method of automatically cutting off the power to a mobile electronic unit by determining the presence of low battery voltage using a battery voltage detection unit during the booting period of the mobile electronic unit and during a period after booting is completed. A digital camera, which is a type of mobile electronic unit, works on a fundamentally different concept than an existing film-type camera. The digital camera photographs a subject through an optical lens system and stores the corresponding photographed image information as digital data in a memory card. The digital data stored in the memory card can be processed in a desired format using a computer and can be easily transmitted via a network. Thus, the demand for digital cameras is expected to substantially increase. The digital camera operates using power supplied from an AC adapter or a battery. A primary battery, such as an alkaline or lithium, and a secondary battery, such as a lithium-ion (Li-ion) or lithium hydrogen, have been widely used. A nominal voltage and an internal capacity of the secondary battery are higher than those of the primary battery, and thus, the secondary battery can guarantee a more stable operation in case of a low battery voltage. However, if the digital camera uses a current higher than a standard consumption current, such as during a flash check after Iris, shutter 1, and shutter 2 in a low battery voltage, the primary battery does not withstand the load. (In a camera with a two-stage shutter release, the shutter 1 mode operates by an ON signal of a first switch in a two-stage shutter release button for use in performing the camera's AE/AF operations and the shutter 2 mode operates by an ON signal of a second switch in the two-stage button arrangement.) Thus, the operating stability of a variety of IC (integrated circuit) units, such as a digital signal processing unit and a microcontroller of the digital camera, cannot be guaranteed, and communication errors between the IC units occur. In order to prevent the digital camera from malfunctioning, the power should be cut off when low battery voltage is present. The present invention provides a method of automatically cutting off the power of a mobile electronic unit by determining that a low battery voltage is present by using a battery voltage detection unit during the booting period of the mobile electronic unit and during a period after the booting is complete. Another embodiment of the present invention is directed to a method of automatically cutting off the power in a mobile electronic unit when a battery voltage is detected in a first check section during a booting period of the mobile electronic unit that is not in a normal state. A further embodiment of the present invention is directed to a method of automatically cutting off the power of a mobile electronic unit when the battery voltage detected in the first check section is in the normal state and when a battery voltage is detected in a second check section during a period of time after the booting period of the mobile electronic unit is completed that is not in a normal state. Preferably, the first check section includes at least one of a first sub-section which corresponds to an on-time section of primary elements at an initial stage of booting, and a second sub-section including a driving period of elements where a current higher than a standard consumption current is used. The method of automatically cutting off the power of a mobile electronic unit of the present invention further includes (a1) checking the battery voltage before a predetermined time from a starting time of the second sub-section; (a2) checking the battery voltage after a predetermined time from an ending time of the second sub-section; and (a3) generating a power cutting off signal when a difference between the battery voltages checked in steps (a1) and (a2) is more than a predetermined reference value. Preferably, the second check section includes at least one of a third sub-section which corresponds to a stabilization period after booting of the mobile electronic unit is completed, a fourth sub-section including a driving period of elements at an initial stage of a predetermined operation mode, and a fifth sub-section including a performance period of an operation mode where a current higher than a standard consumption current is used. The method of automatically cutting off the power of a mobile electronic unit also includes (a) checking the battery voltage before a predetermined time from a starting time of the fourth sub-section; (b) checking the battery voltage after a predetermined time from an ending time of the fourth sub-section; and (c) generating a power cutting off signal when the difference between the battery voltages checked in steps (a) and (b) is greater than a predetermined reference value. Preferably, the method of the present invention further includes comparing a battery voltage detected during a performance period of an operation mode where a current higher than a standard consumption current is used with a threshold voltage and generating a power cut-off signal when the detected battery voltage is less than the threshold voltage for more than a predetermined number of times. Hereinafter, a method of automatically cutting off the power in case of a low battery voltage according to an embodiment of the present invention will be described in detail with reference to the accompanying drawings. FIG. 1 shows an apparatus using a method of automatically cutting off the power in case of a low battery voltage. The apparatus includes a battery voltage detection unit 11 and a low battery voltage determination unit 13. The battery voltage detection unit 11 measures the amount of voltage of a battery unit (not shown) and divides the amount by two, resulting in figure referred to as a “voltage value.” The battery voltage detection unit 11 transmits the voltage value to the low battery voltage determination unit 13. The low battery voltage determination unit 13 corresponds to a microcontroller of a mobile electronic unit, for example, a digital camera. An algorithm for automatically cutting off the power in case of a low battery voltage is stored in the low battery voltage determination unit 13. According to the algorithm, the normal state or low voltage state of the battery is determined by comparing the voltage value to a predetermined set of values according to the specification of the corresponding mobile electronic unit. If it is determined that the battery is in a low voltage state, a power cutting off signal is generated that cuts off the power supplied to a power supply unit (not shown). FIG. 2 is a circuit diagram showing a structure of a battery voltage detection unit 11 of FIG. 1. The battery voltage detection unit 11 includes a first and second transistor Q1 and Q2 and a first through sixth resistors R1 to R6. When the power switch of the mobile electronic unit such as a digital camera is turned on, an enable signal is applied to the base terminal of the second transistor Q2 via a resistor R4, and the second transistor Q2 is turned on. As such, a battery voltage applied to an emitter terminal of the first transistor Q1 is divided by two voltage-division resistors R5 and R6 that are connected to a collector terminal of the first transistor Q1. A connection point CP between the resistors R5 and R6 becomes a check point CP of the battery voltage. A capacitor C1 is connected between check point CP and ground. When resistors R5 and R6 are designed to have the same resistance, the battery voltage is divided exactly in half. If two alkaline batteries having a nominal voltage of 1.5V are used, the battery voltage ranging from about 1.5V to 1.0V is checked at the check point CP. FIG. 3 is a flowchart of the method of automatically cutting off the power when a low battery voltage is present, according to an embodiment of the present invention. The method illustrated by FIG. 3 is performed by the low battery voltage determination unit 13. In FIG. 3, step 31 refers to the step wherein the low battery voltage determination unit 13 determines in, a first check section during the booting period of a mobile electronic unit if the battery voltage is in a normal state. If the battery voltage determination unit 13 determines that the battery voltage is not in the normal state, in step 33, a signal to cut off the power of the mobile electronic unit is generated. The first check section includes at least one of a first sub-section which corresponds to an on-time section of primary elements at an initial stage of booting and a second sub-section including a first driving period of elements where a current higher than a standard consumption current is used. In the case of a digital camera, the primary elements that drive the first sub-section, are a power supply unit, an LCD, and a CCD. The second sub-section includes the charging section of the flash capacitor during the booting period. In the first sub-section, if the checked battery voltage is less than a predetermined reference value, that is, about 1.17V, then the battery is considered to be in a low voltage state, and the power cutting off signal is generated. The predetermined reference value will vary, depending on the specifications of the mobile electronic unit. In the second sub-section, a low voltage state of the battery voltage is determined by a method which will be described later in FIG. 4. If the battery voltage detected by the first check section is in the normal state (Step 31), then in Step 35, the low battery voltage determination unit 13 determines in a second check section after the booting of the mobile electronic unit is completed whether a battery voltage is in the normal state. The second check section includes at least one of a third sub-section which corresponds to a stabilization period after booting of the mobile electronic unit is completed, a fourth sub-section including a driving period of elements at an initial stage of a predetermined operation mode, and a fifth sub-section including a performance period of an operation mode where a current higher than a standard consumption current is used. In case of the digital camera, the fourth sub-section includes a shutter driving section at an initial stage of a photographing mode, and the fifth sub-section includes a charging section of the flash capacitor after a maximum light-emitting photographing mode is executed. As a result of step 35, if the battery voltage is not in the normal state, in step 33, a signal to turn off the power of the mobile electronic unit is generated. As a result of step 35, if the battery voltage is in the normal state, in step 37, a determination is made whether the battery is in the normal state, indicating that the battery is operating according to an operation mode. Step 37 will be described later. In the third sub-section a signal to cut off the power is generated when the checked battery voltage is less than a predetermined reference value, that is, a threshold voltage, i.e., a minimum voltage of, for example, an alkaline battery at which it is possible for the mobile electronic unit to be normally operated. In the fourth and fifth sub-sections, a low voltage state of the battery voltage is determined by a method which will be described later with respect to FIG. 4. In Step 33, a signal is generated to cut off the power of the power supply, resulting in the power being turned off or a warning sound being generated. FIG. 4 is a flowchart showing in detail a method of determining a low battery voltage, which is performed in steps 31 and 35 of FIG. 3. Referring to FIG. 4, in step 41, the battery voltage is checked at a predetermined time, for example, 25 ms prior to each starting time of the second sub-section, the fourth sub-section and the fifth sub-section. In step 43, the battery voltage is checked at a second predetermined time, for example, 25 ms after each ending time of the second sub-section, the fourth sub-section and the fifth sub-section. The predetermined time will vary, depending on the specification of the particular mobile electronic unit. In step 45, the difference between the battery voltages checked in steps 41 and 43 is compared with a predetermined reference value, for example, 50 mV. As a comparison result, in step 47, if the difference between the battery voltages is greater than the reference value, then the battery is considered to be in the low voltage state, and the signal to cut off the power is generated. If the difference between the battery voltages is less than the reference value, then the battery is considered to be in the normal state. FIG. 5 is a graph showing a battery voltage check section according to the present invention. The battery voltage check section is classified into a booting period and a period after booting is completed. A check method used in each section is previously programmed and stored in the low battery voltage determination unit 13. The booting period includes sections a through f, and the period after booting is performed includes sections g through k. Here, section a is a key check section, section b is an on-time section of a power supply unit, an LCD, and a CCD, section c is an Iris and shutter driving section, section d is a zoon motor driving and reverse-break section, section e is a booting sound and LCD backlight on section, section f is a flash capacitor charging and focusing section at an initial stage of booting, section g is a stabilization section after booting, section h is a shutter 1 section, section i is an Iris interworking section including shutter 1, section j is a shutter 2 section including a shutter sound, and section k is a flash capacitor charging section after photographing in a light-emitting mode, respectively. In the present invention, preferably, during the booting period, at least one of sections b and f is used, and during the period after booting is completed, at least one of sections g, h, and k is used. In other words, in the present invention, section b corresponds to the first sub-section, section f corresponds to the second sub-section, section g corresponds to the third sub-section, section h corresponds to the fourth sub-section, and section k corresponds to the fifth sub-section, respectively. FIG. 6 is an enlarged graph showing a first check voltage in a first sub-section of FIG. 5. The first check voltage in the first sub-section is detected with respect to section b 61 at the check point CP. The first check voltage is used to determine whether an alkaline battery is actually used. The alkaline battery voltage used in the digital camera ranges from about 3V to 2V, and the voltage where the battery voltage is divided exactly in half, that is, maximum 1.5V to minimum 1.0V is a detection object voltage at the check point CP. Section b 61 is a section where the driving of the LCD, the CCD, and the power supply unit starts, and a voltage in section b 61 is set to a first check voltage. In this case, if a reference value of the first check voltage is set to about 1.17V and the first check voltage is less than 1.17V, then the state of the battery is considered to be in a low voltage state, and the power of the digital camera is automatically cut off. FIG. 7 is an enlarged graph showing a second check voltage in a second sub-section of FIG. 5. The second check voltage is detected with respect to section f 71 at the check point CP. If it is determined that the first check voltage detected in the first sub-section is in the normal state, then the first check voltage is considered an internally stable power source, and while the initial booting is continuously occurring, the second check voltage is measured. In a charging section of a flash capacitor, that is, section f 71 set during an initial booting period, a larger amount of current consumption compared to other sections occurs. If the battery is in a low voltage state, there is a distinguishable difference between the battery voltage at a predetermined time, for example, 25 ms prior to a charging section of the flash capacitor and the battery voltage at the predetermined time, for example, 25 ms after the charging section of the flash capacitor. Thus, voltages before and after flash capacitor charging at two check points 73 and 75 before and after a predetermined time of a section where charging is actually performed, respectively, are checked. If a voltage difference therebetween is smaller than a predetermined reference value, for example, about 50 mV, then the battery is considered to be in the normal state. If the voltage difference therebetween is greater than the predetermined reference value, then the battery is considered to be in a low voltage state, and the power of the digital camera is automatically cut off. FIG. 8 is an enlarged graph showing a third check voltage in a third sub-section of FIG. 5. The third check voltage is detected with respect to section g 81 at the check point CP. If it is determined that the second check voltage detected in the second sub-section is in the normal state, then the third check voltage is measured. The third check voltage is compared with a threshold voltage, i.e., a minimum voltage of, for example, an alkaline battery at which it is possible for the mobile electronic unit to be normally operated. If the third check voltage is smaller than the threshold voltage, the third check voltage is further checked during a predetermined period, for example, within about 10 ms, a predetermined number of times, for example, at least more than twice. If each of the further-checked third check voltages is less than the threshold voltage, the state of the battery is considered to be in a low voltage state, and the power of the digital camera is automatically cut off. FIG. 9 is an enlarged graph showing a fourth check voltage in a fourth sub-section of FIG. 5. The fourth check voltage is detected with respect to section h 91 at the check point CP. The first through third check voltages are voltages checked during the booting period and during the period after booting is completed, whereas the fourth check voltage is a voltage check with respect to each operation mode generated after booting is completed. In other words, in a photographing mode where the most amount of current consumption occurs, e.g., when a voltage drop occurs in shutter 1. Like in the second check voltage, if a voltage difference detected at two check points 93 and 95 that are set at a point 25 ms before a shutter 1 operation mode and at a point 25 ms after the shutter 1 operation mode, is smaller than a predetermined reference value, for example, about 50 mV, then the state of the battery is considered to be in a normal state. If the voltage difference is greater than the predetermined reference value, the state of the battery is considered to be in a low voltage state, and the power of the digital camera is automatically cut off. FIG. 10 is an enlarged graph showing a fifth check voltage in a fifth sub-section of FIG. 5. The fifth check voltage is detected with respect to section k 101 at the check point CP. Even though it is determined that all of the first through fourth check voltages are in the normal state, remarkable voltage drop occurs after photographing in a maximum light-emitting mode in a low voltage state. Thus, voltages are checked before and after flash capacitor charging at two check points, before and after a predetermined time, for example, 25 ms of a section where charging of the flash capacitor is actually performed after photographing in the maximum light-emitting mode, respectively. If a voltage difference therebetween is smaller than a predetermined reference value, for example, about 50 mV, then the state of the battery is considered to be in the normal state. If the voltage difference therebetween is greater than the predetermined reference value, the state of the battery is considered to be in a low voltage state, and the power of the digital camera is automatically cut off. In addition, in all sections after booting is completed, that is, sections h through k, a battery voltage detected for a predetermined time, for example, two seconds whenever a corresponding operation is performed, is compared with a threshold voltage, i.e., a minimum voltage of, for example, an alkaline battery at which it is possible for the mobile electronic unit to be normally operated. If the detected battery voltage is less than the threshold voltage for a predetermined number of times, for example, at least more than three times, the power of the digital camera is automatically cut off. In the above-described embodiment, the first through fifth check voltages are sequentially checked to determine whether the battery is in a low voltage state. However, the second check voltage and the fourth check voltage are sequentially checked, or the second check voltage, the fourth check voltage, and the fifth check voltage are sequentially checked to determine whether the battery is in the low voltage state. In addition, the first and second check sections and the first through fifth sub-sections may be diversely set according to the specification of the corresponding mobile electronic unit. In the above-described embodiments, the digital camera has been shown. However, the method of automatically cutting off power according to the present invention can be used in a mobile electronic unit, such as a personal digital assistant (PDA) and a mobile phone in which both a primary battery and a secondary battery are used, without modifying hardware of the mobile electronic unit. The present invention can be realized as a computer-readable code on a computer-readable recording medium. A computer-readable medium may be any kind of recording medium in which computer-readable data is stored. Examples of such computer-readable media include ROMs, RAMs, CD-ROMs, magnetic tapes, floppy discs, optical data storing devices, and carrier waves (e.g., transmission via the Internet), and so forth. Also, the computer-readable code can be stored on the computer-readable media distributed in computers connected via a network. Furthermore, functional programs, codes, and code segments for realizing the present invention can be easily analogized by programmers skilled in the art. In a mobile electronic unit in which both a primary battery and a secondary battery are used, a low voltage state of a battery may cause the malfunctioning of some of the elements of the unit, in particular, the digital signal processing unit and the microcontroller. As described above, embodiments of the present invention detect when a low voltage state of a battery exists during the booting period and at predetermined intervals after booting is completed, and the power of the mobile electronic unit is automatically cut off, so that damage of the respective elements is prevented and a stable operation of the mobile electronic unit is guaranteed. While this invention has been particularly shown and described with reference to preferred embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the spirit and scope of the invention as defined by the appended claims. |
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abstract | The burnout of a fuel element in a reactor is determined by first transferring a fuel element from a reactor to a measuring position and then subjecting the transferred fuel element at the position to a neutron flux. A first detector measures the total γ radiation emitted by the transferred fuel element and thereafter, if the radiation measured by the first detector exceeds a predetermined first limit, the transferred fuel element is returned back to the reactor. If not, a second detector measures a magnitude of high energy γ radiation above 1 MeV emitted by the transferred fuel element and thereafter only if the radiation measured by the second detector exceeds a predetermined second limit, the transferred fuel element is transferred back to the reactor. The element is not returned to the reactor if the radiation measured by the second detector is below the second limit. |
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059096543 | summary | FIELD OF INVENTION The present invention relates to the field of processing organic waste, "processing" in the present case referring to the breaking down of said waste via the thermal route with the primary aim of affording opportunities for reducing its volume to thereby lessen handling and storage problems. More particularly, it concerns a new method and new apparatus for processing solid organic sulphur or chloride containing waste in which the carbon content of solid residue resulting from an initial pyrolysis of the waste is gasified. BACKGROUND TO THE INVENTION The new method of the present invention not only achieves the aim of volume reduction, but also provides, for example, such benefits as the elimination of sulphur or chloride content from exhaust gases, and similarly any radioactive content, in an effective and straight forward manner. The invention is therefore especially useful for the processing of ionic exchange media from nuclear facilities, which media display a certain degree of radioactivity and therefore would otherwise require conventional measures in relation to ultimate waste disposal and deposition. The nuclear industry annually produces a significant amount of waste which is classified as radioactively contaminated ion exchange media. Such waste is managed in various fashions prior to ultimate disposal in bedrock chambers or shallow land burial. This management is technically complex and as a rule leads to increased volumes which influences storage costs. A process resulting in diminished volume at reasonable cost is therefore highly commercially desirable. Ion exchange medium is an organic material. The base is usually a styrene polymer with grafted sulfonic acid and amine groups. The material is therefore burnable, but air is supplied during combustion and sulphur and nitrogen oxides are formed which in turn must be separated in some manner. Additionally, during combustion the temperature becomes sufficiently high for radioactive cesium to be partially vaporized. The residual radioactivity will also accompany the resulting fly ash to some extent. This necessitates a very high performance filter system. Accordingly, both technical and economic problems are typically associated with standard combustion techniques. An alternative to straight combustion is pyrolysis. However, previously known pyrolysis methods in this technical field are deficient in several aspects and in particular no one has earlier succeeded in devising a pyrolysis process which provides a comprehensive solution to the problem of sulphur and nitrogen-containing radioactive waste, and to do so under acceptable economic stipulations. The following can be mentioned as examples of the known technology in this respect: U.S. Pat. No. 5,424,042 to J. Bradley Mason et al. describes a system for vitrification of nuclear waste (incorporating of a portion of the nuclear waste into a stable glass matrix) including several subsystems: a feed conditioning system for conditioning "dry waste," "wet waste" or ion exchange resins and "liquid waste"; a feed preparation system for blending the waste types; a feed melter chamber with an upper zone and a lower zone for oxidizing the waste into ash and off-gas; a glass handling system for packing and storing the glass product; and an off-gas cleaning and control system. U.S. Pat. No. 5,470,544 describes a system for the detoxification of hazardous waste utilizing a moving bed evaporator and a steam-reforming detoxification reactor. U.S. Pat. No. 5,427,738 also describes a system for the detoxification of solid waste which first mechanically particularized the waste in a spinning knife cutter, size reduction grinder or like device, and then subjects the particularized waste to a gas flow of a hot gas in the range of 250-750.degree. C. The particularized waste is agitated to enhance exposure to the hot gaseous flow. SE-B 8405113-5 which describes single stage pyrolysis in a fluidised bed followed by conversion of tars in the resulting gas to non-condensable gas using limestone as catalyst. U.S. Pat. Nos. 4,628,837, 4,636,335, and 4,654,172 describe pyrolysis of ion exchange resins where the pyrolysis is carried out in two stages. Both of these stages, however, are directed towards pyrolysis of the ion exchange media itself, i.e. the solid product. Speaking generally, both stages moreover are carried out at relatively low temperatures. Furthermore, none of these specifications recites any comprehensive solution to the problem of solid organic sulphur or chloride containing waste such as is the case with the method of the present invention. SUMMARY OF THE INVENTION The principal objective of the present invention is to provide a method and apparatus for processing solid wastes of the above-mentioned type, which method results in a "dead" (to use a biological term) pyrolysis residue and thereby an effective reduction in the volume of the waste. Another objective of the invention is to provide a method and apparatus which, in addition to the above-mentioned volume reduction, affords effective processing of the resulting exhaust gases. Another objective of the present invention is to provide a method and apparatus for the further reduction of the pyrolysis residue volume through the gasification of at least a portion of the carbon content of the pyrolysis residue. A further objective of the invention is to provide a method and apparatus which also affords an extremely high retention of the radioactivity present in the pyrolysis residue. A still further objective of the invention is to provide a method and apparatus which is straight forward in technical respects and which is therefore also cost effective taking everything into account as regards volume reduction of the solid waste and management of the resulting exhaust gases. The above-mentioned objectives are attained by pyrolyzing the solid waste at a relatively low temperature. The pyrolysis residue is passed through a steam reformer to gasify the carbon content of the solid residue. The resulting gases from the pyrolysis vessel and steam reformer vessel are mixed and oxidized (combusted) in a submerged bed heater. Residual acid gases and particulates are removed from the offgas stream by a fiber bed scrubber. The method and apparatus of the present invention can process sulfur and chloride containing waste having high organic content, including salt depleted resins, steam generator cleaning solutions, antifreeze and sludges. |
abstract | A method for measuring efficiency improvement in a heating system. The method includes the following steps: temperature sensors sense an outflow temperature. A controller records the temperature and a call time at a thermostat burner call. Software calculates a percentage of the reduction of fuel consumption, that the efficiency improvement saves. A display displays the percentage saved. |
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051456370 | claims | 1. A maintenance procedure for a nuclear reactor, said reactor including a vessel having a top, a bottom, and a wall, said reactor having a core within said vessel, said reactor having an instrumentation guide assembly, said instrumentation guide assembly including an incore housing extending through said bottom, said incore housing being bonded to said vessel by a weld, said instrumentation guide assembly also including a guide tube above, coaxial to and bonded to said incore housing, said guide tube extending at least partially through said core, said reactor having an instrumentation module which can be inserted into said core through said guide tube for monitoring said core, said procedure comprising the steps of: shutting down said reactor; removing said top; removing said instrumentation module from said vessel; inserting an ultrasonic probe and an attached shaft from above so that said shaft extends through said core and is at least partially immersed in water and through said guide tube and so that said probe extends at least partially into said incore housing; mounting a drive unit above said core, said drive unit being coupled to said probe via said shaft; operating said drive unit so that it moves said ultrasonic probe vertically and circumferentially in alternation within said incore housing while said ultrasonic probe is activated to detect defects in said weld; removing said probe from said vessel; inserting a new instrumentation module into said guide tube; replacing said top; and restarting the reactor. lowering a probe into the interior of said incore housing from above, said probe being mechanically coupled to a scanning drive mechanism via a shaft extending through said core; clamping said scanning drive mechanism to a top guide at the top of a reactor core of said reactor pressure vessel so as to define a reference circumferential position of said probe within and relative to said reactor pressure vessel, and adjusting the vertical distance between said drive mechanism and said probe so as to define a reference vertical position of said probe within and relative to said reactor pressure vessel; and raster scanning said weld by alternating sweeps in one of the vertical and circumferential dimensions with incremental movements in the other of said dimensions, and while raster scanning said probe includes multiple transducers and multiple eddy current coils all of said multiple transducers are pulsed during the entire ultrasonic examination; a plurality of said multiple transducers have the same focal point during the ultrasonic examination to interrogate the same region simultaneously; and all of said said multiple eddy current coils are energized during the entire eddy current examination. wherein said drive mechanism moves said probe to perform said examination with said multiple eddy current coils and with said multiple transducers so as to cause all of said transducers and all of said eddy current coils to travel from above the highest level of said weld to below the lowest level of said weld during their respective examinations. a probe with at least one ultrasonic transducer; a drive mechanism including electronic means for causing said drive mechanism to alternatively sweep said probe in one of a vertical dimension and a circumferential dimension and step in the other of said dimensions, said electronic means activating and taking data from said ultrasonic transducer, said electronic means coordinating vertical and circumferential movement of said probe with said data collection so the position of any detected defects can be specified. 2. A method of performing a non-destructive examination of a weld attaching an incore housing to a bottom head of a reactor pressure vessel, wherein said weld at least partially surrounds the external periphery of said incore housing, said method comprising the steps of: 3. A method as recited in claim 2 further comprising examining the interior surface and near surface thickness of said incore housing using at least one eddy current coil with said probe. 4. A method as recited in claim 3 wherein: 5. A method as recited in claim 4 wherein said incore housing at least partially penetrates said bottom head; and 6. A method as recited in claim 5 wherein said drive mechanism moves said probe so that all of said transducers and all of said eddy current coils travel from at least 40 millimeters above the highest level of said weld to at least 40 millimeters below the lowest level of said weld. 7. A method as recited in claim 6 wherein said examinations are done with vertical sweeps followed by rotations of about 5.degree. until said probe has rotated at least 360.degree.. 8. A method as recited in claim 7 wherein said probe includes at least one longitudinal wave transducer that faces normal to the surface of said incore housing, said respective examinations beginning with said longitudinal transducer facing so as to traverse past the highest level of said weld in the first vertical sweep of said ultrasonic examination. 9. A system for the non-destructive examination of welds between an incore housing and a bottom head of a reactor pressure vessel of a boiling-water reactor, said system comprising: 10. A system as recited in claim 9 wherein said vertical adjustment means includes an extension tube coupling said drive shaft and said probe, said adjustment means also including extension locking means for holding said extension tube in a fixed vertical position, said extension being free to rotate 360.degree. and slide lengthwise through its entire length in said shaft prior to locking. 11. A system as recited in claim 10 wherein said probe includes an eddy current coil. 12. A system as recited in claim 11 wherein said probe includes multiple ultrasonic transducers having the same focal point, said probe also including multiple eddy current coils. 13. A system as recited in claim 12 wherein said multiple eddy current coils include coils of both the absolute and differential type. 14. A system as recited in claim 13 wherein said probe comprises six ultrasonic transducers and four eddy current coils. 15. The system of claim 14 wherein, of said six ultrasonic transducers, three are oriented so their beams are perpendicular to the length of said probe. 16. A system as recited in claim 15 wherein a hoist cable capable of raising and lowering said drive mechanism, said extension, and said probe is attached to said drive mechanism. |
045044392 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention concerns a gas cooled nuclear reactor with several loops connected with the core by means of channels, with heat absorbing components such as heat exchangers, split tubular furnaces or the like being arranged in the loops with hot gas flowing through said components from bottom to top and their external surfaces being exposed to cold gas. 2. Background of the Prior Art In nuclear reactor installations, the thermal energy produced in the reactor core is transferred by means of a cooling gas medium to heat absorbing components such as heat exchangers, split tubular furnaces, and the like, which may be arranged in circuits or loops. A circuit or loop consists of a shaft or pod in the form of a cylinder arranged vertically in the prestressed concrete pressure vessel connected with the core by means of one or several channels. The circuit or loop houses different reactor components such as the heat exchanger equipment and the like. The cooling medium for example (helium) flows through the core from top to bottom. The cooling medium is conducted through a hot gas channel to the heat absorbing components, wherein it discharges a substantial portion of its thermal energy. It is then passed to the blower and returned to the core, whereby the cooled gas usually flows along the outside of the heat absorbing components. The above-described gas circuit may be supplemented by means of further reactor components, for example, a gas turbine. A plurality of circuits or loops are utilized in the conventional high temperature, gas cooled nuclear reactor. An essential characteristic of a nuclear reactor installation is that it is designed for a predetermined operational state. Excessive temperatures over extended periods of time or extreme fluctuations around its mean value may result in damage to the reactor installation. Grave problems are associated with the failure of certain heat absorbing components (such as with split tubular furnaces), as access to them is not possible for operational reasons alone (operational at gas temperatures of around 900.degree. C.). SUMMARY OF THE INVENTION It is the object of the invention to vary the gas temperature in a circuit of a gas cooled nuclear reactor installation without affecting the temperature of the gas in the other circuits. In an extreme case, it should be possible to take the affected circuit or loop out of service. This object is attained according to the invention by arranging a means for mixing gases under the heat absorbing component in the hot gas channel. The mixing means is connected with a cold gas conduit. The invention essentially consists of a mixing means arranged under a heat absorbing component located in the circuit or loop of a gas cooled nuclear reactor. Cold gas is fed into the hot gas flowing into the loop. The mixing means comprises at least one valve and a plurality of tubes projecting into the hot gas channel. The valves have a regulatory function, whereby the temperature of the gas being conducted to the heat absorbing component may be continuously varied within a temperature range of approximately 200.degree. to 900.degree. C. The simplest embodiment of the mixing means would have the potential disadvantage that the gas would not be adequately intermixed. A substantial temperature gradient within the mixed gas would possibly be created. Such a temperature gradient could have a negative effect (deformation) on the heat absorbing component. An adequate mixing of the hot gas with the cold gas may be obtained by feeding the cold gas by means of distributor heads into the hot gas. The distributor head has a plurality of orifices. The principal axis of each orifice is arranged perpendicularly to the surface of the distributor head. The turbulent flow of the gas flowing through the loop is prevented by inserting an apertured plate at the point of diversion in the hot gas channel. The tubes are arranged in the partition between the hot gas channel and the cold gas channel and may be located according to the invention either above or under the apertured plate. One embodiment is applied advantageously in cooling systems, wherein a reduction of the temperature from approximately 900.degree. to approximately 200.degree. is required. The tubes thereby pass through the apertured plate. The distributor heads are arranged above the apertured plate. In a further possible embodiment the tubes and the distributor heads are arranged under the apertured plate. A very good intermixing of the cold gas with the hot gas may be obtained in keeping with the invention, by causing the gas emerging from the mixing device to flow in a direction opposite to the direction of flow of the hot gas. Gas mixed in this manner displays no appreciable temperature gradient. The cold gas flows according to the invention through valves controlled by means of a regulator through the tubes into the hot gas channel. During normal operation, the actual temperature coincides with the nominal temperature and the valves are closed. If it is necessary to reduce the temperature of the gas flowing into the loop, the valves are opened slowly, whereby the cold gas flows into the hot gas. The heat absorbing components are, in a known manner, equipped with thermocouples to measure the temperature of the gas. In case of a failure of a heat absorbing component, it is desirable to reduce its temperature. This may be effected by opening the valves completely. By reducing the temperature, progressive failures of the heat absorbing components may additionally be prevented. The advantages obtained by the invention consist in particular of being able to vary the temperature of the gas flowing through the loop within a wide range of temperatures without thereby affecting the other loops, and particularly, without affecting the components arranged therein. |
054066000 | summary | TECHNICAL FIELD OF THE INVENTION The present invention relates to containers used for transportation and short term storage of spent nuclear fuel. BACKGROUND OF THE INVENTION As the nuclear utility industry matures, there is an ever-increasing need for additional storage space to safely contain spent nuclear fuel. One method that has been developed in recent years for storage of spent nuclear fuels is dry storage in horizontal storage modules, which are shielded bunkers in which containerized spent fuel is stored and monitored for definite periods of time. One conventional technique for horizontal modular dry storage of spent nuclear fuel rods is disclosed in U.S. Pat. No. 4,780,269 to Fischer et al. A basic procedure for dry storage of spent nuclear fuel is to position a dry shielded canister into a shielded transfer cask. The canister and cask are filled with deionized water, which is then lowered into a pool containing the spent nuclear fuel. Spent fuel assemblies are then placed into the canister, and a shielded end plug is positioned to close the canister. The canister and cask are then removed from the pool, and the cask and canister are drained and dried. The exterior of the cask is decontaminated, followed by closure of the cask with a closure plate. The closed transportation cask is then lowered onto a transport trailer and secured by tie-downs. The transport trailer carries the cask to the sight of the horizontal dry storage modules. The cask is opened and docked with an entry port of a dry storage module. The canister is then transferred from the cask into the module, such as by passing a ram through the dry storage module from an end opposite the entry port, through the entry port and into the opened cask. The canister can then be grasped and pulled into the dry storage module, after which both the entry port and access port are sealed. A critical aspect of this process is the safe containment and transfer of the spent nuclear fuel within the canister from the original pool storage to the final dry horizontal storage site. The transport cask must be constructed with adequate structural strength and shielding to both physically protect the dry shielded canister within, and to provide biological shielding to minimize personnel radiation dosages during canister transfer and transport operations. During the canister transfer and transport process, the cask must be able to withstand any foreseeable impact, such as could occur by accidental dropping of the cask from the transport trailer or exposure to tornadoes or other natural disasters. In the United States, federal regulations setting forth requirements that transport casks must meet are found in 10 C.F.R. 72, including subpart G, as well as 10 C.F.R. 71 and 10 C.F.R. 50. In particular, the cask must be able to withstand impacts due to a drop of 30' onto an essentially, unyielding fiat horizontal surface, without structural failure. Even if structurally damaged, no leakage of the contents from the cask is permitted. It is thus important to design casks with high structural integrity. At the same time, it is desirable to maximize the quantity of spent fuel that can be transported within the cask at any given time, and to minimize the cost of constructing the cask. While strength considerations typically warrant constructing the cask from thicker sections of metal and other materials, this requirement may reduce the quantity of spent fuel that can be transported within the cask. External dimensions of the cask are limited by constraints such as the total weight of the loaded cask, and clearances required to transport the casks through tunnels, under bridges and overpasses, and the like. Currently, conventional casks are often constructed from a polished austentitic stainless steel, such as 304 stainless steel, for corrosion prevention. However, such stainless steel is limited in strength and may fail under high stresses. To combat this potential, conventional casks are constructed from thick metal sections, and must be reinforced with gusset plates and other reinforcing members. Additionally, locations on the casks that are subjected to force during transport must be reinforced with additional metal plates welded to the cask structure. For example, conventional casks are outfitted with cylindrical trunnions welded or bolted directly to a cylindrical structural shell of the cask at diametrically opposed locations. These trunnions are grasped by hooks, and serve as pivot points while lifting the cask during the transportation process. Because of the stresses transferred to the cask structure from the trunnions during use, the shell is typically reinforced in the area surrounding the trunnions by welding additional plates of metal. The trunnions themselves are conventionally permanently secured to the structural shell of casks by welding or bolting directly to the shell. In the case of welding, the welded joint is subjected to substantial stress during hoisting of the cask. In the case of boring the trunnions in place, the bolts are subjected to extreme shear and tensile loads during hoisting of the cask. Again, the trunnions must be heavily reinforced to withstand such loads, increasing the weight and overall dimensions of the cask, and thus decreasing the spent fuel containment capacity and increasing the cost of manufacture. When sealed joints, such as elastomeric (e.g., O-ring) seals or metal seals are utilized, the base metal used to form the structural shell is conventionally machined to form the sealing surfaces. Thus, for example, when 304 stainless steel is used to construct the shell, annular surfaces on the shell are machined and polished to form sealing surfaces. While functioning adequately in most situations, extreme impact to the seal area, such as by accidental dropping of the cask at an oblique angle whereby force is concentrated on the seal area, may result in permanent deformation of the metal seal surface, and subsequent leakage potential. SUMMARY OF THE INVENTION The present invention provides a container designed for use as a cask for short-term containment and transporting of spent nuclear fuel. In the first aspect of the present invention, the container is formed from a structural shell defining a cavity for receiving spent nuclear fuel, and first and second end apertures opening into the cavity. The shell has a first end portion formed of a first material and a second end portion formed of a second material. The first end portion is joined to the second end portion to form the structural shell. A bearing surface is defined on the first end portion of the shell and is engageable to enable hoisting of the container. The first end portion of the shell is constructed from a first material that has a higher load bearing strength than the second material, to handle the hoisting stress. The container also includes a first closure securable to the first end portion of the shell to seal the first end aperture, and a second closure securable to the second end portion of the shell to seal the second end aperture. The container further includes a radiation absorbing shield layer, which may include both gamma radiation and a neutron radiation absorbing materials. The container is thus constructed so that those areas of the container that are subjected to the greatest stress, e.g. the first end portion, is constructed from the strongest material, such as a high-strength metal alloy. However, those portions of the cask that are not exposed to as high a stress are produced from lower cost materials having a strength that is adequate for the lower loads to be imposed on those portions. In a further aspect of the present invention, a cask is provided that includes a tubular inner shell defining a cavity for receiving spent nuclear fuel, and first and second ends. A tubular outer shell having first and second ends is assembled coaxially over the inner shell to define an annular space therebetween. A radiation absorbing material fills the annular space. An annular member defining a central aperture and a first annular sealing surface is secured about its perimeter to the first ends of the inner shell and the outer shell to create airtight joints with both the inner shell and the outer shell. A first closure plate is releasably securable to the annular member and defines a second annular sealing surface corresponding to the first annular sealing surface defined by the annular member. A seal is positioned between the second annular sealing surface of the first closure plate and the first annular sealing surface of the annular member to create an airtight seal between the first closure plate and the annular member. The cask also includes a second closure plate secured proximate its perimeter to the second ends of the inner shell to create airtight joints with the inner shell. In a further aspect of the present invention, a cask is provided that includes a structural shell defining a cavity for receiving spent nuclear fuel and first and second end apertures. A first closure is securable to the shell to seal the first end aperture. A second closure is securable to the shell to seal the second end aperture. A radiation absorbing shield layer is affixed to the shell. First and second pairs of trunnion mounting structures, preferably configured as tubular sleeves are secured in opposing disposition within apertures formed in the structural shell. First and second trunnions, each defining a base and a beating surface, are included. The base of each trunnion is releasably securable to a corresponding one of the trunnion mounting structures, whereby the bearing surfaces of the first trunnions can be grasped to hoist the container. The second trunnions are used to provide a point of support and rotation for loading and unloading the cask from its conveyance. In a preferred embodiment, the trunnion mounting structures are configured as annular sleeves that are welded to the structural shell of the cask, within which sleeves the base of the trunnions are received. Because of this construction, fasteners such as bolts used to secure the trunnions to the mounting structures are substantially isolated from tensile and shear loads. In a further aspect of the present invention, the trunnion mounting structures are preferably formed from a high-strength material such as is used to form the portion of the outer shell to which the first trunnions are mounted, thereby providing a strong trunnion mounting without requiring additional plate reinforcement. In a still further aspect of the present invention, improved seal joints are included in the cask. Sealing surfaces of the cask are formed utilizing hardened metal weld overlays, thereby providing sealing surfaces that are not readily subject to permanent deformation upon impact of the cask. In the preferred embodiment, sealing surfaces of closure plates on the cask include grooves formed to define a half-dovetailed cross section for receiving seals. This enables use of either metal or elastomeric seals in the joints, and enables assembly of the joints while the cask is in either the horizontal or vertical disposition. In a still further aspect of the present invention, a cask is disclosed that includes a tubular structural shell defining a cavity for receiving spent nuclear fuel and first and second opened ends. The first closure plate is releasably securable to the first opened end of the shell, whereby when secured to the shell, the first opened end of the shell is sealed, and when released from the shell, loading and unloading of spent nuclear fuel through the first open end into the cavity is permitted. The second closure plate is secured to and seals the second open end of the shell. The second closure plate defines a central access aperture. An access cover plate is releasably securable to the second closure plate to seal the central access aperture. When released from the second closure plate, entry of a ram through the access aperture into the cavity of the shell to facilitate unloading of spent nuclear fuel through the first open end of the shell is permitted. Shield plugs filled with a radiation-absorbing material are provided to cover the trunnion mounting structures and central access aperture formed in the cask during short-term storage and transportation. In another aspect, the present invention relates to a skid for transporting a nuclear fuel transportation cask and containment vessel. The skid supports the cask around the neutron radiation shielding material. The skid includes a supporting member and a retaining member that each include a plurality of parallel spaced-apart plates lying in planes perpendicular to a longitudinal axis of the cask which are connected by a plurality of longitudinal fins parallel to the longitudinal axis of the cask. The longitudinal fins are positioned to mate with structural elements associated with the neutron radiation shielding material to transfer loads from the cask to the skid. The present invention thus provides a cask that is less costly to construct, yet that provides improved safety under impact conditions. Exposure of workers to radiation during transport procedures is also reduced. |
summary | ||
052290654 | description | DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows a prior art circuit 1 for sampling, and for measuring the temperature of, the primary coolant fluid of a pressurized-water nuclear reactor. The sampling and measuring circuit comprises small-diameter pipes 2 which are connected to a hot leg 3, to a cold leg 4 and to an intermediate leg 5, termed cross-over leg, of a loop of the primary circuit of the reactor. The intermediate leg of a pressurized-water nuclear reactor connects the outlet of the primary part of the steam generator to a primary pump for circulating the coolant fluid in the loop, the delivery part of which is connected to the cold leg 4. The pipes 2 of the sampling and measuring circuit are connected to the legs 3, 4 and 5 via devices for sampling the coolant fluid from, or for reintroducing the coolant fluid into the primary circuit, traversing the wall of the corresponding primary pipe and generally referred to as scoops. The pipes 2 of the sampling circuit 1 are connected to the hot leg via three scoops 6 distributed around the pipe constituting the hot leg 3. The ducts 2 are also connected to the cold leg 4 and to the intermediate leg 5 via scoops 7 and 8 respectively. The sampling and measuring circuit 1 also comprises couplings 10, gates and flaps 11, headers 12 and measuring probes 13. The sampling and measuring circuit 1 has a complex structure and the total length of the ducts 2, in the case of a conventional pressurized-water nuclear reactor, can be of the order of 60 m. As a result, there are risks of leakage and of radioactive contamination of the environment and of the operatives charged with the maintenance of the circuit. A measuring device 15 according to the invention can be seen in FIG. 2, arranged on a part of a hot leg 16 of the primary circuit of a pressurized-water nuclear reactor, which hot leg has a substantially horizontal axis 17. As can be seen in FIGS. 2 and 3, the measuring device 15 comprises three scoops 19, 20, 21 for sampling coolant fluid and an element 22 for reintroducing the sampled coolant fluid into the primary pipe 16. The sampling scoops 19, 20 and 21 are spaced by 120.degree. about the axis 17 of the pipe 16, in a substantially vertical, straight section of this pipe. The scoops 20 and 21 are arranged beneath the axis 17 of the pipe 16 and the scoop 19, at the upper part of the straight section of the pipe 16. The scoops 19, 20 and 21 are equipped with temperature-measuring probes, as will be explained hereinbelow. The reintroduction device 22 is placed in a position diametrically opposite to the scoop 21 situated beneath the axis of the pipe 16 and in the vicinity of its lower part. Each of the scoops, 19, 20, 21 is connected to the reintroduction element 22 via a small-diameter pipe, 24, 25 and 26, respectively. The water sampled at the scoops 19, 20 and 21 circulates outside the duct 16 in the pipes 24, 25 and 26 and is reintroduced into the pipe 16 by the element 22, as indicated by the arrows 27 shown in FIG. 3. The whole of the device 15 having the shape of a portion of a ring is arranged around the pipe 16. A scoop such as 19 can be seen in FIG. 4, fixed in an opening 28 traversing the wall of the pipe 16. The scoop 19, in the shape of a glove finger, is fixed by a weld seam 30 in the wall of the pipe 16. The portion of scoop 19 situated inside the pipe 16 is pierced by openings 31 opening out into the axial central channel 32. A projection 34 is fixed to the outer surface of the pipe 16 by a weld seam 35 in such a way that the internal bore 36 of the projection 34 is in the extension of the bore 32 of the scoop 19. The bore of the projection 34 and the upper part of the bore 32 are machined so as to receive the support 38 for a temperature-measuring probe 39. The upper part 36a of the bore 36 is tapped so as to receive a threaded part of the support 38 which is screwed into the tapped part 36a of the bore 36. A weld seam 40 makes it possible to ensure a sealed joint between the support 38 and the projection 34. The projection 34 is also pierced in order to form a channel 41 opening out into the bore 36, and a seating receiving the end of the pipe 24 of the device 15 which is fixed by a weld seam 43 to the projection 34 and which is brought into communication with the bore 36 by the channel 41. The scoops 20 and 21 are formed in the same manner as the scoop 19 and comprise a part in the shape of a glove finger traversing the wall of the duct 16, and a connecting projection to which are fixed respectively the pipe 25 and the pipe 26. The element 22 for reintroducing the coolant water into the primary pipe 16 constituting the hot leg can be seen in FIG. 5. The wall of the pipe 16 is traversed by an opening 45, and the element 22 for reintroducing the coolant water consists of a projection 46 fixed by a weld seam 47 to the outer surface of the pipe 16 and comprising a bore 48 arranged in the extension of the bore 45 traversing the wall of the pipe 16. The bore 48 is machined to a diameter greater than the diameter 45 and comprises a tapped part 48a so as to be able to receive a probe support 50 arranged over the entire length of the bore 48 and comprising an end part of reduced diameter, on which is fixed a temperature probe 51 projecting slightly into the opening 45. The probe support 50 comprises a threaded end part which is screwed into the tapped opening 48a, and two frustoconical parts joining its threaded end part to its end of reduced diameter. The support 50 is engaged, by screwing, into the bore 48 and fixed in a sealing fashion on the end of the projection 46 by a weld seam. The sealed joint between the probe support 50 and the projection 46 could also be provided by a seal. The projection 46 is also traversed by three channels arranged at substantially 120.degree. about its axis, such as the channel 52 opening out into the bore 48, in a space provided in this bore, at the periphery of a frustoconical part of the support 50. The end of a pipe such as 24 or 25 or 26, joining one of the scoops to the elements 22 for reintroducing the coolant water into the primary duct, is fixed in the region of each of the channels 52. Each of the pipes such as 24 is fixed in a sealing fashion on the projection 46 by a weld seam such as 54. The probe support 50 could, optionally, be replaced by a plug closing the part of the bore 48 situated above the channel 52, fixed onto the projection 46 by screwing and by welding. The temperature probes such as 39 arranged in the scoops 19, 20 and 21, and the temperature probe 51 associated with the projection for reintroducing pressurized water 46, are connected via conducting wires traversing the corresponding probe bodies to a module for processing the signals from the probes, arranged in the control room of the reactor. The signals emitted by the probes are representative of the temperature of the water sampled at the scoops 19, 20 and 21, or of the temperature of the coolant water reintroduced into the primary pipe 16 at the reintroduction element 22, and are collected by the processing module which enables values representative of the temperature of the coolant water of the primary circuit to be obtained. The mean of the three values obtained from the probe 39 is calculated, which makes it possible to obtain and display a value for the temperature of the coolant water in the hot leg, eliminating some of the effects of stratification of the coolant water in the hot leg which are likely to give rise to errors and disturbances in the measurements. The mean temperature obtained by the electronic calculation means of the module can be compared with the temperature obtained from the probe 51 which corresponds to the actual mean temperature of the fluid reintroduced into the primary pipe 16 via the pipes 24, 25 and 26, and via the projection 46. Indeed, a certain mixing of the quantities of water sampled at each of the scoops 19, 20 and 21, takes place in the bore 48 of the projection 46. The temperature measured by the probe 51 therefore corresponds to the mean temperature of the coolant water reintroduced into the duct 16. This comparison makes it possible, to detect any operational anomaly of the device 15 where a substantial discrepancy exists between the two mean values obtained. Flowmeters such as the flowmeter 29 arranged on each of the ducts 24, 25 and 26 make it possible to check that the fluid is circulating correctly in the ducts of the measuring device 15. The device 15 therefore makes it possible to effect a certain mixing and homogenization of the coolant fluid so as to obtain a value representative of the temperature of the primary fluid. This mixing and homogenization are necessary to the extent that substantial stratification of the coolant fluid occurs in the hot leg 16 of the primary circuit. In the case of a cold leg or of an intermediate leg such as the leg 55 shown in FIG. 3a, the measurements of the temperature of the coolant water of the primary circuit can be carried out using a single temperature probe arranged in a scoop 56 having openings for the passage of the coolant water in its part arranged inside the pipe 55, or possibly a plurality of aligned probes at the upper part of the pipe. The values supplied by this probe or these probes are also processed, displayed and taken into account in the control room of the reactor. In the case of the cold and the intermediate leg no substantial stratification of the fluid occurs, so that measurement of the temperature in the upper part of the pipe is sufficient to obtain a representative value. The probe 51 of the element 22 for reintroducing the coolant water can be used as an emergency probe if one of the probes associated with a scoop 19, 20 or 21 measuring the outlet temperature of the coolant fluid is faulty. If the projection 46 of the element 22 for reintroducing the coolant water is closed by a plug, the element 22 then not having a temperature probe, the circuit consisting of the pipes 24, 25 and 26 serves solely to ensure satisfactory mixing of the fluid in the region of the three probes arranged in the scoops 19, 20 and 21, on the outlet of the coolant water. In all cases, the device according to the invention makes it possible to measure, in a simple manner, the temperature of the coolant water of the reactor in the vicinity of the primary pipes, using a small-sized device with reduced risks of a coolant leakage and of contamination of the environment of the reactor and of the maintenance staff. The structure and arrangement of the scoops equipped with temperature probes, and of the element for reintroducing the fluid into the primary pipe may be different from that which has been described and illustrated. The measurement signals emitted by the probes arranged on the hot leg, on the cold leg or on the intermediate leg can be processed in any manner known from the prior art and displayed, or cause alarms to be triggered. The invention is applied not only to pressurized-water nuclear reactors but also to any nuclear reactor comprising a primary coolant fluid circulating in large-diameter pipes, at least certain parts of which have a substantially horizontal arrangement. |
claims | 1. A drop system for cooling a nuclear reactor, the drop system comprising:a. a heat exchanger having a first side and a second side, wherein the heat exchanger is within a nuclear reactor, the heat exchanger comprising:i. an inner pipe having an expandable lip, wherein one or more gaskets are circumferencially engaged with an outer circumference of the inner pipe;ii. an outer pipe, wherein the inner pipe and the outer pipe are nested, wherein a first end of the nested inner and outer pipes is in communication with the first side of the heat exchanger, wherein a second end of the nested inner and outer piper is in communication with the second side of the heat exchanger, wherein the inner pipe slidingly engages an interior surface of the outer pipe;iii. one or more burst discs, wherein the one or more burst discs are configured to rupture;b. one or more fusible links attached to the outer pipe, wherein the fusible links are collapsible, wherein the one or more fusible links are heat activated;c. at least one locking mechanism having a counter weight,wherein the at least one locking mechanism is in communication with the outer pipe,wherein the locking mechanism raises and lowers the outer pipe, and wherein liquid nitrogen absorbs heat as it flows through the heat exchanger, and wherein the liquid nitrogen flows through the heat exchanger after the one or more burst discs rupture. 2. The system of claim 1, further comprising a gas-powered generating unit, for generating electricity from the liquid nitrogen gas as it expands. 3. The system of claim 2, further comprising a hydraulic system powered by the gas-powered generating unit. 4. The system of claim 3, wherein hydraulic system opens and shuts valves. 5. The system of claim 1, comprising a relief valve to relieve excess gaseous nitrogen pressure from the system. 6. The system of claim 5, further comprising an expansion tank, wherein the relief valve evacuates the excess gaseous nitrogen pressure to the expansion tank. 7. The system of claim 1, wherein the fusible links collapse at a predetermined temperature, and wherein the collapsing of the links cause the heat exchanger to automatically drop into a position at a lowered state relative to an elevated state. 8. The system of claim 1, wherein the several pipes communicate the liquid nitrogen through the first side of the heat exchanger. 9. The system of claim 1, wherein the liquid nitrogen comprises compressed atmospheric nitrogen. 10. The system of claim 1, further comprising a hydraulic system in communication with the heat exchanger. |
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047042354 | abstract | A method of decontaminating radionuclide-contaminated acid insoluble corrosion products from primary system surfaces in pressurized water reactors by oxidation and concurrent dissolution in an acidic decontamination solution of the corrosion products which have been made acid-soluble by the oxidation. The characterizing feature of the method is that the oxidation is carried out at relatively low temperatures with a water-based oxidation agent having a pH below 7 and containing cerium nitrate, chromic acid and ozone. |
047284838 | summary | CROSS REFERENCE TO RELATED APPLICATION Reference is hereby made to the following copending application dealing with related subject matter and assigned to the assignee of the present invention: "Position Sensing Apparatus" by George S. Jewell, assigned U.S. Ser. No. 678,520 and filed Dec. 5, 1984 now U.S. Pat. No. 4,583,297. BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to fuel assemblies for nuclear reactors and, more particularly, is concerned with an apparatus used in an automated system for inspecting a fuel assembly for envelope, channel spacing and length and also for correcting error in the inspection fixture of the apparatus. 2. Description of the Prior Art In most nuclear reactors, the reactor core is comprised of a large number of elongated fuel assemblies. Conventional designs of these fuel assemblies include a multiplicity of fuel rods held in an organized array by grids spaced along the fuel assembly length. The grids are attached to a plurality of control rod guide thimbles. Top and bottom nozzles on opposite ends of the fuel assembly are secured to the guide thimbles which extend above and below the opposite ends of the fuel rods. The fuel rods which contain fissile material are grouped together in a closely-spaced array within each fuel assembly and the fuel assemblies, in turn, are mounted in side-by-side closely-spaced relationship with one another between the upper and lower core plates so as to provide a neutron flux in the core sufficient to support a high rate of nuclear fission and thus the release of a large amount of energy in the form of heat. In view of the densely-packed condition of the fuel rods and fuel assemblies in the core, the dimensional standards of envelope and length of each fuel assembly and the channel spacing between the adjacent fuel rods of each fuel assembly must be met within very close tolerances. Thus, at the completion of manufacture of each fuel assembly, quality control inspections are carried out to determine whether the fuel assembly meets the aforementioned dimensional standards. Currently, the fuel assembly quality inspection is performed at three separate stations: (1) envelope measurement; (2) channel spacing; and (3) length measurement. At the envelope measurement station, the out-of-straightness of the fuel assembly is quantified. The sides of a fuel assembly are normally not perfectly straight. The fuel assembly commonly exhibits a slight bow and twisting. Quantifying this behavior is performed by measuring the relative position of the grids to each other and inspecting for excessive displacements. The current method of envelope measurement uses twelve LVDT sensors mounted in a configuration of three sensors per side. A set of distance measurements is taken at each grid location to signify whether the grid is located either left/right or back/front of the center of the fuel assembly. At the channel spacing station, the distance between adjacent fuel rods within a fuel assembly is checked. Currently, an operator manually pulls a strain gauge probe through a channel and a computer translates the sensor output into distance measurements. At the length measurement station, the fuel assembly length is measured using a stick micrometer. The fuel assembly is set upright on a level table and an inspector measures the distance from the table surface to the bottom edge of the top nozzle. From the foregoing brief description of current practices, it will be readily understood that these stations are manual in nature, requiring an inspector to monitor equipment and process data. Consequently, a need has emerged to improve and automate the way in which fuel assembly inspection is carried out. SUMMARY OF THE INVENTION The present invention provides an apparatus for use in an integrated fuel assembly inspection system designed to satisfy the aforementioned needs. In contrast to the previous practices, the fuel assembly inspection apparatus of the present invention allows a fully integrated system wherein all inspections are performed in one station in a completely automated manner and all measurements are made using non-contact sensing techniques. Also, for increased accuracy of fuel assembly envelope measurement, the apparatus facilitates performance of correction for inspection fixture error on a real-time basis instead of on a sampled basis. Accordingly, the present invention is directed to fuel assembly inspection apparatus, comprising: (a) an elongated fixture mounted in a stationary upright position; (b) upper means mounted to an upper portion of the fixture and lower means mounted adjacent to a lower portion of the fixture for supporting a nuclear fuel assembly therebetween and extending along the fixture; (c) a bottom carriage having a central opening adapted to receive the fuel assembly therethrough when supported between the upper means and the lower means such that the bottom carriage will surround all sides of the fuel assembly, the bottom carriage being mounted to the fixture for generally vertical movement along the fixture and the fuel assembly; and (e) drive means for selectively moving the bottom carriage. In addition, means are disposed on the bottom carriage for measuring the fuel assembly envelope when the bottom carriage is moved to and stationed at selected axial positions along the fuel assembly. The envelope measuring means includes a single-axis positioning table disposed on each side of the bottom carriage adjacent a side of the fuel assembly, a proximity sensor mounted on each positioning table for movement along the adjacent side of the fuel assembly, and power means coupled to each sensor for stationing the sensor at a home position while the bottom carriage is moving along the fuel assembly and for sweeping the sensor relative to the side of the fuel assembly away from and back to the home position once the carriage is positioned at one of the selected axial positions along the fuel assembly. Further, means are disposed on the upper means and the bottom carriage for continuously monitoring fixture out-of-straightness and performing correction of the envelope measurement in response thereto. The monitoring and correction performing means includes a pair of X-Y axes lasers mounted on one of the upper means and the bottom carriage adjacent the fuel assembly, and a pair of matched X-Y photodetectors mounted on the other of the upper means and the bottom carriage adjacent the fuel assembly. The respective lasers provide straight line reference used to excite the corresponding photodetectors. The pairs of lasers and photodetectors facilitate measuring of both translational and rotational motion of the bottom carriage as the same moves up along the fuel assembly for facilitating adjustment of the envelope measurment for any fixture error at each of the axial positions along the fuel assembly. Still further, means are disposed on the bottom carriage for measuring channel spacing between fuel rods of the fuel assembly. The channel spacing measuring means includes a single-axis positioning table located on each of a pair of adjacent sides of the fuel assembly, a capacitive probe mounted on each of the tables for movement along the side of the fuel assembly, and motive means for driving the probe to specified channels locations along the fuel assembly side for taking channel spacing measurements once the bottom carriage is positioned at one of the selected axial positions along the fuel assembly. Finally, means are disposed on the bottom carriage and the fixture for measuring fuel assembly length when the bottom carriage has been moved between the bottom and top nozzles of the fuel assembly. The fuel assembly length measuring means includes a photoswitch mounted on each side of the bottom carriage adjacent a side of the fuel assembly and operable to detect an edge of the respective bottom and top nozzles of the fuel assembly, and means forming an optical scale mounted on the fixture and the bottom carriage for determining the position of the carriage as it moves along the fuel assembly when each photoswitch detects the respective edges of the bottom and top nozzles for deriving the length of the fuel assembly. These and other advantages and attainments of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention. |
abstract | The present invention relates to a method and device for determining the energetic composition of electromagnetic waves. It is the object of the present invention to provide a method and device for X-ray spectroscopy that allows simultaneous detection of the individual energies at a comparatively higher resolution and/or across a comparatively wider energy range. According to the invention, at least one reflective zone plate (12) is used that comprises a multitude of predefined wavelength-selective regions (14) arranged next to one another, wherein the wavelength-selective regions (14) each include a multitude of reflecting arched portions (20), which extend exclusively and continuously across the respective wavelength-selective region (14). |
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H00004073 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The invention in one embodiment converts neutron energy (E.gtorsim.1 MeV) to electric current and the concomitant development of an electrical voltage, and it produces deep ultraviolet radiation at the same time. With reference to FIG. 1, consider a swarm of 10-14 MeV neutrons (n) incident upon one or two opposing sides of a closed hollow container 11. The container sidewall or sidewalls 13a and 13b upon which the neutrons are incident is each a thin, substantially planar metallic cathode, each of thickness h, and a substantially planar screen anode 15 is positioned parallel to, and approximately equidistance from, the two cathodes as shown. Each of the cathodes contains a low atomic weight element such as Li.sup.6 as a constituent. The energetic neutrons collide with the Li.sup.6 particles in the cathodes and produce ionized helium and tritium according to reactions such as EQU n+Li.sup.6 .fwdarw.He.sup. ++ +T.sup.+ +3e.sup.-. The thickness .DELTA.h of each cathode is chosen so that the mean free path of the He.sup.++ and the T.sup.+ thus produced is greater than .DELTA.h. The mean free path of an alpha particle in a metal of effective atomic charge Z is estimated to be EQU .lambda.(He.sup.++).apprxeq.10.sup.-1 /Z.sup.2 (cm) so that the thickness of each cathode will probably be required to be .DELTA.h<0.1 mm. A substantial fraction (>36 percent) of the energetic He.sup.++ and T.sup.+ ions enter che interior of the container 11 where they encounter a high pressure gas (P=1-30 atm.) 17, such as He, Ne or Ar. The He.sup.++ and the T.sup.+ ions that enter the container gas deposit energy in the gas and produce additional charged particles and excited states through "Coulomb drag", and a plasma is formed. The plasma radiates deep ultraviolet photons (h.nu.=10-20 eV), many of which strike the adjacent metal cathodes and eject photo-electrons of energy determined by EQU eV=h.nu.-.PHI..sub.c, h.nu.=representative photon energy.apprxeq.10-20 eV, PA1 .PHI..sub.c =cathode work function.apprxeq.1.6-6 eV. PA1 .rho.=(.eta..sub.e1 Q-1/.eta.laser) E.sub.L .sup..nu. prf, PA1 .eta..sub.e1 =electrical energy conversion efficiency (assumed=0.4 here), PA1 Q=fusion target energy gain (assumed=80 here), PA1 .eta..sub.laser =laser energy efficiency (assumed=0.05 here), PA1 E.sub.L =laser energy delivered to target (assumed=1 MJ here), PA1 .nu..sub.prf =target fusion pulse repetition frequency. PA1 .eta.=F.sub.f I.sub.sc V.sub.oc /P.sub.in, PA1 F.sub.F (.apprxeq.0.8)=photovoltaic system fill factor, PA1 I.sub.sc =short circuit current of system, PA1 V.sub.oc =open circuit voltage of system (.apprxeq.0.85 E.sub.gap), PA1 P.sub.in =power input to system, PA1 E.sub.gap =energy gap of suitable photovoltaic semiconductor (.apprxeq.1.45-5 eV). The voltage V corresponding to the ejected photo-electrons is further reduced by the product, IR.sub.plasma of the current flow I (away from the cathode and towards the screen anode as indicated in FIG. 1) and the plasma resistance R.sub.plasma to a resultant voltage EQU V.sub.r =V-JR.sub.plasma, specific, The plasma specific resistance value is expected to be quite low here (R.ltorsim.0.3 ohm-cm.sup.2) because of the presence of the plasma in the container gas volume. The product JR.sub.plasma is no more than 3 volts for J.ltorsim.10 amps/cm.sup.2, if the container space is neutralized by the plasma; and the resultant voltage developed is then V.sub.r =5-15 volts. If one positions an impedance or other electric load 19 between. and connects it to, the screen anode and a metal wall cathode, an electric voltage will be developed across the load between cathode and anode. The action is analogous to that of a gas-filled photodiode, and the presence of the plasma avoids space charge limitations on current flow from cathode to anode in the container gas. In order to stop or substantially decelerate a neutron with initial kinetic energy E.congruent.10 MeV, one would need an equivalent thickness of cathode-anode material of at least 25-50 cm of metallic material. Thus, one can concatenate the basic cathode/anode/gas container many times, using adjacent units as indicated in FIG. 2, and repeat the above-described sequence of events several hundred times substantially simultaneously. This improves the efficiency of conversion of the energy of the neutron swarm; and since the resulting voltages V.sub.r are additive the net resulting voltage from one end of the concatenated structure to the other can be of the order of 1 kV. The presence of He.sup.+ ions in the container gas volume can promote the following reactions: EQU He.sup.+ +2He.fwdarw.He.sub.2.sup.+ +He, EQU He.sub.2.sup.+ +e.sup.-.fwdarw.He.sup.* +He, EQU He.sup.* +2He.fwdarw.He.sub.2.sup.* +He, ##STR1## FIG. 3 schematically exhibits the energy levels of the excited monomers and dimers (excimers) of interest here, He.sup.* and He.sub.2 *. The He.sub.2 * excimer will preferentially dissociate to He+He in the presence of a third particle (such as He) with emission of UV radiation at a wavelength .lambda..sub.d.apprxeq. 840 .ANG. (E=14.7 eV); this represents about 60% of the internal energy binding the two He particles. Radiation of wavelength .lambda.=640 .ANG., which is also produced by the decay shown in FIG. 3, is radiation trapped and will probably not be emitted from the container. If neon or argon gas is used rather than helium gas in the container, the UV radiation would appear at a wavelength .lambda..sub.d .congruent.1100 .ANG. (E=11.3 eV) or .lambda..sub.d .congruent.1300 .ANG. (E=9.5 eV), respectively, with about the same conversion efficiency. Thus, with approximately 60% conversion efficiency, deep UV radiation (E=9.5-14.7 eV) can be produced and emitted from the container. Much of this radiation can be used to conduct in situ experiments within the container 11, if desired. A second embodiment of the invention also provides for generation of electricity, and the concomitant development of an electrical voltage, from fast neutron reactions. with reference to FIG. 4, a closed hollow sphere 21 is provided with a sufficiently thick shell 22 to withstand internal pressures of at least p=100 atmospheres, and the sphere interior is filled with a noble or other inert gas such as He or Ne at a pressure of substantially p=100 atm. The inner diameter of the sphere should be substantially d=20 M and should have one or more dedicated sectors 23 for delivering a sequence of fusion targets T (and, separately, two or more fusion laser beams h.nu..sub.L) to the geometric center of the sphere, where laser-induced fusion occurs. Target fusion produces a plurality of high energy neutrons (n) that move through and collide with the noble gas and produce sequences of reactions such as: EQU n(fast)+He.fwdarw.He.sup.+ (fast)+n+e.sup.-, EQU n(fast)+He.fwdarw.He.sup.++ (fast)+n+2e.sup.-, EQU He.sup.++ (fast)+He.fwdarw.He.sup.++ (fast)+He.sup.+ +e.sup.-, EQU He.sup.+(fast)+He.fwdarw.He.sup.+ (fast)+He.sup.+ +e.sup.-, EQU He.sup.++ (fast)+He.fwdarw.He.sup.++ (fast)+He*, EQU He*+2He.fwdarw.He.sub.2 *+He.fwdarw.3He+h.nu..sub.d (.lambda..sub.d .apprxeq.840 .ANG.), EQU He.sup.+ +2He.fwdarw.He.sub.2.sup.+ +He, EQU He.sub.2.sup.+ +e.sup.- .fwdarw.He.sub.2 *.fwdarw.2He+h.nu..sub.d (.lambda..apprxeq.840 .ANG.). With Ne substituted for He, radiation of wavelength .lambda..apprxeq.1100 .ANG. is produced from Ne.sub.2 * decay. Assuming a reaction cross-section of .sigma.(n, He)=10.sup.-24 cm.sup.2, the mean free path between collisions of an energetic neutron with He particles is .lambda.=1N.sigma.=370 cm so that substantially all neutron energy is absorbed in (n, He) collisions within the gas before a neutron reaches the wall of the sphere. As noted above, the He or He.sup.+ or He.sup.++ or He.sub.2.sup.+ gas particles have no electronic states that may be excited by radiation of wavelength .lambda.=840 .ANG. so that the He gas is substantially transparent to such radiation. Only modest Rayleigh scattering of the radiation occurs within the gas so that most of the radiation (at photon energies up to 14.5 eV) will ultimately reach the sphere walls. With reference to FIGS. 5 and 6, these photons will substantially all pass through a first layer of thin anode plates 25, positioned adjacent to and substantially parallel to the sphere walls and spaced apart therefrom in the sphere interior. The photons then strike (and mildly heat) a thick metal cathode laye 22 (the sphere wall) that is treated in bulk with Cs, Th, Ba oxide, Sr oxide or a similar suitable atomic material to reduce the work function of the metal cathode material to .PHI..sub.c .ltorsim.1.6 eV. Preferably, the material comprising the thin anode plates 25 has a work function .PHI..sub.a >>1.6 eV at the temperatures of operation. The photons scatter from atoms or molecules within the cathode material; and if the photon energy satisfies EQU h.nu.>E.sub.F +.PHI..sub.a, where E.sub.F is the Fermi level of the electrons in the cathode material, each such photon that scatters within the cathode material may produce one or more photo-electrons that escape from the cathode material and move to one or the other of the anodes 25, thus creating a potential difference across the cathode anode circuit (FIG. 5). With the electron work function of the anode material chosen to be >>1.6 eV, photo-electrons will be preferentially created in, and will preferentially exit from, the cathode material vis-a-vis the anode material so that electron flow from the sphere walls (cathode) to the anode plates will predominate. Each anode plate 25 is positioned inside the sphere so that the space 26 between each anode plate and the adjacent sphere wall 22 is also filled with the noble gas at p=100 atm pressure. The spacing d.sub.1 (=10-50 .mu.m) by insulating soacers 18 between anode plane and sphere wall in FIG. 5 is chosen small enough so that most photo-electrons emitted by the cathode will experience only a modest number of scatterings by noble gas particles in the space separating the anode and cathode and will ultimately reach and and be absorbed on one of the anode plates 25. An approximate relation for the power produced by a target fusion event is given by If one gigawatt of power is required here, the target fusion pulse repetition rate must be at least .nu..sub.prf =83 Hz, which is probably achievable with current technology. With reference to FIG. 5, the anode plates adjacent to the sphere wall can be fabricated from very thin plates of a light metal such as Al of transverse area 2-100 cm.sup.2 and with adjacent anode plates being spaced apart a small distance determined by the static voltage stand-off requirements for two such adjacent plates. The anode plates themselves may be made very thin (<100 .mu.m thickness) as the plates are not required to maintain any pressure differential across themselves. With reference to FIG. 6, one may replace the cathode-anode arrangement of FIG. 5 with a photovoltaic means 25' that is adjacent to but spaced apart from the sphere wall 22, with a sequence of diodes 29 electrically connecting the photovoltaic array means 25 and the sphere wall 22. The photovoltaic means 25' should preferably have a radial thickness of at least two mean free paths for absorption of the photons of characteristic energy h.nu..sub.d. As a photocurrent is generated in 25' and moves through the diode (optional) or other electrical load, a substantial electrical voltage may be generated between the photovoltaic means 25 and the sphere wall 22. Use of the photovoltaic array of FIG. 6 offers certain advantages over use of the cathode-anode array of FIG. 5. First, the photovoltaic means 25' is capable of repairing itself in the high gamma flux and high neutron flux environment by annealing higher photon flux produces a corresponding higher current, with no direct deleterious effects. Second, the current developed acros the diodes is not space charged-limited, whereas the current flowing from cathode to anode in FIG. 5 may be so limited. For steady state operation, the power conversion efficiency for the photovoltaic process is approximately To achieve the highest conversion efficiency available (20-50 percent) for the photon energies used, one might use a semiconductor material with as high a value of E.sub.gap as possible (e.g., diamond or ZnS with E.sub.gap .apprxeq.5.4 or 3.9 eV, respectively), consistent with other physical requirements. The foregoing description of preferred embodiments of the invention is presented for purposes of illustration only and is not intended to limit the invention to the precise form disclosed; modification and variation may be made without departing from what is regarded as the scope of the invention. |
048511560 | claims | 1. A nitric acid free method of processing combustible nuclear waste material so as to retain volatile radionuclides, said method consisting of the steps of: (1) heating and agitating said nuclear waste material with about 5 to about 12 liters of concentrated sulfuric acid per kilogram of said nuclear waste material at a temperature between about 250.degree. C. and about 330.degree. C. to form dispersed elemental carbon, in an environment which permits said elemental carbon to reduce said volatile radionuclides to nonvolatile forms; and (2) removing said acid from said nuclear waste material by heating at a temperature between about 350.degree. C. and about 450.degree. C., thereby evaporating said sulfuric acid. (1) agitating said waste material with about 5 to about 12 liters of sulfuric acid per kilogram of waste material, at a temperature between about 250.degree. C. and about 330.degree. C., whereby said waste is converted into particulate carbon plus an inert residue, and said carbon prevents volatile ruthenium compounds from forming and, if formed, reduces them to non-volatile compounds; (2) evaporating said sulfuric acid from said waste material by heating at a temperature between about 350.degree. C. and about 450.degree. C. (3) desulfating by heating at a temperature between about 700.degree. C. and about 900.degree. C. until sulfur dioxide is no longer evolved; (4) adding, at any previous step in said process, about 2 to about 20% by weight glass formers, based on total solids, including said glass formers; and (5) heating at a temperature between about 1000.degree. C. and about 1200.degree. C. to form a glass and to contain dispersed radioactive waste therein. (1) agitating said waste material with about 5 to about 12 liters of sulfuric acid per kilogram of waste material, at a temperature between about 25.degree. C. and about 330.degree. C., to form dispersed elemental carbon, whereby said waste is converted into particulate carbon plus an inert residue, and said carbon prevents volatile ruthenium compounds from forming and, if formed, reduces them to non-volatile compounds; (2) evaporating said acid from said waste material by heating at a temperature between about 350.degree. C. and about 450.degree. C.; (3) adding, at any previous step in said process, about 2 to about 20% by weight glass formers, based on total solids, including glass formers; and (4) pressing into a wafer; and (5) sintering said wafer to form a ceramic which contains dispersed radioactive waste therein. 2. A method according to claim 1 including the additional steps of adding about 2 to about 20% borosilicate type glass forming compounds to said waste material to form a mixture, desulfating said carbonized waste materials and fusing said mixture into a glass. 3. A method according to claim 2 wherein said desulfating is performed at 700.degree. to 900.degree. C. until sulfur dioxide is no longer evolved. 4. A method according to claim 2 wherein said glass is formed at about 1000.degree. to about 1200.degree. C. 5. A method according to claim 1 including the additional steps of adding borosilicate glass forming compounds to said waste material and sintering said glass forming compounds to form a ceramic. 6. A nitric acid free method of immobilizing radioruthenium-containing combustible nuclear waste material in glass so as to retain radioruthenium and dispersed radioactive residues within said glass, said method consisting of the steps of 7. A method according to claim 6 wherein said glass is formed in containers used for immobilization. 8. A method according to claim 6 wherein said glass formers are borosilicate type. 9. A nitric acid free method of immobilizing radioruthenium-containing combustible nuclear waste material in ceramic so as to retain radioruthenium and dispersed radioactive residues within said ceramic, said method consisting of the steps of 10. A method according to claim 9 wherein said glass-formers are borosilicate type. |
summary | ||
054266815 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The combined active/passive ECCS network having three active divisions and one passive division in accordance with a preferred embodiment of the invention is depicted in FIG. 1. The active components are primarily those of the ABWR with either a reactor core isolation cooling (RCIC) or a reactor heat vent and level control (RHVLC) and with either one or two high-pressure core flooder (HPCF) units. The high-pressure active components for the configuration evaluated are not needed to satisfy the design basis requirements. Therefore, the high-pressure units can be classified as non-safety in the three active divisions and fourth passive division concept in accordance with the invention. However, retaining the HPCF system in Division III will result in a lower core damage frequency. For this case, the number of active components in the active/passive system will be the same as in the ABWR. Alternatively, it would be possible to apply the combined active/passive ECCS network using the ABWR ECCS active safety systems. In this case, consideration for a third safety-related HPCF and non-safety classification for the RCIC can be entertained depending on the desire or need to perform on-line maintenance of the high-pressure ECCS equipment. As shown in FIG. 1, each active division (Divisions I to III) has a low-pressure flooder (LPFL) comprising a diesel generator 10 which drives a pump 12. Pump 12 pumps water from the suppression pool 8 to the RPV 2 via a reactor heat removal heat exchanger (RHR-HX) 14. Each RHR-HX is in turn coupled to the ultimate sink (e.g., seawater, riverwater or cooling pond) by an intermediate heat exchanger (RSW-HX) 16. Each RSW-HX pump (not shown) is also driven by a respective diesel generator 10. The RHR-HX/RSW-HX pairs combine to remove heat from the water supplied to the RPV from the suppression pool. The LPFL also operates in other modes: (1) means (not shown) are provided for water from the RHR-HX to be piped back to the suppression pool in a pool cooling mode; (2) the RHR-HX is connected to a drywell sparger (not shown) for spraying cooling water into the drywell when the drywell temperature is too high; and (3) the RHR-HX is connected to a wetwell sparger (not shown) for spraying cooling water above the suppression pool when the wetwell temperature is too high. Division I of the active system further includes a reactor heat vent and level control (RHVLC) which comprises a high-pressure pump 18 which is powered by the steam turbine 52. The RHVLC pumps water from the suppression pool to the feedwater line 20, which carries the water to the RPV. RHVLC turbine 52 also drives a generator (not shown) which is used to recharge the batteries. Alternatively, a conventional RCIC steam turbine system could be used. Division II of the active system further includes a high-pressure core flooder (HPCF) 22, which comprises a high-pressure pump driven by a diesel generator. HPCF 22 pumps water from the suppression pool to RPV 2 via an injection line. The passive division components include the GDCS, the PCCS, the RHR-CND and the release valves. Each of these systems will be described in detail below. The GDCS provides reactor vessel inventory to the annulus region of the reactor after a LOCA and reactor depressurization. The GDCS is composed of a short-term pressurized subsystem (GDCS-ST) 24, which is responsible for filling and reflooding the core following a LOCA, and a long-term subsystem (GDCS-LT) 26, which provides long-term reactor inventory supply following discharge of the GDCS-ST. Both subsystems provide cooling water under force of gravity from water pools located within the drywell at an elevation above the active core region to replace RPV water inventory lost during a LOCA event and subsequent decay heat boil-off. The GDCS-ST provides short-term water makeup; the GDCS-LT provides long-term post-LOCA reactor vessel inventory makeup to meet long-term core decay heat boil-off requirements. The GDCS-ST 24 may consist of one tank (as shown in FIG. 1) or two or more tanks (as shown in FIG. 5) located between the suppression pool and the condenser pool. The GDCS-ST and GDCS-LT are both placed in the PCV 6 at the same regional elevation of the main steam line 28. FIG. 5 shows the arrangement in PCV 6 of a GDCS having two GDCS-LT tanks and two GDCS-ST tanks, where access is provided between the GDCS-ST tanks for removal or entry of equipment from or into the drywell 11. The GDCS-ST 24 (as shown in FIG. 2) is slightly pressurized at 100 psia and is initiated by a low reactor water level signal. This subsystem operates following reactor depressurization. A time delay is provided for the opening of the squib valves 54 in the line from the GDCS-ST pool after a Level 1 signal is received. A check valve 56 is provided in the line leading to the RPV to eliminate any backflow from the RPV to the GDCS-ST tank. The GDCS-ST tank has a sufficient supply of water to flood the RPV to a depth above the fuel rods. The GDCS-LT 26 consists of two tanks which supply reactor water after a time delay of approximately 1/2 to 3 hr after the initial low-level signal is received, at which time the GDCS-LT squib valves 54 are fired electrically to open. Upon opening of the squib valves, the gravity head causes water from the GDCS pools to flow into the RPV 2. An additional valve in the GDCS line is called a thermal-actuated deluge valve 30. This valve opens when the temperature in the lower drywell rises above a predetermined threshold, resulting in water from the GDCS-LT filling the lower drywell cavity surrounding the RPV. The GDCS-ST and GDCS-LT supply water to the RPV by way of two nozzles on the vessel. Each RPV injection line nozzle contains a flow limiter (not shown) of a venturi-like shape. On each injection line there is a locked-open, manually operated isolation maintenance valve 58 located near the vessel nozzle and another such valve located near the water source. A test connection valve 60 downstream of the check valve allows the latter to be tested during refueling outages. In each GDCS injection line, a check valve 56 is located upstream of squib valve 54. The squib valve is designed to withstand reactor pressure without leakage during operation. Once RPV Level 1 is reached, vessel depressurization is initiated and timers are started in GDCS logic which will actuate the squib valves. Once actuated, the squib valves provide a permanent open flow path from the GDCS source to the RPV. The check valves prevent backflow to the pools after the squib valves are actuated and the vessel pressure is still higher than the pool pressure plus its gravity head. Once the vessel pressure has decayed below the pool pressure, the differential pressure will open the check valve and allow water to begin flowing into the vessel. The check valves are designed such that they remain partially open when zero pressure difference exists across the valve. This minimizes the potential for sticking in the closed position during long periods of non-use. Pool suction lines have an intake strainer (not shown) to prevent entry of debris material that might be carried into the pool, such as during a large break LOCA. The thermally actuated deluge valves are connected respectively to two GDCS-LT downcomer lines 32, which provide a means to flood the lower drywell cavity in the event of a core melt sequence that causes failure of the lower vessel head and allows molten fuel to reach the cavity floor. The PCCS 34 (as shown in FIG. 3) maintains the containment 6 within its pressure limits for design basis accidents. The system is designed as a passive system with no components that function actively, and it is also designed for conditions that equal or exceed the upper limits of containment reference severe accident capability. The PCCS operates by natural circulation. The PCCS is initially driven by the pressure difference created between the containment drywell and the suppression pool during a LOCA and then by gravity drainage of steam condensed in the tubes, so they require no sensing, control, logic or power-actuated devices to function. The PCCS is an extension of the safety-related containment and has no isolation valves. The PCCS consists of two low-pressure, totally independent loops, each loop containing a heat exchanger that condenses steam and transfers heat to water in a large condenser pool which is vented to atmosphere. Each PCC condenser (as well as each RHR-CND) is submerged in a respective compartment of the condenser pool 36 (see FIG. 6) located high in the reactor building at approximately the same elevation as the fuel pools. The condenser pool is above and outside of the drywell (see FIG. 1). A combination of one PCCS unit and one RHR-CND unit is designed to discharge into a single GDCS-LT unit. The suppression pool may be connected to the GDCS-LT pool via the LPFL pumps 12 for replenishing the gravity-driven water supply from the suppression pool. Each PCC condenser (one of which is shown in FIG. 3) has an upper drum and a lower drum connected by condenser tubes. A steam-gas mixture enters the PCCS through a line 38 directly from the drywell, so that no valves need be opened and the PCCS is always in a "ready standby" mode. The steam is then condensed in the condenser tubes and falls to the lower drum. From the lower drum, the noncondensables can be vented through a line 40 which is submerged in the suppression pool 8. The condensed water is drained to the GDCS-LT pool 26. A U-shaped bend 42 in the pipeline in the GDCS-LT pool is provided to form a water trap that prevents steam from entering the PCCS condenser from the GDCS pool airspace. Heat from the PCCS condenser will cause the condenser pool temperature to rise to a point where it will boil. The resulting steam will vent outside the reactor building. A continuous circulation of steam entering the PCCS condenser, being condensed and flowing to the GDCS-LT pool, is established to remove containment heat. A GDCS line, which is part of the GDCS system, is shown in FIG. 3 connecting the GDCS-LT pool to the RPV. This line is included as part of FIG. 3 to show the capability for the condensed PCCS water that enters the GDCS-LT pool to subsequently return to the RPV. The two PCC condensers are made of two identical modules. The two condensers provide containment cooling after a LOCA and limit containment pressure to less than its design pressure for at least 72 hours after a LOCA. The PCCS condenser has a central steam supply pipe which is open to the containment at its lower end and which feeds two horizontal headers through two branch pipes at its upper end. Steam is condensed inside vertical tubes and the condensate is collected in two lower headers. The vent and the drain lines from each lower header are routed to the drywell through a single containment penetration. The condensate drains into an annular duct around the vent pipe and then flows in a line which connects to a large common drain line which also receives flow from the other header. As seen in FIG. 6, each condenser is located in a respective subcompartment of the condenser pool 36, and all pool subcompartments communicate at their lower ends to enable full utilization of the collective water inventory, independent of the operational status of any given subloop. A valve is provided at the bottom of each condenser pool subcompartment that can be closed so the respective subcompartment can be emptied of water to allow condenser maintenance. Condenser pool water can heat up to about 101.degree. C. (214.degree. F.). The steam which is formed, being nonradioactive and having a slight positive pressure relative to station ambient pressure, is vented from the steam space above each PCC condenser segment. The steam is released to the atmosphere through large-diameter discharge vents 46. A moisture separator is installed at the entrance to the discharge vent lines to preclude excessive moisture carryover and loss of condenser pool water. Condenser pool make-up clean water supply for replenishing the level is provided from a so-called "make-up demineralized system" (not shown). Level control is accomplished by using an air-operated valve in the make-up water supply line. The valve opening/closing is controlled by a water level signal sent by a level transmitter sensing water level in the condenser pool. Cooling/clean-up of the condenser pool water is performed by a fuel and auxiliary pools cooling system (not shown). Several suction lines, at different locations, draw water from the sides of the condenser pool at an elevation above the minimum water level that is required to be maintained during normal plant operation. The water is cooled/cleaned and then returned to the condenser pool. On the return line for condenser pool water recirculation flow, there is also a post-LOCA pool water make-up connection. As seen in FIG. 6, the passive division incorporates a pair of RHR condensers 48 located in respective subcompartments of the condenser pool. One such RHR-CND is illustrated in FIG. 4. Operation of the RHR-CND is initiated by opening the associated release valve 50 connected to the main steam line 28 from the RPV. When release valve 50 is opened, the difference between the pressure in RPV 2 and the pressure in the associated GDCS-LT tank 26 forces the steam flow from the RPV to the RHR-CND. This steam is condensed in the condenser tubes and then the condensate is returned to the GDCS-LT pool 26. The steam condensation heats the condenser pool 36 to the boiling point, producing nonradioactive steam which is vented outside the reactor containment building. With the low-pressure GDCS-LT system connected downstream to RHR-CND 48, the operating pressure of RHR-CND 48 will be in the low to intermediate range (i.e., 100-200 psi). The pressure drop across the RHR-CND is of the order 10-30 psi at full capacity. The RHR-CND can be classified as non-safety, and the unit operation is designed for emergencies involving multiple failures of the other active components. The release valves (RLVs) 50, one of which is shown in FIG. 4, provide a similar function as the plant automatic depressurization system (ADS) valves attached to the steam lines from the RPV. The valve capacity is about half that of an ADS valve. The RLVs provide a backup depressurization function, and allow RHR-CND 48 to come into use under the conditions of the RHR-HXs 14 becoming unavailable or the suppression pool temperature approaching the safety limit. The manually operated RLVs 50 need to be designed so that the RHR-CND does not exceed its MW capacity or design pressure. The RHR-CND could be used in conjunction with the RPV in its operating pressure range of 50 to 1250 psia. The RHR-CND is designed primarily for accident control and mitigation. The condenser pool associated with the RHR-CND does not require safety-related pool cooling. The RHR-CND flow/heat-load is governed by the forced pressure difference between the RPV and the GDCS-LT system. The forced operation feature of the RHR-CND enables it to utilize its full capacity and perform several functions. These functions include: (1) controlled reactor depressurization; (2) backup to the active RHR-HXs in the event of pump failure or high suppression pool temperature; and (3) reactor inventory supply in conjunction with the GDCS-LT. During accident progression, active systems are the first to initiate, followed by passive systems in the event the active system components fail or as conditions deteriorate. Severe accident features (such as containment venting) are a last resort in the accident progression. The passive systems act as a buffer between the initiation of active systems and severe accident features. Therefore the probability of approaching a severe accident condition, which requires exercising the severe accident features, is greatly reduced. The passive components included in the fourth division provide multiple backups to the active components covering the BWR safety functions of reactor inventory supply, containment and shutdown cooling, ultimate sink, reactor depressurization and station blackout mitigation. The GDCS-ST and GDCS-LT serve as backups to the LPFL and HPCF/RHVLC for reactor inventory supply. The PCCS and RHR-CND serve as backups to the RHR-HXs for suppression pool/containment cooling. The condenser pool serves as a backup to the ultimate sink. The RLVs serve as backup to the ADS valves for reactor depressurization. The RLVs/GDCS/RHR-CND serve as backup to the RHVLC for station blackout mitigation. Passive components by their nature require less maintenance than active systems. The addition of a passive fourth division to the three active divisions of the ABWR will have a minimum impact on the ABWR total maintenance activities for the ECCS network. With the passive division, ECCS on-line maintenance can be performed. This would reduce or eliminate the ECCS maintenance activities during outage, with the consequence that the outage is shortened in duration and plant availability is improved. Maintenance activities on the passive components are much reduced as compared to the maintenance required on active systems. Passive component maintenance mainly involves valve testing, flushing of common injection lines to prevent crud build-up and unplug the line, and ultrasonic inspection of passive heat exchanger tubes. The addition of a passive fourth division to the three active divisions of the ABWR also provides redundancy and diversity in power supply and component design with the resultant improvement in plant safety. In particular, the passive division meets the BWR design basis requirements with on-line maintenance by providing additional sources of reactor inventory supply (GDCS) and primary containment cooling (PCCS). The passive division also provides additional and diverse severe accident mitigating features. The PCCS stabilizes the containment following core damage. The presence of the PCCS in the PCV with no moving parts would reduce or avoid containment venting. With the PCCS, the conditional containment failure probability will be lower. The GDCS enables lower drywell flooding. The thermalactuated deluge valves open when the temperature rises to a predetermined threshold in the drywell, resulting in emptying of the GDCS tank and filling of the lower drywell cavity surrounding the RPV. The RHR-CND serves to mitigate station blackout and provide back-up to the RHR-HXs with N+2 capability. The passive division also enables classification of the high-pressure active components as non-safety systems. Classifying the high-pressure components of the HPCF and RHVLC as non-safety simplifies plant operation and response to transients. Limiting the combined active and passive concept to one passive division enables the optimization of the ECCS network, thereby minimizing the number and sizes of the passive components. However, the size of the PCV may need to be increased slightly to compensate for the GDCS tanks and associated piping. Also, the reactor building fueling floor must be rearranged to incorporate the condenser pool. With this in mind, it is expected that the plant initial cost will increase somewhat, with the increase in initial cost being offset at least in part by a decrease in operating maintenance cost. The foregoing preferred embodiment of the invention has been disclosed for the purpose of illustration. Variations and modifications of the disclosed apparatus will be readily apparent to practitioners skilled in the art of safety systems for BWRs. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter. |
abstract | Systems and methods for calibrating and controlling leaves of a multi-leaf collimator are disclosed. According to an exemplary method, a controller may receive images of collimator leaves and may determine a minor offset between an imaging marker and the tip of the respective leaf. In addition, the controller may quantify a barrel distortion effect associated with a leaf-imaging camera. The controller may correct leaf position data using the minor offsets and barrel distortion quantification, and may use the corrected leaf positions to accurately place the leaves during radiotherapy. Advantageously, a desired beam shaping window may be formed with the leaves, ensuring that healthy tissue is minimally irradiated while also ensuring that the target tissue receives the correct radiation dose. Embodiments of the present disclosure provide collimator calibration techniques which may be faster than prior calibration techniques, allowing shortened calibration times and faster radiotherapy sessions. |
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053496175 | claims | 1. A pressurized water nuclear reactor, comprising: a. a vessel forming a first ferrule (2); b. a second ferrule (6) contained in the vessel (2) and concentric therewith, said first and second ferrules defining an external annular compartment (3) through which primary coolant flow water circulates to provide a hairpin-shaped flow path F of said primary coolant from top to bottom of said vessel (2); c. a central cylindrical compartment (10) containing a core (4) for heating said water, said flow path F of primary water continuing from the bottom of said vessel, through said core, to the top of cylindrical compartment; d. an apparatus for removing residual power from the core of said reactor, said apparatus comprising a third ferrule (7) defining a complementary annular space (8) between said external annular compartment (3) and said central cylindrical compartment (10), said complementary annular space (8) including a first orifice (12) for permitting communication of flow water between a lower part of said complementary annular space (8) and a lower part of said external annular compartment (3), said complementary annular space (8) also including a second orifice (11) for permitting communication of flow water between said cylindrical compartment (10) and an upper part of said complementary annular space, said apparatus also comprising at least one self-contained auxiliary heat exchanger (9) for removing said residual power, said heat exchanger being located in said complementary annular space (8) and being supplied autonomously with a second heat transfer fluid separate from said flow water; and e. means for increasing vacuum effect in the vicinity of the lower part of the complementary annular space, wherein said means for increasing vacuum effect comprises means for partially closing the first orifice and a series of cylindrical radial pipes (16) extending, in the vicinity of the base of the core, into the external annular compartment (3), said cylindrical radial pipes (16) issuing onto openings (15) provided on the periphery of the lower part of the intermediate ferrule in the complementary annular space and being provided on their wall with longitudinal slots (17) for communication with the external annular compartment (3). a. a vessel forming a first ferrule (2); b. a second ferrule (6) contained in the vessel (2) and concentric therewith, said first and second ferrules defining an external annular compartment (3) through which primary coolant flow water circulates to provide a hairpin-shaped flow path F of said primary coolant from top to bottom of said vessel (2); c. a central cylindrical compartment (10) containing a core (4) for heating said water, said flow path F of primary water continuing from the bottom of said vessel, through said core, to the tope of said cylindrical compartment; d. an apparatus for removing residual power from the core of said reactor, said apparatus comprising a third ferrule (7) defining a complementary annular space (8) between said external annular compartment (3) and said central cylindrical compartment (10), said complementary annular space (8) including a first orifice (12) for permitting communication of flow water between a lower part of said complementary annular space (8) and a lower part of said external annular compartment (3), said complementary annular space (8) also including a second orifice (11) for permitting communication of flow water between said cylindrical compartment (10) and an upper part of said complementary annular space, said apparatus also comprising at least one self-contained auxiliary heat exchanger (9) for removing said residual power, said heat exchanger being located in said complementary annular space (8) and being supplied autonomously with a second heat transfer fluid separate from said flow water; and e. means for increasing vacuum effect in the vicinity of the lower part of the complementary annular space, wherein the means for increasing vacuum effect comprises means for partially closing the first orifice and in the external annular compartment (3) below the auxiliary heat exchanger (9), an annular chamber (22) linked by a series of openings (24) with the complementary annular space (8) and, by an annular slot (20), with the external annular compartment (3), said annular chamber (22) having an extension in the radial direction of the vessel (2) such that it creates, in the external annular compartment (3), a narrowing (19) which leads to an increase, at the location of the slot (20), to the flow rate of the primary flow descending into the external annular compartment (3) of the auxiliary heat exchanger (9). 2. A reactor according to claim 1, wherein there are two such longitudinal communicating slots (17) on each pipe (16) said slots being positioned azimuthly on the surface of the cylindrical pipes at an angle .phi. close to 80.degree. with the main flow path in the external annular compartment (3). 3. A pressurized water nuclear reactor, comprising: 4. A pressurized water nuclear reactor including a reactor vessel (2) having a primary water circulation in accordance with a hairpin path in the reactor vessel (2) and having two concentric ferrules (2, 6) defining an external annular compartment (3), in which cold primary water describes a downward path and a central cylindrical compartment (10) containing the actual core (4), in which the primary water flows from bottom to top, accompanied by heating, through the core, a third ferrule (7) defining a complementary annular space (8) between the two preceding compartments (3, 10), said annular space (8) being linked in this lower part by a first orifice (12) issuing into the external annular compartment (3) with the cold water of the primary circuit and in its upper part and by a second orifice (11) issuing into the central compartment (10) with the hot water of the primary circuit and an auxilliary heat exchanger (9) located in said complementary annular space (8), said auxiliary exchanger (9) being supplied autonomously by a second heat transfer fluid, which is independent of the primary cooling water of the reactor core, and means for increasing vacuum effect in the vicinity of the lower part of the complementary annular space, said means for increasing the vacuum effect incorporating means for partially closing the first orifice and a series of cylindrical radial pipes (16) extending, in the vicinity of the base of the core, into the external annular compartment (3), said cylindrical radial pipes (16) issuing onto openings (15) provided on the periphery of the lower part of the intermediate ferrule in the complementary annular space and being provided on their wall with longitudinal slots (17) for communication with the external annular compartment (3). 5. A reactor according to claim 4, wherein there are two such longitudinal communicating slots (17) on each pipe (16), said slots being positioned azimuthly on the surface of the cylindrical pipes at an angle .phi. close to 80.degree. with the main downward flow direction in the external annular compartment (3). 6. A reactor according to claim 4, wherein the means for increasing vacuum effect comprises means for partially closing the first orifice and, in the external annular compartment (3) below the auxiliary heat exchanger (9), an annular chamber (22) linked by a series of openings (24) with the complementary annular space (8) and, by an annular slot (20), with the external annular compartment (3), said annular chamber (22) having an extension in the radial direction of the vessel (2) such that it creates, in the external annular compartment (3), a narrowing (19) which leads to an increase, at the location of the slot (20), to the flow rate of the primary flow descending into the external annular compartment (3) of the auxiliary heat exchanger (9). |
051270293 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIGS. 1A and 1B are schematic views, respectively, showing an X-ray exposure apparatus according to an embodiment of the present invention. In FIG. 1A the apparatus includes an X-ray directing means 20, and in FIG. 1B one component of the directing means 20 is illustrated in detail. In FIGS. 1A and 1B, denoted at 1 is an emission point of a SOR source 10; at 2 is a cylindrical mirror having a convex surface, which is one component of the directing means 20; at 3 is a mask; and at 4 is a semiconductor wafer. In this embodiment, the SOR source 10 and the directing means 20 cooperate to provide an illumination system. The mask 3 has a semiconductor circuit pattern formed thereon, which pattern is at the surface 3a to be exposed. In a zone at the peripheral part of this surface 3a, one or more alignment marks are formed. The mask 3 is supported by a mask stage 30. On the other hand, the wafer 4 has a surface which is coated with a resist, and the wafer 4 is placed on a wafer stage 40. Each of the mask stage 30 and the wafer stage 40 is movable vertically and horizontally as viewed in FIG. 1A, as well as in a direction perpendicular to the sheet of the drawing. With the displacement of the stages 30 and 40, the mask 3 and the wafer 4 can be aligned into a predetermined positional relationship. The apparatus of the present embodiment is an exposure apparatus of proximity type. In the present embodiment, an X-ray beam emitted from the emission point 1 of the SOR source 10 is received by the directing means 20 which serves to adjust the sectional shape of the received X-ray beam and to correct the intensity distribution thereof. The directing means 20 directs the X-ray beam to the mask 3 and irradiates the pattern 3a of the mask 3 with this X-ray beam. Then, the X-ray beam is projected to the wafer 4 through the mask 3, by which the pattern of the mask 3 is transferred to the resist on the wafer 4. The mirror 2 of the directing means 20 serves to expand the diameter of the X-ray beam, obliquely incident on the reflection surface thereof, with respect to a sectional plane perpendicular to the generating line of the mirror 2. Also, the mirror 2 serves to reduce non-uniformness in intensity distribution of the X-ray beam on the mask 3 surface. Further, the mirror 2 is disposed so that its generating line extends substantially in parallel to the horizontal orbit plane of the SOR source 10. While in FIG. 1B only the mirror 2 is illustrated as a component of the directing means 20, actually the directing means 20 include some other components. Examples are: a beryllium (Be) window provided to shield the inside of the exposure apparatus against the outside atmosphere to allow that the inside of the apparatus is maintained at a vacuum or it is filled with an He gas; a stop member for determining the size of the X-ray beam in accordance with the size of the pattern 3a region of the mask 3; a shutter mechanism for controlling the amount of exposure by the X-ray beam. FIG. 2 is a sectional view of the illumination system shown in FIGS. 1A and 1B, taken on a plane including the optical axis thereof and along the vertical direction (direction y.sub.a), as viewed in FIG. 1B. Like numerals as in FIGS. 1A and 1B are assigned to the elements corresponding to those of FIGS. 1A and 1B. In FIG. 2, denoted at d.sub.1 is the distance from the emission point 1 of the SOR source 10 to the center of effective X-ray beam diameter on the reflection surface of the mirror 2; at d.sub.2 is the distance from the center of effective X-ray beam diameter on the reflection surface of the mirror 2 to the mask 3; at R is the radius of curvature of the mirror 2 in a sectional plane perpendicular to the generating line of the mirror; at .alpha. is the angle defined at the center of effective X-ray beam diameter on the reflection surface of the mirror 2, between the reflection surface and the X-ray beam projected thereto from the SOR source 10; at .sigma.' is a standard deviation (angle: rad) of a distribution of intensities of X-rays from the SOR source 10, having different emission angle, in a sectional plane perpendicular to the generating line of the mirror 2 and including the center 1a of the effective diameter, at the gravity center wavelength of the X-ray beam in the used wavelength region. Here, in a sectional plane (x.sub.a -y.sub.a plane) perpendicular to the horizontal orbit plane of the SOR source 10 (x.sub.a -z.sub.a plane in FIG. 1B), the rays of the X-ray beam from the SOR source 10 having different emission angles have a distribution of intensity which is usually in the form of a Gaussian distribution. In FIGS. 1A, 1B and 2, the X-ray beam emitted from the emission point 1 of the SOR source 10 goes along a path in the neighborhood of a plane parallel to the horizontal orbit plane of the SOR source 10, and impinges on the convex mirror 2 having a cylindrical shape. Since the mirror 2 has a curvature with respect to the plane of vertical section (x.sub.a -y.sub.a plane), it serves to reflect the received X-ray beam so as to expand the angle of divergence of the X-ray beam in the vertical direction (y.sub.a direction). As a result, on the mask 3 surface, an X-ray beam expanded sufficiently in the vertical direction, is obtainable. Further, in the present embodiment, the mirror 2 is structured so as to satisfy equations to be set forth below, to thereby reduce the non-uniformness in intensity distribution (Gaussian distribution) of the X-ray beam on the mask 3 surface, with respect to the vertical (y.sub.a) direction. EQU R=(2d.sub.1 d.sub.2 .sigma.')/{[.DELTA.-(d.sub.1 +d.sub.2).sigma.'].multidot..alpha.} . . . (1--1) where d.sub.1 : the distance from the emission point of said X-ray source to the center of effective X-ray beam diameter on said reflection surface; PA0 d.sub.2 : the distance from the center of effective X-ray beam diameter on said reflection surface to the center of effective X-ray beam diameter on the surface to be exposed; PA0 .alpha.: the angle defined at the center of effective X-ray beam diameter on said reflection surface, between the X-ray beam and said reflection surface; PA0 .sigma.': a standard deviation of a distribution of intensities of the X-rays having different angles of emission from said X-ray source, in a sectional plane perpendicular to a generating line of said mirror, at the gravity center wavelength of the X-ray beam from said X-ray source; PA0 .DELTA.: 0.43a.ltoreq..DELTA..ltoreq.4.0a (1-2); and PA0 a: the length of the surface to be exposed, with respect to a direction which is substantially perpendicular to the generating line of said mirror. EQU R=(2d.sub.1 .multidot.d.sub.2)/{[.DELTA.'-(d.sub.1 +d.sub.2)].multidot..alpha.} (2-1) where EQU 4.3.times.10.sup.2 a.ltoreq..DELTA.'.ltoreq.4.0.times.10.sup.4 a (2--2) Next, description will be made of the significance and the derivation of equations (1-1), (1-2), (2-1) and (2-2) that determine an appropriate curvature radius R of the cylindrical convex mirror 2. First, the focal length f of the mirror 2 in the sectional plane (x.sub.a -y.sub.a plane) perpendicular to the generating line of the mirror 2 is given by: EQU f=-Rsin .alpha./2.perspectiveto.-R.alpha./2 (3) where EQU R>0, .alpha.>0 and f<0 Approximating the mirror 2 as a thin lens, examination will now be made as to the degree of expansion, on the mask 3 surface, of the paraxial rays emitted from an object point on the axis (SOR emission point 1). FIG. 3 illustrates the paraxial relation of the thin lens. In this Figure, denoted at 1' is the center position of the emission point 1 of the SOR source 10; at 2' is a thin lens which is an approximation of the mirror; at 3' is a surface that corresponds to the mask 3; at .phi. is the refracting power (=1/f) of the thin lens (i.e. the mirror); d.sub.1 is the distance from the emission point center position 1' to the thin lens 2'; at d.sub.2 is the distance from the thin lens 2' to the surface 3'; at u.sub.1 and u.sub.2 are half angles of divergence of the X-ray beams just emitted from the emission point center position 1' and the thin lens 2', respectively; at h.sub.1 and h.sub.2 are the radii (heights of incidence) of the X-ray beams, in the y.sub.a direction, incident on the thin lens 2' and the surface 3', respectively. The signs are such as those illustrated in the drawing. From the paraxial relationship, the following relations are established: EQU h.sub.1 =-d.sub.1 .multidot.u.sub.1 (4) EQU u.sub.2 =u.sub.1 +h.sub.1 4 (5) EQU h.sub.2 =h.sub.1 -d.sub.2 u.sub.2 (6) By erasing h.sub.1 and u.sub.2 in equations (4)-(6), h.sub.2 is expressed as follows: EQU h.sub.2 =-u.sub.1 (d.sub.1 +d.sub.2)+d.sub.1 d.sub.2 u.sub.1 4 (7) Further, by substituting the relation of equation (3) into equation (7), then EQU h.sub.2 =-u.sub.1 (d.sub.1 +d.sub.2)-2u.sub.1 d.sub.1 d.sub.2 /R.alpha. (8) This is the paraxial relation that represents the expansion of the X-ray beam upon the surface 3', in the vertical (y.sub.a) direction. Accordingly, if a standard deviation of a distribution of intensities of the rays having different emission angles, in the vertical direction, at the gravity center wavelength in the used wavelength region of the X-ray beam is denoted by .sigma.' (>0), the height (radius) h'.sub.2 from the optical axis and on the surface 3' of the X-ray beam as emitted from the center position 1' at an angle substantially corresponding to this standard deviation .sigma.', can be determined by substituting u.sub.1 =-.sigma.' into equation (8), as follows: EQU h'.sub.2 =.sigma.'(d.sub.1 +d.sub.2)+2.sigma.'d.sub.1 d.sub.2 /R.alpha. (9) By taking the height h'.sub.2 in equation (9) as an effective radius .DELTA. nd by solving the equation with respect to R, then equation (1-1) is obtained. Also, as will be understood from the foregoing description, equation (1-2) is the one that defines the range for the position of the X-ray beam that determines the standard deviation. With smaller .DELTA., a beam of a small expansion is obtainable, and with larger .DELTA., a beam of large expansion is obtainable. Next, the range of the value .DELTA. suitable for the exposure will be explained. FIG. 4 is a graph showing the relationship between (i) a/.DELTA. (the ratio of the length a (exposure area) of the surface 3a, to be exposed, in the y.sub.a direction, to the effective radius .DELTA. of the X-ray beam on the surface 3' in the y.sub.a direction) in an occasion where, in the x.sub.a -y.sub.a plane, the rays of the X-ray beam having different emission angles have a distribution of intensity which is a Gaussian distribution, and (ii) the non-uniformness on the surface 3a, to be exposed, in the y.sub.a direction (i.e. the value of "(maximum exposure strength - minimum exposure strength)/maximum exposure strength)" on the surface 3a to be exposed, with respect to the y.sub.a direction). Here, it is assumed that there is no variation in intensity (i.e. non-uniformness in exposure) with respect to the z.sub.a direction. It is seen from this graph that, in order to maintain the non-uniformness in exposure not higher than 50%, it is necessary to keep the ratio a/.DELTA. within a range not greater than about 2.3. In other words, unless .DELTA. is made not less than about 0.43a, the magnitude of the difference between the maximum exposure strength and the minimum exposure strength on the surface 3a to be exposed (exposure area), with respect to the maximum exposure strength, is reduced to a half or less. As a result, large non-uniformness in exposure is produced on the surface 3a to be exposed, which is not proper for use in an exposure apparatus. FIG. 5 is a graph showing the relationship between (i) the ratio a/.DELTA. as described which is under the same condition as FIG. 4, and (ii) the ratio of used light quantity (the ratio of the amount of exposure on the surface 3' to the total quantity of the X-ray beam emitted from the light source). Here, it is assumed that the length of the surface 3a to be exposed, in the z.sub.a direction is substantially equal to the width of the X-ray beam in the z.sub.a direction and that it is unchangeable. In other words, this graph is one that shows to what extent the X-ray beam from the SOR source is effectively used in exposure, in dependence upon the value a/.DELTA.. If, in this graph, the ratio of used light quantity is less than 1/10, the efficiency is too low for use in an exposure apparatus. It is seen from this graph that, in order to make the ratio of used light quantity not less than 1/10, it is necessary to keep a/.DELTA. not less than 0.25. Namely, .DELTA..ltoreq.4a is the condition for an X-ray exposure apparatus. As will be understood from the foregoing, the condition for .DELTA. in an X-ray exposure apparatus is determined, in consideration of the length a of the surface 3a, to be exposed, with respect to the X-ray beam expanding direction (y.sub.a direction in FIG. 1), as depicted by equation (1-2), and it is determined as follows: EQU 0.43a.ltoreq..DELTA..ltoreq.4.0a Accordingly, as showed by equation (1-1), the radius of curvature of the mirror 2 is expressed as follows: EQU R=(2d.sub.1 .multidot.d.sub.2 .multidot..sigma.')/{[.DELTA.-(d.sub.1 +d.sub.2).multidot..sigma.'].multidot..alpha.} where EQU 0.43a.ltoreq..DELTA..ltoreq.0.40a Namely, the condition is ##EQU2## On the other hand, if .DELTA.'=.DELTA./.sigma.', as showed by equation (2-1), equation (1-1) can be rewritten as follows: EQU R=(2d.sub.1 .multidot.d.sub.2)/{[.DELTA.'-(d.sub.1 +d.sub.2)].multidot..alpha.} where .DELTA.'=.DELTA./.sigma.' Generally, in a SOR light source, the value .sigma.' is in the range of EQU 0.1.times.10.sup.-3 .ltoreq..sigma.'.ltoreq.1.0.times.10.sup.-3 (rad) (11) Therefore, from equations (1-2) and (11), as showed by equation (2-2), the following relation is obtained: EQU 4.3.times.10.sup.2 a.ltoreq..DELTA.'.ltoreq.4.0.times.10.sup.4 a Thus, equations (2-1) and (2-2) can be rewritten as follows: ##EQU3## From these equations, the following relation is determined: ##EQU4## Examples will now be described in detail. The parameters in equation (1-1) were set as follows: EQU d.sub.1 =5000 (mm) EQU d.sub.2 =5000 (mm) EQU .sigma.'=5.0.times.10.sup.-4 (rad) EQU .alpha.=1.0.times.10.sup.-2 (rad) EQU a=30 (mm) By setting the parameters in this manner, from equation (10) the following relation is obtained: EQU 2.2.times.10.sup.4 .ltoreq.R.ltoreq.3.2.times.10.sup.5 (13) (unit:mm) With the parameters set in accordance with the above-described described conditions, the relationship between the surface 3' to be exposed (exposure area) and the X-ray exposure strength was examined by using cylindrical convex mirrors of R=250 m (.DELTA.=a/2), R=100 m(.DELTA.=a) and R=45.5 m (.DELTA.=2a), respectively. The results are illustrated in FIG. 6. In this Figure, the axis of abscissa depicts the position in the y.sub.a direction with the origin being at the center of the surface 3' to be exposed, while the axis of ordinate depicts the X-ray intensity at each point (relative value of the X-ray intensity at each point, relative to the value of the intensity of the X-ray beam emitted from the cylindrical convex mirror as integrated with respect to the area on the surface 3' to be exposed). As a comparative example, FIG. 7 shows a graph of the X-ray intensity distribution which was obtained in a similar manner as in FIG. 6, but with use of a cylindrical convex mirror of R=8.5 m (.DELTA.=10a) and a mirror R=.infin.(.DELTA.=a/6), namely, a flat mirror, which were out of the range of the curvature radius as determined in accordance with the present invention. As will be understood from these Figures, if R is within the range as defined in accordance with the present invention, the non-uniformness in exposure can be kept within a tolerable range (the light quantity at the peripheral portion is not less than a half of that of the central portion). Also, the proportion of the quantity of used light to the total quantity is not too small. With the X-ray exposure apparatus of the present embodiment, as described hereinbefore, the whole surface to be exposed can be exposed at a time. This avoids the possibility of local thermal distortion of the mask due to displacement of the X-ray beam. Further, the apparatus includes a mirror having a radius of curvature which is suitable for ensuring the amount of exposure as required in a practical exposure apparatus, and also which is effective to reduce the non-uniformness in intensity distribution of the X-ray beam. Therefore, the present invention is effective to make it easier to provide a practical X-ray exposure apparatus. While the invention has been described with reference to the structures disclosed herein, it is not confined to the details set forth and this application is intended to cover such modifications or changes as may come within the purposes of the improvements or the scope of the following claims. |
abstract | The present invention provides a method of manufacturing a core shroud for a nuclear plant and a nuclear power plant structure in which a groove portion is easily assembled in the case of manufacturing the core shroud having a weld structure of a nuclear power plant by a laser welding, and it is possible to obtain a weld joint portion in which a plastic distortion region and a residual stress are as small as possible, going with a solidification shrinkage of the weld portion. At a time of welding butted portions of a plurality of members constructing a core shroud, a root face is provided in the butted portion, a length of the root face is set to 25% to 95% of a thickness of the thinner one of the butted portions of a plurality of members, a narrow groove is provided in the other than the root face, and the butted portions are welded by a laser welding using a weld wire. |
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040452883 | description | DESCRIPTION OF THE INVENTION Referring now more particularly to FIG. 1, there is shown a partially cutaway sectional view of a nuclear fuel assembly 10. This fuel assembly 10 consists of a tubular flow channel 11 of generally square cross section provided at its upper end with lifting bale 12 and at its lower end with a nose piece (not shown due to the lower portion of assembly 10 being omitted). The upper end of channel 11 is open at 13 and the lower end of the nose piece is provided with coolant flow openings. An array of fuel elements or rods 14 is enclosed in channel 11 and supported therein by means of upper end plate 15 and a lower end plate (not shown due to the lower portion being omitted). The liquid coolant ordinarily enters through the openings in the lower end of the nose piece, passes upwardly around fuel elements 14, and discharges at upper outlet 13 in a partially vaporized condition for boiling reactors or in an unvaporized condition for pressurized reactors at an elevated temperature. The nuclear fuel elements or rods 14 are sealed at their ends by means of end plugs 18 welded to the cladding 17, which may include studs 19 to facilitate the mounting of the fuel rod in the assembly. A void space or plenum 20 is provided at one end of the element to permit longitudinal expansion of the fuel material and accumulation of gases released from the fuel material. A nuclear fuel material retainer means 24 in the form of a helical member is positioned within space 20 to provide restraint against the axial movement of the pellet column, especially during handling and transportation of the fuel element. The fuel element is designed to provide an excellent thermal contact between the cladding and the fuel material, a minimum of parasitic neutron absorption and resistance to bowing and vibration which is occasionally caused by flow of the coolant at high velocity. A nuclear fuel element or rod 14 is shown in a partial section in FIG. 1 constructed according to the teachings of this invention. The fuel element includes a core or central cylindrical portion of nuclear fuel material 16, here shown as a plurality of fuel pellets of fissionable and/or fertile material positioned within a structural cladding or container 17. In some cases the fuel pellets may be of various shapes such as cylindrical pellets or spheres, and in other cases different fuel forms such as a particulate fuel may be used. The physical form of the fuel is immaterial to this invention. Various nuclear fuel materials may be used including uranium compounds, plutonium compounds, thorium compounds, and mixtures thereof. A preferred fuel is uranium dioxide or a mixture comprising uranium dioxide and plutonium dioxide. Referring now to FIG. 2, the nuclear fuel material 16 forming the central core of the fuel element 14 is surrounded by a cladding 17 which in this invention is also referred to as a composite cladding. The composite cladding has a substrate 21 selected from conventional cladding materials such as a stainless steel and zirconium alloys and in a preferred embodiment of this invention the substrate is a zirconium alloy such as Zircaloy-2. The substrate 21 has metallurgically bonded on the inside diameter thereof a metal barrier 22 so that the metal barrier forms a shield of the substrate from the nuclear fuel material inside the composite cladding. The metal barrier preferably forms about 1 to about 4 percent of the thickness of the cladding and is comprised of a metal selected from the group consisting of niobium, aluminum, copper, nickel, stainless steel and iron. The metal barrier 22 has metallurgically bonded on the inside diameter thereof an inner layer 23 so that the inner layer is the portion of the composite cladding closest to the nuclear fuel material 16. The inner layer preferably forms about 5 to about 15 percent of the thickness of the cladding and is comprised of conventional cladding materials such as stainless steel and zirconium alloys and in a preferred embodiment of this invention the substrate is a zirconium alloy such as Zircaloy-2. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products which have either diffused through or corroded through the inner layer 23 and protects the cladding from contact and reaction with such impurities and fission products. In another preferred embodiment of this invention, the substrate and the inner layer are comprised of the same material and a preferred material is a zirconium alloy such as Zircaloy-2. The composite cladding of the nuclear fuel element of this invention has a metal barrier metallurgically bonded to the substrate and an inner layer metallurgically bonded to the metal barrier. Metallographic examination shows that there is sufficient cross diffusion between the substrate and the metal barrier and between the metal barrier and the inner layer to form metallurgical bonds, but insufficient cross diffusion to alloy with the metal barrier itself. Also from FIG. 2 it is apparent that the metal barrier could be termed a "buried" metal barrier. It has been discovered that a metal barrier of the order preferably of at least about 1 to 4 percent of the wall thickness of the cladding metallurgically bonded to the substrate and the inner layer provides chemical resistance sufficient to prevent propagation of failures from the inner layer to the substrate of the cladding. The metal barrier provides significant chemical resistant to fission products and gases that may be present in the nuclear fuel element and prevents these fission products and gases from contacting the substrate of the composite cladding protected by the metal barrier. For a typical fuel element the substrate of the composite cladding ranges in thickness from 24 to 30 mils, the metal barrier ranges in thickness from 0.5 to 1 mils and the inner layer is approximately 3 mils. The composite cladding used in the nuclear fuel elements of this invention can be fabricated by any of the following methods. In one method a tube of the metal selected to be the metal barrier is inserted into a hollow billet of the material selected to be the substrate, a tube of the material selected to be the inner layer is inserted into the metal barrier tube, and then the assembly is subjected to explosive bonding of the tubes to the billet. The composite is extruded using conventional tube shell extrusion at elevated temperatures of about 1000.degree. to 1400.degree. F. (about 538.degree. to 760.degree. C.). Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. In another method, a tube of the metal selected to be the metal barrier is inserted into a hollow billet of the material selected to be the substrate, a tube of the material selected to be the inner layer is inserted into the tube of the metal barrier and then the assembly is subjected to a heating step (such as at 750.degree. C. for 8 hours) to give diffusion bonding between the tubes and the billet. The composite is extruded using conventional tube shell extrusion such as described above in the immediately preceding paragraph. Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. In still another method, a tube of the metal selected to be the metal barrier is inserted into a hollow billet of the alloy selected to be the substrate, a tube of the material selected to be the inner layer is inserted into the metal barrier tube and the assembly is extruded using conventional tube shell extrusion as described above. Then the extruded composite is subjected to a process involving conventional tube reduction until the desired size of cladding is achieved. The foregoing processes of fabricating the composite cladding of this invention gives economies over other processes used in fabricating cladding such as electroplating or vapor deposition. The invention includes a method of producing a nuclear fuel element comprising making a composite cladding container which is open at one end, the cladding container having a substrate, a metal barrier metallurgically bonded to the inside surface of the substrate and an inner layer metallurgically bonded to the inside surface of the metal substrate, filling the composite cladding container with nuclear fuel material having a cavity at the open end, inserting a nuclear fuel material retaining means into the cavity, applying an enclosure to the open end of the container leaving the cavity in communication with the nuclear fuel, and then bonding the end of the clad container to said enclosure to form a tight seal therebetween. The present invention offers several advantages promoting a long operating life for the nuclear fuel element including the reduction of hydriding of the cladding substrate, the minimization of localized stress on the cladding substrate, the minimization of stress and strain corrosion on the cladding substrate, the reduction of the probability of a splitting failure in the cladding substrate and the prevention of the propagation of stress corrosion cracks through the composite cladding. The invention further prevents expansion (or swelling) of the nuclear fuel into direct contact with the cladding substrate, and this prevents localized stress on the cladding substrate, prevents initiation or acceleration of stress corrosion of the cladding substrate and prevents bonding of the nuclear fuel to the cladding substrate. An important property of the composite cladding of this invention is that the foregoing improvements are achieved with a negligible to moderate neutron penalty (depending on choice of barrier material). Such a cladding is readily accepted in nuclear reactors since the cladding would have minimal eutectic formation (depending on choice of barrier material) in the substrate portion of the cladding during a loss of cooling accident or an accident involving the dropping of a nuclear control rod. Further the composite cladding has a very small heat transfer penalty in that there is no thermal barrier to transfer of heat such as results in the situation where a separate foil or liner is inserted in a fuel element. Also the composite cladding of this invention is inspectable by conventional non-destructive testing methods during various stages of fabrication. In addition to the foregoing, when the zirconium alloy is selected as the substrate and the inner layer, the inside and outside surfaces of the composite cladding are compatible with manufacturing processes for light water nuclear reactor cladding and this enables the use of current manufacturing procedures, lubricants, etchants, etc. Those skilled in the art will gain a further understanding of this invention from the following illustrative, but not limiting, examples of this invention. EXAMPLES 1-4 Billets and inserts were machined, cleaned and assembled by standard procedures for example, and all dimensions were chosen so that the composite billets could be extruded into hot extrusion press. The billets were normal Zircaloy-2 conforming to ASTM B353, Grade RA-1, and the inserts were made of high purity niobium and 304L Stainless Steel (ASTM-A 312). All billet bores and inserts had an 8 mil per in. taper and were pressed together to ensure a good contact between the mating surfaces. The dimensions of the machined parts were as follows: __________________________________________________________________________ Inner Buried Diameter Billet Barrier Barrier Outer Inner Outer Inner Outer Inner Length X Dia. X Dia. Dia. Dia. Dia. Dia. __________________________________________________________________________ 1. Buried Nb Metal Barrier 9.5 .times. 5.74 .times. 2.59 2.59 - 2.44 2.44 - 1.66 2. Buried Nb Metal Barrier 9.5 .times. 5.74 .times. 2.59 2.59 - 2.44 2.44 - 1.66 3. Buried SS Metal Barrier 9.5 .times. 5.74 .times. 2.64 2.64 - 2.44 2.44 - 1.66 4. Buried SS Metal Barrier 9.5 .times. 5.74 .times. 2.56 2.56 - 2.44 2.44 - 1.66 __________________________________________________________________________ Prior to assembling the billets and inserts the mating surfaces were given a light etch to remove traces of impurities. The etchant used for the Zircaloy-2 was a solution of 70 ml H.sub.2 O, 30 ml HNO.sub.3, and PA1 5 ml HF; PA1 7.5 ml H.sub.2 SO.sub.4, PA1 4 ml HNO.sub.3, PA1 31 ml H.sub.2 O, and PA1 2 ml HF. PA1 Extrusion rate -- 6 in/min, PA1 Reduction ratio -- 6:1, PA1 Temperature -- 1,100.degree. F. and PA1 Extrusion force -- 3500 tons. PA1 Outer Diameter -- 2.500 inches, PA1 Inner Diameter -- 1.640 inches, and PA1 Length -- 5 Feet. and for the niobium a solution of 7.5 ml HCL, The stainless steel was polished with fine emery paper and cleaned with acetone and de-ionized water. To improve the chances for a satisfactory bond between the inserts and the billets during extrusion, it was decided to prebond the assemblies. This was accomplished by pressing the tapered inserts into the tapered bore in the billets in vacuum .ltoreq.20 .mu.m while maintaining the billet temperature at 1,400.degree. F. for 8 hours. Forces applied to the inserts during initial pressing ranging from 30-45,000 lbs. To reduce end-losses during the extrusion a 2 inch piece of Zircaloy-2 billet was welded on each end of the composite billets and machined flush. The extrusion of the billets into the tube shells was done using the following parameters: All billet surfaces except the bore and also the floating mandrel were lubricated with a water soluble lubricant which was baked on at 1,300.degree. F. for 1 hour. Both ends of the tube shells were cut clean and the inner diameter was honed to remove possible surface flaws and to improve the finish. Final dimensions for the tube shells were: The final reduction of the tube shells to fuel tubing followed the standard procedure which includes four reductions with cleaning and annealing between each step. The parameters for this process are listed in Table 1. TABLE 1 __________________________________________________________________________ CO-EXTRUDED TUBE REDUCTION PARAMETERS Inner Diameter Outer Thickness of Metal Barrier % Step Diameter Composite Insert Tube Reduction Qe* __________________________________________________________________________ Start with Tube Shell 2.500 .430 1.650 -- -- Clean for anneal (degrease - soap base caustic) Anneal - 1250.degree. F - 1 Hour First Pass 1.687 .270 1.147 57 1.2 Clean for anneal Anneal 1150.degree. F - 1 Hour Second Pass 1.125 .160 .805 60 1.4 Clean for anneal Anneal 1150.degree. F - 1 Hour Third Pass .750 .085 .580 64 1.7 Clean for anneal Anneal 1150.degree. F - 1 Hour Fourth Pass .495 .028 .439 70 2.3 Clean for anneal Anneal 1070.degree. F - 21/2 to 4 Hours Etch to .494 .028 .438 __________________________________________________________________________ *Qe is defined as the ratio of percentage of change in wall thickness to percentage of change in mean diameter. Dimensions of the final products are listed in Table 2. TABLE 2 ______________________________________ Dimensions in Mils of Inner Outer Metal Inner Diameter Diameter Barrier Layer ______________________________________ Example 1 0.438 0.494 1.0 .+-. .2 3.1 .+-. 0.6 Example 2 0.438 0.494 1.0 .+-. .2 3.2 .+-. 0.5 Example 3 0.438 0.494 1.4 .+-. .2 3.6 .+-. 0.2 Example 4 0.438 0.494 1.0 .+-. .1 3.0 .+-. 0.2 ______________________________________ As will be apparent to those skilled in the art, various modifications and changes may be made in the invention described herein. It is accordingly the intention that the invention be construed in the broadest manner within the spirit and scope as set forth in the accompanying claims. |
abstract | A device for sealing a tube in an opening which is located in a component. The tube is in particular a lance shaft and the component is in particular a nozzle connected to a cover of a reactor pressure vessel. A pressure element is in operative connection with the tube. As a result, the tube is pressed against the component. Provision is made for the pressure element to be driven hydraulically. |
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summary | ||
054024544 | summary | CROSS-REFERENCE TO RELATED APPLICATION This application is a Continuation of International Application Ser. No. PCT/DE92/00679, filed Aug. 13, 1992. BACKGROUND OF THE INVENTION FIELD OF THE INVENTION The invention relates to a process and a device for obtaining a sample from the atmosphere in a closed gastight vessel, preferably from the reactor safety vessel of a nuclear power station, wherein the sample is introduced into a sample-taking container and constituents of the sample which are soluble and/or condensable in a transport or vehicle fluid are discharged from the vessel together with the transport fluid and gaseous constituents of the sample. Such a process and a device suitable for carrying out the process are known from German Published, Non-Prosecuted Application DE 39 32 712 A1, corresponding to Published European Application No. 0 419 994 A1 and to U.S. application Ser. No. 07/590,151, filed Sep. 28, 1990, now abandoned. Gastight vessels are frequently used for enclosing installations in which substances that must not pass into the environment of the vessel are reacted. Such vessels, which are also known as containments, are normally accessible while the installation enclosed in them is working correctly, and at the same time they also permit problem-free monitoring of the atmosphere contained in them. However, in the event of faults, the vessels are hermetically sealed in accordance with regulations, so that difficulties are encountered in obtaining samples from the outside which are representative of the atmosphere. When taking samples, the different states of the atmosphere, such as "dry" and "moist", as well as the behavior of the substances in gas or vapor form and of air-borne solid and liquid aerosols are of considerable importance. In particular, effects such as depositions of condensing vapors or large aerosols (>1 .mu.m) upstream (in the direction of flow) of a sample taking device may result in a highly erroneous assessment of the composition and of a radioactive contamination of the atmosphere. In the device referred to above, sample taking fittings which are connected serially are provided in the run of a pipe loop, with each of them being able to be operated by means of a pneumatic or hydraulic line. Two penetrations of the wall of the vessel are then needed for the pipe loop in addition to one such penetration for each pneumatic or hydraulic line. Since the strength and tightness of the vessel must not be impaired thereby, it entails an expense which is not inconsiderable. Furthermore, the mechanically active sample taking fittings must also remain capable of operating under and after accident conditions, such as with temperatures above 500.degree. C. and extreme radiation loads of over 10 KGy/h, so that stringent demands on materials, particularly for moving parts, have to be met. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a process and a device for obtaining samples from the atmosphere in a closed gastight vessel, preferably from the reactor safety vessel of a nuclear power station, which overcome the hereinafore-mentioned disadvantages of the heretofore-known methods and devices of this general type and which obtain representative samples of the atmosphere prevailing in the vessel from the outside, without impairing the strength and tightness of the latter, while depositions of constituents of the samples, which would falsify the results of the measurement, are reliably avoided and the taking of samples is effected with components which can also be constructed for faulty conditions and which are preferably mechanically passive. With the foregoing and other objects in view there is provided, in accordance with the invention, a process for obtaining a sample from an atmosphere in a closed gastight vessel, preferably from a reactor safety vessel of a nuclear power station,which comprises passing a sample through a venturi nozzle immediately upon entry of the sample into a sample-taking container in a vessel; mixing the sample in the venturi nozzle with a transport fluid serving as a washing liquid; and subsequently discharging gaseous constituents of the sample being soluble and/or condensable in the washing liquid together with the washing liquid from the sample-taking container and from the vessel by triggering by a sudden pressure reduction. In accordance with another mode of the invention, an inlet channel, through which the sample flows before reaching the venturi nozzle, is flushed with the washing liquid contained in the sample-taking container, before the sample is discharged. In accordance with a further mode of the invention, the velocity of flow of the sample in the venturi nozzle or nozzles is slightly, preferably 10% to 30%, below the critical nozzle velocity, so long as no condensation of the sample occurs in the washing liquid, and the velocity of flow is increased to the critical nozzle velocity as soon as the sample condenses, at least partially, in the washing liquid. In accordance with an added mode of the invention, there is provided a process which comprises reacting gaseous constituents of the sample and of the transport fluid chemically with one another. In accordance with an additional mode of the invention, there is provided a process which comprises varying a level height of the washing liquid, preferably in the inlet channel, by pressure changes in the transport fluid, and raising the washing liquid at least once after the sample has flowed in to a height of an inlet opening at a free end of the inlet channel for the sample. In accordance with yet another mode of the invention, there is provided a process which comprises adjusting a difference in pressure between an atmosphere in the vessel and an interior of the sample-taking container to up to 5000 hPa. In accordance with yet a further mode of the invention, there is provided a process which comprises adjusting a temperature of the washing liquid at the beginning of a sample taking process to be slightly lower than that of an atmosphere in the vessel. In accordance with yet an added mode of the invention, there is provided a process which comprises retaining elementary organic iodine, CO, CO.sub.2 and other gas from the sample in the washing liquid, by inactive iodine additions and variation of the pH value of the washing liquid. In accordance with yet an additional mode of the invention, there is provided a process which comprises diluting the washing liquid after being drawn off by suction from the sample-taking container, until the radioactivity of the sample is lower than 10.sup.9 Bq/m.sup.2. In accordance with again another mode of the invention, there is provided a process which comprises separating the sample before its assessment into gaseous constituents and washing liquid containing other parts of the sample, and drawing off the sample by suction through a throttle working in the laval velocity range, through a water separator and into a vacuum vessel. In accordance with again a further mode of the invention, there is provided a process which comprises distributing each individual sample over a plurality of transport containers for transport purposes. In accordance with again an added mode of the invention, there is provided a process which comprises subjecting the sample-taking container to superatmospheric pressure, such as by introducing nitrogen, until a bursting disk at a free end of the inlet channel breaks, to initiate obtaining a sample. With the objects of the invention in view, there is also provided a device for obtaining samples from an atmosphere in a closed gastight vessel, preferably from a reactor safety vessel of a nuclear power station, comprising a sample-taking container having a bottom and a given volume; a washing liquid being disposed in the sample-taking container and having a volume being at most approximately equal to half of the given volume; a venturi nozzle dipping into the washing liquid in the sample-taking container above the bottom; and an inlet channel leading into the sample-taking container below the venturi nozzle. In accordance with another feature of the invention, the volume of the washing liquid is slightly greater than the volume of an inlet channel, serving for the admission of the samples, between the free end of the channel and the bottom of the sample-taking container. In accordance with a further feature of the invention, the venturi nozzle is replaced by filling bodies, which serve as flow distributors, and a plurality of nozzles in the bottom of the sample-taking container, with an inlet opening of the inlet channel at the free end of the latter being closed by a bursting disk during the normal use of the vessel. In accordance with an added feature of the invention, the sample-taking container has a dome, and there is provided a filling and emptying line at the bottom for the washing liquid, and a gas line connected to the dome. In accordance with an additional feature of the invention, the sample-taking container has a side, the filling and emptying line is guided upwards at the side, and there is provided an injector connecting the filling and emptying line to the gas line laterally above the dome. In accordance with yet another feature of the invention, the sample-taking container is disposed inside the vessel, preferably inside a reactor safety vessel. In accordance with yet a further feature of the invention, there is provided a line starting from the injector and passing to the outside through an outer wall of the vessel, and a throttle in the line for limiting a flow through the line. In accordance with yet an added feature of the invention, there is provided a sorption filter for organoiodine being inserted into the line. In accordance with yet an additional feature of the invention, there are provided polished or teflon-coated surfaces, preferably in the inlet channel, being in contact with the sample. In accordance with again another feature of the invention, the sample-taking container has built-in fittings, and the inlet channel, the sample-taking container with all of the built-in fittings, the filling and emptying line, the gas line, the injector and the line starting from the injector are formed essentially of radiation-resistant material, such as special steel. In accordance with again a further feature of the invention, volume of the washing liquid filling the sample-taking container is substantially between 2 and 3 liters. In accordance with a concomitant feature of the invention, there are provided means for maintaining a constant velocity of flow of the sample by throttling in the venturi nozzle and/or with a throttle disposed outside the vessel. The process according to the invention and the device according to the invention are very advantageous because they make it possible to obtain unadulterated samples of the atmosphere in a hermetically sealed vessel and for this purpose require only a single bushing for a pipe through the wall of the vessel. Impairment of the strength and tightness of the vessel are thereby virtually impossible. This is still true if, in the event of the direct extraction of washing liquid, a second pipe is passed through the wall of the vessel. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a process and a device for obtaining samples from the atmosphere in a closed gastight vessel, preferably from the reactor safety vessel of a nuclear power station, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
043084605 | abstract | A radioactive material storage system for use in the laboratory having a flat base plate with a groove in one surface thereof and a hollow pedestal extending perpendicularly away from the other surface thereof, a sealing gasket in the groove, a cover having a filter therein and an outwardly extending flange which fits over the plate, the groove and the gasket, and a clamp for maintaining the cover and the plate sealed together, whereby the plate and the cover and the clamp cooperate to provide a storage area for radioactive material readily accessible for use or inventory. Wall mounts are provided to prevent accidental formation of critical masses during storage. |
claims | 1. A method for radiographic density evaluation of a material, said method comprising:capturing a radiographic image of the material with a cassette by projecting X-rays from a source through the material, the cassette and/or software is configured to obtain information to perform intensity normalization and standardization of the radiographic image;performing intensity standardization of the radiographic image;analyzing the radiographic image to evaluate radiographic tissue density of the biological material; andwherein the intensity standardization is performed comprises estimating a background, the background created by an instrument source-detector geometry and baseline responses; and subtracting the background from the radiographic image, and converting the radiographic image from a grayscale image into color-coded intensity images that form an intuitive colormap based on a calibration bar with a predetermined radiographic signature on the cassette to serve as reference for performing the intensity standardization. 2. The method of claim 1 wherein the source is an X-ray radiography machine. 3. The method of claim 1 wherein the cassette is placed under the subject relative to a direction of the source. 4. The method of claim 1 further comprising diagnosing a disease or condition in the biological material when the radiographic tissue density is outside of a preselected threshold. 5. The method of claim 1 wherein the cassette is configured for use with the source producing X-ray exposures of varying kVp, mAs and time. 6. The method of claim 1 wherein the cassette further comprises a radio-opaque backing having an X-ray radiographic signature and estimating a source-detector geometrical inhomogeneity, and placing at least one calibration bar or other device with a predetermined radiographic signature on the cassette to serve as references for performing the intensity standardization. 7. The method of claim 1 further comprising placing at least one calibration bar with a predetermined radiographic signature on or within the cassette, the at least one calibration bar configured to have a graduated radio-opacity inset. 8. The method of claim 1 wherein the material is a living human subject. 9. The method of claim 1 wherein the material is a biological tissue or organ or organelle. 10. The method of claim 1 wherein the material is a non-biological. 11. The method of claim 1 wherein the source is an X-ray microscope. |
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description | This application claims priority from Korean Patent Application No. 10-2011-0141175, filed on Dec. 23, 2011, in the Korean Intellectual Property Office, the contents of which are incorporated herein by reference in its entirety. 1. Field of the Invention The present invention relates to a nuclear fuel rod for a fast reactor, and more particularly, to a nuclear fuel rod for a fast reactor in which a reactor core of a sodium-cooled fast reactor (SFR) is designed compact-sized. 2. Description of the Related Art The fast reactor is generally known for pyroprocessing in which nuclear waste from the light water reactor is recycled. The fast reactor, and more specifically, the sodium-cooled fast reactor (SFR) continuously breeds more fuel for fission in the reactor than it consumes. Further, the nuclear waste from the light water reactors has relatively long half life and nucleus with high radiotoxicity is split into more stable nuclei with reduced radiotoxicity by the transmutation in SFR. With high utilization of the fuel, sustained energy supply can be provided while waste is reduced by SFR. FIG. 5 illustrates a constitution of a conventional nuclear fuel rod for a fast reactor. Referring to FIG. 5, the conventional nuclear fuel rod for the fast reactor includes a slender cylindrical fuel material 10, and a cladding tube 20 in a tubular form surrounding the material 10. The fast reactor is designed to be cooled by liquid metal such as sodium (Na), lead (Pb), lead-bismuth (Pb—Bi) alloy instead of water which is used as a coolant in the light- or heavy-water reactor. It is also possible to cool the heat generated from the slug 10 by feeding liquid metal 30 also into the cladding tube 20. Meanwhile, in the conventional nuclear fuel rod for the fast reactor, plenum, which is the space to collect fission products (specifically, fission gas), is placed above the materials, taking up approximately 42% of the entire length of the nuclear fuel rod. The plenum is considered to be one of the important factors that influence inner pressure of the fuel rod due to release of fission products (specifically, fission gas) according to the consumption of the nuclear fuel. In determining the overall length of the nuclear fuel rod and determining height of the reactor core, it is necessary to reduce the overall length of the nuclear fuel rod by reducing the length of the plenum for the purpose of compacter reactor. By way of example, Korean Patent Application No. 10-2007-0082795 (Korean Patent No. 1009156020000) (title: Fuel rod coated with oxide film on inner surface of the cladding and its manufacturing method) proposes a technology to extend the lifespan of the fast reactor by forming an oxide film on an inner surface of the cladding tube of the nuclear fuel rod for fast reactor, thereby suppressing fuel cladding chemical interaction (FCCI) between metallic fuel slug and cladding tube and increasing maximum allowable degree of burnup and maximum allowable temperature. FIG. 6 illustrates another constitution of a conventional nuclear fuel rod for fast reactor. Referring to FIG. 6, the nuclear fuel rod for fast reactor of Korean Patent No. 1009156020000 has a film of oxide selected from a group consisting of chromium oxide (Cr2O3), Vanadium oxide (V2O3) and zirconium oxide (ZrO2) which is coated on an inner surface of the cladding tube, in which the oxide film suppresses chemical reaction between the metallic slug and the cladding tube to thus increase temperature at the outlet of the fast reactor. However, the above-mentioned patent inherently has a problem of difficulty of designing compact-sized reactor core, due to the metallic slug that is structurally identical to the conventional fuel slug. Korean Patent Application No. 10-2008-0011094 (Korean Patent No. 10-0959152) (title: Metallic fuel element with nitride-coated layer on cladding inner surface for fast nuclear reactor and manufacturing method thereof) discloses a technology to lift up the operational temperature limit by efficiently suppressing inter-diffusion reaction between metallic fuel slug received in the cladding tube and the cladding tube by stably forming nitride-coated layer of the transition elements on inner wall of the cladding tube, thereby preventing generation of inter-reaction layer between the metallic fuel slug and the inner wall of the cladding tube. FIG. 7 illustrates a constitution of yet another conventional nuclear fuel rod for fast reactor. Referring to FIG. 7, the metallic nuclear fuel for use in fast reactor includes a metallic fuel slug 1 of cylindrical configuration, a hollow cylindrical cladding tube 3 surrounding the metallic fuel slug 1, a nitride-coated layer 2 of transition element formed on inner wall of the cladding tube 3, and liquid metal 4 filled between the metallic fuel slug 1 and the cladding tube 3. While the conventional constitution aims to lift up the operational temperature limit by preventing generation of the inter-reacting layer between the metallic fuel slug 1 and the inner wall of the cladding tube 3 due to the nitride-coated layer 2, this still does not address the problem associated with the difficulty of forming compact-sized reactor core since the metallic fuel slug 1 is formed structurally identical to the conventional fuel slug. Korean Pat No. 10-0963472 mentions metallic fuel rod including in the cladding tube a metal sheath having therein metallic fuel particles formed by atomizing technique, and a preparation method thereof. However, Korean Pat No. 10-0963472 does not describe about ensuring plenum space by defining a hollow space in the center occupied by the fuel material. That is, the patent only inserts metal sheath into cladding tube to prevent interaction between the fuel material and the cladding tube. Accordingly, in order to solve the problems occurring in the prior art, the inventors have researched for a method for designing a compact-sized reactor core by structurally altering a nuclear fuel rod for a fast reactor and thus completed the present invention. An aspect of the present invention provides a nuclear fuel rod in which a reactor core of the fast reactor can be compact-sized by reducing the length of the nuclear fuel rod to be smaller than the length of the conventional one. In order to achieve the aspect of the invention, the present invention provides a nuclear fuel rod for a fast reactor which may include tubular fuel (that is, fuel slug or fuel particles) having a hollow portion therein, a tubular inner pipe inserted into the hollow portion of the tubular fuel materials to prevent collapse of the tubular fuel rod due to combustion of the nuclear fuel, and a liquid metal, or He gas or vacuum applied in gaps between a tubular cladding pipe surrounding the tubular fuel materials and the tubular fuel materials and the tubular cladding pipe, in which the tubular inner pipe collects nuclear fission products such as fission gas therein which is generated due to combustion of the nuclear fuel. Further, the nuclear fuel rod for the fast reactor according to an embodiment may additionally include a plenum provided on an upper end of the tubular fuel rod to collect nuclear fission products such as fission gas which is generated due to combustion of the nuclear fuel. Further, the tubular fuel materials may include at least one element selected from a group consisting of uranium (U), plutonium (Pu), zirconium (Zr), americium (Am), neptunium (Np), and curium (Cm). The cross-section fraction occupied by the tubular fuel materials may take up 50% to 90% of the total cross section of the fuel rod. An outer circumference of the tubular inner pipe and an inner circumference of the tubular fuel materials may be brought into close contact with each other so that the tubular inner pipe supports the tubular fuel materials. The tubular inner pipe may be formed from molybdenum (Mo), tungsten (W), niobium (Nb), tantalum (Ta), or alloy thereof. The tubular cladding pipe may be formed from stainless steel. The liquid metal may be sodium, but may be He gas or vacuum depending on embodiments. According to the nuclear fuel rod for a fast reactor in one embodiment, reactor core of the fast reactor can be compact-sized by reducing the length of the nuclear fuel rod from the length of the conventional one. Further, in one embodiment, operation margin can be increased by reducing the core temperature of the nuclear fuel rod. Further, in one embodiment, safety is ensured in the event of design basis accident (DBA) where fuel materials melting occurs. Meanwhile, in one embodiment, the tubular fuel materials may be formed by casting metal fuel material. To be specific, the tubular fuel materials may be formed into metal fuel particles by spraying, and vibro-packed into the tubular fuel materials. Reference will now be made in detail to the embodiments of the present invention, examples of which are illustrated in the accompanying drawings, wherein like reference numerals refer to the like elements throughout. The embodiments are described below to explain the present invention by referring to the figures. The invention relates to a nuclear fuel rod for a fast reactor in which a reactor core of the fast reactor, or a reactor core of a sodium-cooled fast reactor (SFR) to be more specific, is designed compact-sized by reducing the length of the nuclear fuel rod to be smaller than the length of the conventional nuclear fuel rod. FIG. 1 illustrates a constitution of a nuclear fuel rod for a fast reactor according to an embodiment. Referring to FIG. 1, the nuclear fuel rod for a fast reactor according to an embodiment may include a tubular fuel materials 100, a tubular inner pipe 200, a tubular cladding pipe 300 and a liquid metal 400 (or He gas or vacuum depending on embodiments), which are assembled in the form of a nuclear fuel assembly at predetermined intervals from each other and mounted to the interior of the fast reactor. Although not illustrated, the nuclear fuel rod for the fast reactor according to an embodiment may additionally include a plenum which is provided on an upper end of the tubular fuel materials 100 to collect fission gas which is generated due to combustion of the nuclear fuel. FIG. 2 illustrates a constitution of the tubular fuel materials of the nuclear fuel rod for the fast reactor according to an embodiment. Referring to FIG. 2, the tubular fuel materials 100 may include a hollow portion 110 formed therein. Generally, as illustrated in FIG. 5, the fuel materials provided in the conventional nuclear fuel rod for the fast reactor have an elongated, slender cylindrical shape which does not have a hollow portion therein. However, in one embodiment, the fuel materials 100 may have the hollow portion 110 formed therein (FIG. 2). To be specific, the hollow portion 110 is provided to collect the fission products such as fission gas released in accordance with the combustion of the nuclear fuel, and in one embodiment, due to the presence of the hollow portion 110 defined in the fuel materials 100, the length of the plenum, playing the same role as the hollow portion 110, may be reduced, and as a result, the overall length of the nuclear fuel rod may be reduced. That is, instead of reducing the length of the plenum, which works as the space to collect the fission products such as fission gas, an embodiment additionally forms the hollow portion 110 in the fuel materials 100 to play the same role as the plenum, thereby reducing the length of the nuclear fuel rod and thus enable designing of a compact-sized reactor core of the fast reactor. The tubular fuel materials 100 include fission fuel therein, and may be sealed off by a receptacle such as the tubular cladding pipe 300 which has good compatibility with the liquid metal 400 and no reactivity, and excellent heat conductivity. The space filled with liquid metal may alternatively be filled with He gas or vacuum as need arises. To be specific, the tubular fuel materials 100 may include at least one element selected from a group consisting of uranium (U), plutonium (Pu), zirconium (Zr), americium (Am), neptunium (Np), and curium (Cm). As explained above, in one embodiment, the output from the fast reactor may decrease from that of the conventional fuel materials due to the presence of the hollow portion 110 formed in the fuel materials 100 and subsequent decrease in the amount of fission products such as fission gas. However, this decrease in the output may be compensated by increasing the diameter of the fuel materials. Meanwhile, the cross section fraction occupied by the fuel materials in the fuel rod may control the heat generated during the nuclear fission. Accordingly, if the area is too small, heat output will deteriorate, while if the area is too large, mechanical interaction with the cladding pipe may increase due to expansion of the fuel materials following the nuclear fission release. In consideration of the above, the cross section fraction occupied by the tubular fuel materials 100 may take up 50% to 90%, and preferably, take up approximately 75% of the total cross section. Referring to FIG. 1, the tubular inner pipe 200 may be inserted into the hollow portion 110 of the tubular fuel materials 100 to prevent collapse of the tubular fuel materials 100 due to combustion of the nuclear fuel. To be specific, since an outer circumference 200a of the tubular inner pipe 200 and an inner circumference 100a of the tubular fuel materials 100 are brought into close contact with each other, the tubular inner pipe 200 support the tubular fuel materials 100 and therefore, collapse of the tubular fuel materials 100 due to combustion of the nuclear fuel can be prevented. FIG. 3 illustrates a constitution of the tubular inner pipe of the nuclear fuel rod for a fast reactor according to an embodiment. Referring to FIG. 3, the tubular inner pipe 200 may include a collecting space 210 therein to collect fission products such as fission gas generated due to the combustion of nuclear fuel. To be specific, the collecting space 210 may be formed in the inner pipe 200 and like the plenum, may collect the fission gas which is generated due to the combustion of the nuclear fuel. Further, the tubular inner pipe 200 may be formed from molybdenum (Mo), tungsten (W), niobium (Nb), tantalum (Ta), or alloy thereof. Referring to FIG. 1, the tubular cladding pipe 300 may cover the tubular fuel materials 100 from outside. FIG. 4 illustrates a constitution of the tubular fuel materials 100 of a nuclear fuel rod for a fast reactor according to an embodiment. Referring to FIG. 4, the tubular cladding pipe 300 may be formed into a tubular configuration having a hole formed therein, and referring to FIG. 1, may cover the tubular fuel materials 100 from outside as the tubular fuel materials 100 is placed in the hole of the tubular cladding pipe 300. Further, the tubular cladding pipe 300 may be formed from stainless steel to prevent leakage of the products of the nuclear fission, while also preventing chemical interaction between the nuclear fission materials and the coolant placed outside the nuclear fuel rod by blocking direct contact thereof. To be specific, the tubular cladding pipe 300 may be formed from high chrome stainless steel containing 8 to 12 wt % chrome, and may consist of element such as iron, chrome, tungsten, molybdenum, vanadium, or niobium. Referring to FIG. 1, the liquid metal (or He gas or vacuum depending on embodiments) 400 may be filled in gap between the tubular fuel materials 100 and the tubular cladding pipe 300, and may be sodium. To be specific, gaps may be formed between the tubular fuel materials 100 and the tubular cladding pipe 300 to accommodate the thermal expansion of the tubular fuel materials 100 during combustion, and by charging liquid sodium with good thermal conductivity in this gap between the tubular fuel materials 100 and the tubular cladding pipe 300, the thermal conductivity may be enhanced between the tubular fuel materials 100 and the tubular cladding pipe 300. That is, by releasing the heat generated due to nuclear reaction of the tubular fuel materials 100 during combustion of the nuclear fuel using the liquid metal (or He gas or vacuum depending on embodiments) 400 to outside the tubular cladding pipe 300, the core temperature of the tubular fuel materials 100 can be decreased. As explained above, in the nuclear fuel rod for the fast reactor according to an embodiment, since the length of the nuclear fuel rod can be reduced from the length of the conventional one, the reactor core of the fast reactor can be designed to be compact-sized, and since the core temperature of the nuclear fuel rod is decreased, the operation margin can be increased, and safety is ensured in the event of the design basis accident (DBA) due to fuel materials melting. While the present invention has been particularly shown and described with reference to exemplary embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the spirit and scope of the invention as defined by the appended claims. |
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description | The present invention relates generally to a medical cancer therapy facility and, more particularly, to a medical particle delivery system having a compact gantry design. It has been known in the art to use a particle accelerator, such as a synchrotron, and a gantry arrangement to deliver a beam of particles, such as protons, from a single source to one of a plurality of patient treatment stations for cancer therapy. In such systems, the cancerous tumor will be hit and destroyed by particles in a precise way with a localized energy deposition. Thus, the number of ion interactions on the way to the tumor through the healthy body cells is dramatically smaller than by any other radiation method. A position of the center of the tumor inside the body defines a value of the particle energy. The transverse beam raster is defined by the transverse size of the tumor with respect to the beam, while the width of the tumor defines the beam energy range. The energy deposition is localized around the “Brag” peak of the “implanted particles” and remaining energy is lost due to particle interaction with the tumor cells. Such cancer treatment facilities are widely known throughout the world. For example, U.S. Pat. No. 4,870,287 to Cole et al. discloses a multi-station proton beam therapy system for selectively generating and transporting proton beams from a single proton source and accelerator to one of a plurality of patient treatment stations each having a rotatable gantry for delivering the proton beams at different angles to the patients. A duoplasmatron ion source generates the protons which are then injected into an accelerator at 1.7 MeV. The accelerator is a synchrotron containing ring dipoles, zero-gradient dipoles with edge focusing, vertical trim dipoles, horizontal trim dipoles, trim quadrupoles and extraction Lambertson magnets. The beam delivery portion of the Cole et al. system includes a switchyard and gantry arrangement. The switchyard utilizes switching magnets that selectively direct the proton beam to the desired patient treatment station. Each patient treatment station includes a gantry having an arrangement of bending dipole magnets and focusing quadrupole magnets. The gantry is fully rotatable about a given axis so that the proton beam may be delivered at any desired angle to the patient. The gantry of typical particle beam cancer therapy systems accepts a particle beam of a required energy from the accelerator and projects it with a high precision toward a cancerous tumor within a patient. The beam from the gantry must be angularly adjustable so that the beam can be directed into the patient from above and all sides. Because of these requirements, the gantry of a conventional particle beam cancer therapy facility is typically the most expensive piece of equipment of the treatment facility and its magnets are generally very large and heavy. For example, the proton-carbon medical therapy facility described by R. Fuchs and P. Emde in “The Heavy Ion Gantry of the HICAT Facility” includes an isocentric gantry system for delivery of protons, Helium, Carbon and Oxygen ions to patients. The gantry system has a total weight of 630 tons and the required beam line elements for transporting and delivering fully stripped Carbon and Oxygen ions with 430 MeV/nucleon kinetic energy have a total weight of 135 tons. The rotating part of the isocentric gantry system weighs about 570 tons due to its role to safely transport and precisely delivers ions to the patients. Advances in particle accelerator design have resulted in accelerators utilizing smaller and less complex magnet arrangements. For example, a nonscaling fixed field alternating gradient (FFAG) accelerator has recently been developed which utilizes fixed field magnets, as opposed to much larger and more complex variable magnetic field coil magnets. Such advances, however, have heretofore not been applied to the gantry design of typical cancer therapy facilities. Accordingly, it would be desirable to improve upon the prior art medical cancer therapy facilities by providing a simpler, less expensive and more compact gantry design utilizing some of the advances made in the field of particle accelerators. The present invention is a particle therapy gantry for delivering a particle beam to a patient. The gantry generally includes a beam tube having a curvature defining a particle beam path and a plurality of fixed field magnets sequentially arranged along the beam tube for guiding the particle beam along the particle path. In a preferred embodiment, each of the fixed field magnets is a combined function magnet performing a first function of bending the particle beam along the particle path and a second function of focusing or defocusing the particle beam. Also, the magnets are preferably arranged in triplets, wherein each triplet has two focusing magnets and one defocusing magnet disposed between the focusing magnets. The focusing magnets perform the combined function of bending the particle beam and focusing the particle beam and the defocusing magnet performs the combined function of bending the particle beam and defocusing the particle beam. The defocusing magnets are preferably positive bending magnets for bending the particle beam along an arc defined by a positive center of curvature and the focusing magnets are preferably negative bending magnets for bending the particle beam along an arc defined by a negative center of curvature, wherein the positive and negative centers of curvature are oriented on opposite sides of the beam pipe. In one embodiment, the fixed field magnets are permanent magnets including a ferromagnetic core having a curvature defined by a center of curvature and forming a beam tube receiving cavity having the beam tube supported therein. The core is shaped to provide a magnetic field in the beam tube which grows stronger in a direction toward the core center of curvature. In an alternative embodiment, the fixed field magnets include superconducting coils adjacent the beam tube for providing the combined function. In either case, the beam tube of the gantry preferably includes a particle beam entry point, a transition point, a particle beam exit point, a first curved particle beam path arc length extending between the entry point and the transition point and a second curved particle beam path arc length extending between the transition point and the exit point. The first arc length bends about ninety degrees and the second arc length bends about one hundred eighty degrees in a direction opposite the first arc length. Two half-triplets are preferably disposed in juxtaposed orientation at the beam tube transition point and a half-triplet is preferably disposed at each of the beam tube entry point and the beam tube exit point. Each of the half-triplets includes a defocusing magnet and a focusing magnet. The present invention further involves a method for delivering a particle beam to a patient through a gantry. The method generally includes the steps of bending the particle beam with a plurality of fixed field magnets sequentially arranged along a beam tube of the gantry, wherein the particle beam travels in the beam tube, and alternately focusing and defocusing the particle beam traveling in the beam tube with alternately arranged combined function focusing and defocusing fixed field magnets. In a preferred embodiment, the combined function fixed field magnets are arranged in triplets, wherein each triplet includes two focusing magnets and one defocusing magnet disposed between the focusing magnets. The focusing magnets perform the combined function of bending the particle beam and focusing the particle beam and the defocusing magnet performs the combined function of bending the particle beam and defocusing the particle beam. The defocusing magnets are preferably positive bending magnets for bending the particle beam along an arc defined by a positive center of curvature and the focusing magnets are preferably negative bending magnets for bending the particle beam along an arc defined by a negative center of curvature, wherein the positive and negative centers of curvature are oriented on opposite sides of the beam pipe. The gantry of the present invention may be utilized in a medical particle beam therapy system having a source of particles, a particle accelerator, an injector for transporting particles from the source to the accelerator, one or more patient treatment stations including rotatable gantries of the present invention for delivering a particle beam to a patient and a beam transport system for transporting the accelerated beam from the accelerator to the patient treatment station. The preferred embodiments of the particle beam gantry of the present invention, as well as other objects, features and advantages of this invention, will be apparent from the following detailed description, which is to be read in conjunction with the accompanying drawings. The scope of the invention will be pointed out in the claims. FIG. 1 shows a typical medical particle delivery therapy facility 10. The facility 10 generally includes an injector 12, a particle accelerator 14, and a beam delivery network 16 including a rotatable gantry treatment room 18 for delivering a beam to a patient. The beam delivery network 16 may also be designed to divert independent beams to various other applications as desired. For example, the beam delivery network 16 may be designed to deliver a beam to a beam research room 20 and a fixed beam treatment room 22. The research room 20 may be provided for research and calibration purposes, with an entrance separate from the patient areas, while the fixed beam treatment room 22 may include separate beam lines for such therapeutic applications, such as eye treatments. The beam injector module 12 can be a conventional LINAC or a tandem Van de Graaf injector with an injection kicker, which completes the task of particle injection into the accelerator 14. In the case of proton particles, the injector typically provides proton beam pulses at 30 Hz with a pulse width varying between 25 and 100 nanoseconds at a delivered energy of 7 MeV. The particle accelerator 14 can be a synchrotron, cyclotron or some other conventional design known in the prior art. The accelerator 14 accelerates particles to a desired energy level for extraction and delivery to the patient treatment rooms 18 and 22. Variation of the extraction energy is achieved by adjusting, for example, an RF frequency within the accelerator 14. Again for proton particles, extraction typically occurs when the kinetic energy of the particles is in the range 60 to 250 MeV. The beam delivery network 16 connects the accelerator 14 to the treatment rooms 18 and 22 and the beam research room 20. The network 16 generally includes an extraction line 26, a switchyard 28 and a plurality of beam transport lines 30. The switchyard 28 is typically an arrangement of switching magnets for diverting the particle beam to a desired beam line 30. The beam transport lines 30 take the particle beam from the switchyard 28 to the different treatment rooms of the facility. Referring additionally to FIG. 2, the rotatable gantry treatment room 18 includes a rotating gantry 24, which is rotatable by plus or minus 200 degrees from the vertical about a point of rotation 32 to deliver a particle beam to a patient 33 at a gantry iso-center 34. The gantry system accepts particles already accelerated to required energy. The first part 24a of the gantry bends particles within a quarter of a circle for 90 degrees. The second part 24b of the gantry bends the particles in a half of a circle and brings the particles straight towards the required direction 34. The gantry 24 is constructed as a three-dimensional structure supported on the treatment room side by a bearing 36 and, on the beam inlet side, by a bearing 38. The gantry 24 is further preferably balanced around its rotation axis. Gantry movement can be realized by a gear motor/gear ring drive 40 that allows high precision positioning. Each gantry 24 is preferably controlled by means of an individual independent computer unit that ensures mutual braking of the main drive units, soft start and soft deceleration functions, control of the auxiliary drive units for the treatment room, and supervision of the limit switches. The gantry 24 further includes a nozzle 42 for delivering the particle beam to the patient 33. Referring now to FIG. 3, the optical components of the gantry 24 according to the present invention are shown. The gantry 24 generally includes a hook-shaped beam pipe 44 and a series of identical fixed-field magnet triplets 46 arranged in sequence around the beam pipe. The beam pipe 44 can be provided as a continuous pipe, or it can be assembled from a plurality of beam pipe segments welded or otherwise fastened together in a conventional manner. The beam pipe 44 and the magnet triplets 46 are enclosed in a gantry housing 47. Referring additionally to FIGS. 4-6, the magnet triplet 46 is considered the “unit cell” and contains a relatively long combined function bending/defocusing magnet (QD) 48 flanked by a pair of shorter combined function bending/focusing magnets (QF) 50. The cell 46 is symmetric with respect to the center of the defocusing magnet 48. Thus, the gantry 24 is made of densely packed identical “triplet” cells 45. Three combined function magnets make a cell. The central magnet 48 produces major bending and has a linear horizontal defocusing gradient. Two smaller identical but opposite bending magnets 50 are placed on both sides of the major bending magnet 48. They have a linear focusing gradient. Each of the combined function magnets 48 and 50 performs two functions. The first function is to bend the particle beam along an orbital path, while the second function is to focus or defocus the particle beam as it travels around the path. The defocusing magnet (QD) 48 has a strong central field and a negative gradient (horizontally defocusing) at the center, while the focusing magnets (QF) 50 have a positive gradient (horizontally focusing). Both magnets 48 and 50 are fixed field dipole-type magnets using a very strong focusing and small dispersion function. The horizontal and vertical betatron functions βx and βy and the dispersion function in the basic cell 46, at the reference momentum, are shown in FIG. 7. The minimum required aperture for the two combined function magnets major bend with the defocusing gradient and the opposite bend with the focusing field are presented in FIGS. 8 and 9, respectively. Thus, the QD and QF magnets 48 and 50 are arranged in a non-scaling, fixed field alternating gradient (FFAG) configuration. Such FFAG configurations have been used before in particle accelerators, but have heretofore never been proposed in a therapeutic particle delivery gantry of a medical facility. Also, both types of magnets 48 and 50 are somewhat arc-shaped or wedge-shaped when viewed in a direction perpendicular to the path of the beam pipe 44. Thus, each magnet 48 and 50 is defined by an axis 48a and 50a, which may represent the center of curvature in the case of an arc-shaped magnet, or an intersection point of the two outside faces in the case of a wedge-shaped magnet. In either case, each defocusing magnet (QD) 48 of each magnet triplet 46 is arranged along the beam pipe 44 so that its axis 48a falls on the same side of the beam pipe 44 as the beam pipe's center of curvature 44a. Conversely, each flanking pair of focusing (QF) magnets 50 of each magnet triplet is arranged along the beam pipe 44 so that their axes 50a falls on the opposite side of the beam pipe 44 as the beam pipe's center of curvature 44a. In this manner, each defocusing magnet (QD) 48 can be termed a “positive bending” magnet, wherein the shape and arrangement of this magnet bends the particles passing therethrough in a path generally matching the curvature of the beam pipe, as shown in FIGS. 3-6. Each focusing magnet (QD) 50, on the other hand, can be termed a “negative bending” magnet, wherein the shape and arrangement of these magnets bend the particles passing therethrough in a path generally opposite to the curvature of the beam pipe. It has been found that such alternating arrangement of positive and negative bending magnets results in a particle beam having a reduced dispersion. Referring now to FIG. 10, each defocusing magnet (QD) 48 includes a ferromagnetic core 52 made up of an upper 53 and a lower half 54 forming a dipole magnet. The upper 53 and lower halves 54 are identical in cross-section and can be solid ferromagnetic masses, as shown in FIG. 10, or they can be made from a series of stacked laminates. In either case, the upper core half 53 includes an angled face 53a and the lower core half includes an angled face 54a. The angled faces 53a and 54a of the upper and lower core halves 53 and 54 face each other and form a beam pipe receiving cavity 56 when the core halves are assembled together to form the magnet core 52. Referring to FIG. 11, each focusing magnet (QF) 48 is similarly constructed. Specifically, each focusing magnet 50 includes a ferromagnetic core 58 made up of an upper 59 and a lower half 60 forming a dipole magnet. Again, the upper 59 and lower halves 60 can be solid ferromagnetic masses or they can be made from a series of stacked laminates 55. Also, the upper core half 59 includes an angled face 59a and the lower core half includes an angled face 60a. The angled faces 59a and 60a of the upper and lower core halves 59 and 60 face each other and form a beam pipe receiving cavity 62 when the core halves are assembled together to form the magnet core 58. As mentioned above, each magnet 48 and 50 is a combined function arc magnet combining the functions of bending the particle beam and focusing or defocusing the particle beam. The bending function is achieved by the curvature of the magnet, while the focusing or defocusing function is achieved by the arrangement of the magnet cores 52, 58. In particular, the upper 53 and the lower 54 halves of the defocusing magnet core 52 are arranged together respectively above and below the beam pipe 44 so as to provide a magnetic field in the beam pipe which grows stronger in a direction toward the center of curvature 48a of the core, as shown in FIG. 10, whereas the upper and the lower halves 59 and 60 of a focusing magnet core 58 are arranged together respectively above and below the beam pipe so as to provide a magnetic field in the beam pipe which grows weaker in a direction toward the center of curvature of the defocusing core 48a, but which grows stronger in a direction toward the center of curvature 50a of its own core. Thus, in a defocusing combined function magnet 48, as shown in FIG. 10, a proton, or other particle, in the beam pipe 44 radially further from the core center of curvature 48a and the beam pipe center of curvature 44a (to the right in FIG. 10) is subject to a weaker magnetic field and bends less, while a proton, or other particle, closer to the beam pipe center of curvature (to the left in FIG. 10) sees a stronger magnetic field and bends more. This results in a more dispersed horizontal concentration of protons, but a denser vertical concentration, in the beam pipe just downstream of a defocusing combined function magnet. Conversely, in a focusing combined function magnet 50, as shown in FIG. 11, a proton, or other particle, in the beam pipe 44 radially further from the beam pipe center of curvature 44a, or closer to the core center of curvature 50a, (to the right in FIG. 11) is subject to a stronger magnetic field and bends more, while a proton closer to the beam pipe center of curvature, or away from the core center of curvature, (to the left in FIG. 11) sees a weaker magnetic field and bends less. This results in a greater horizontal concentration of particles, but a weaker vertical concentration of particles in the beam pipe just downstream of a focusing combined function magnet. The above defocusing effect is achieved by orienting the angled surfaces 53a and 54a of the upper and lower core halves 53 and 54 of the defocusing magnet core 52 so that they form an intersection point 64 that falls on the same side of the beam pipe 44 as the beam pipe center of curvature 44a, as shown in FIG. 10. A focusing magnet 50 is formed by orienting the angled surfaces 59a and 60a of the upper and lower core halves 59 and 60 of the focusing magnet core 58 so that they form an intersection point 66 that falls on the side of the beam pipe 44 opposite the beam pipe center of curvature 44a, as shown in FIG. 11. In other words, the angled faces 53a and 54a of a defocusing magnet 48 meet adjacent the inner arc of the beam pipe 44, whereas the angled faces 59a and 60a of a focusing magnet 50 meet adjacent the outer arc of the beam pipe, with respect to the center of curvature 44a of the beam pipe. Accordingly, not only are the positive and negative bending functions alternately arranged, but also the focusing and defocusing functions of the magnets are alternately arranged. Such alternate arrangement of the positive and negative bending and the focusing and defocusing functions provides to the present invention the feature of net strong particle beam focusing in both horizontal and vertical planes. At the transition point 68 of the gantry 24, where the beam pipe 44 reverses its curvature, and/or at the beam entry point 70 and/or at the beam exit point 74, modifications of the magnet triplet 46 can be utilized to provide the desired bending and focusing/defocusing functions. For example, a half-triplet 76 consisting of a single negative-bend focusing magnet 50 and a reduced length, positive bend defocusing magnet 48a can be utilized at the beam entry point 76 and/or the beam exit point 72 of the gantry to achieve the desired bend angle and focusing at these points. Similarly, at the beam pipe curvature transition point 68, two half-triplets 76, as described above, can be assembled together in a juxtaposed orientation to form a “straight” magnet triplet 78. For proton therapy systems, the combined function defocusing magnet 48 and the combined function focusing magnet 50 used in the gantry can be very small permanent magnets, as described above. For example, a suitable magnetic field of about 1.8 T can be achieved using defocusing magnets 48 that measure about 6 cm×8 cm×10 cm. For larger particles, such as carbon, the magnets can utilize high-temperature superconductor tapes (HTS) or superconducting Niobium-Tin coils to achieve the required greater magnetic fields of about 6 T. In either case, the magnets are still fixed-field magnets. FIG. 12 shows a cross-section of a fixed field combined function magnet 80 utilizing high-temperature superconductor tapes (HTS) or superconducting Niobium-Tin coils 82 surrounding the beam tube 44, without an iron core. FIG. 13 shows a cross-section of a similar superconducting magnet 84 having a super ferric core 86 and superconducting coils 88 surrounding the beam tube 44. As a result of the present invention, the size of the gantry in a particle therapy facility can be dramatically reduced and the control system for a gantry treatment room can be greatly simplified. Specifically, the gantry 24 can be made about 20 meters long, from the rotation point 32 to the iso-center 34, with a height of about 3.2 meters. The gantry 24 preferably has a free space of about 1.6 meters from the last magnet to the isocenter 34. Thus, the present gantry invention reduces the weight of the gantry system by using a non-scaling Fixed Field Alternating Gradient (FFAG) triplet structure with permanent, superconducting or high-temperature superconducting combined function magnets. This invention allows a very close control of focused ion transport through the beam line with different energies but under the fixed magnetic field. The ions are delivered to the isocentric non-scaling FFAG gantry system at the same entrance position. This invention can achieve presented goals due to a very large momentum acceptance and very strong focusing properties of the non-scaling FFAG structures. The ions with different energies transported through the system arrive at the end of it with small differences in positions (−2.5 up to +3.2 mm) easily adjusted by the raster-scanning focusing part of the gantry. Although preferred embodiments of the present invention have been described herein with reference to the accompanying drawings, it is to be understood that the invention is not limited to those precise embodiments and that various other changes and modifications may be affected herein by one skilled in the art without departing from the scope or spirit of the invention, and that it is intended to claim all such changes and modifications that fall within the scope of the invention. |
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052231810 | description | DETAILED DESCRIPTION OF THE INVENTION The present invention provides a process for reducing the volume of thorium bearing radioactive waste for disposal from radioactive contaminated sites, thereby significantly reducing the cost for radioactive burial. The present process also allows for the recovery of valuable magnesium compounds for resale. The process is also economical to run on large volumes of material, using reagents that can easily be brought to the site for processing and can be recycled, and does not result in further disposal problems for the reagents or by-products from the process. Specifically, the present process extracts magnesium from magnesium slag. However, for the present invention the magnesium slag contains radioactive thorium (.sup.232 Th and .sup.230 Th) and radioactive Th daughters. The term "radioactive Th daughters" means .sup.232 Th or .sup.230 Th daughters. These radioactive Th daughters include, as .sup.232 Th daughters, actinium-228 (.sup.228 Ac), bismuth-212 (.sup.212 Bi), lead-212 (.sup.212 Pb), polonium-212 (.sup.212 Po), polonium-216 (.sup.216 Po), radium-224 (.sup.224 Ra), radium-228 (.sup.228 Ra), radon-220 (.sup.220 Rn), thallium-208 (.sup.208 Tl), and thorium-228 (.sup.228 Th) and, as .sup.230 Th daughters, astatine-218 (.sup.218 At), bismuth-210 (.sup.210 Bi), bismuth-214 (.sup.214 Bi), lead-210 (.sup.210 Pb), lead-214 (.sup.214 Pb), mercury-206 (.sup.206 Hg), polonium-210 (.sup.210 Po), polonium-214 (.sup.214 Po), polonium-218 (.sup.218 Po), radium-226 (.sup.226 Ra), radon-222 (.sup.222 Rn), thallium-206 (.sup.206 Tl), and thallium-210 (.sup.210 Tl). When the term "average .sup.232 Th daughters" is used, it is defined as ##EQU1## where [.sup.228 Ac] means the concentration of .sup.228 Ac measured using a germanium [Ge] gamma detector, and similarly for all the other indicated isotopes. Also [.sup.208 Tl] is divided by 0.35 to account for its branching ratio. Only the above indicated five isotopes given in the formula are measured for the radioactive Th daughters. The other isotopes will behave similarly and decay over time to one of the measurable isotopes. Typically, the non-radioactive components of the magnesium slag include as the major component, hydromagnesite [4 MgCO.sub.3.Mg(OH).sub.2.4 H.sub.2 O], and as minor components BaMg(CO.sub.3).sub.2 and Mg.sub.6 Al.sub.2 CO.sub.3 (OH).sub.16.4H.sub.2 O and others. Thus the starting material used in the present process termed "magnesium slag" includes both the radioactive and non-radioactive components. The magnesium slag is typically a heterogeneous mixture of the components. The process operates under pressure and uses carbon dioxide and water as the reagents. The basic processing technology is well known and was first used in the mid 1800's for separating magnesium from calcined dolomite. This process is often referred to as the Pattinson process (British Patent 9102, issued Sep. 24, 1841). Several modifications of the Pattinson process have been reported throughout the years. The selectivity in the present process needed to achieve a radioactive material volume and weight reduction is exceedingly high, as minute quantities of thorium and its daughters can cause the extracted magnesium material (for example, MgO, MgCl.sub.2, Mg metal, particularly MgCO.sub.3) to be radioactive and thus prevent its sale and pose a further disposal problem. There are ten radioactive .sup.232 Th daughters. Five of these daughters can be analyzed by gamma spectroscopy. The value obtained from the gamma spectroscopy measurement gives an estimate of the .sup.232 Th activity since the daughters for the slag that was utilized were found to be in equilibrium with the parent .sup.232 Th. Surprisingly, the Pattinson process as modified by the present invention does not result in the dissolution of radioactive thorium. The process of the present invention is highly efficient in that it uses reduced quantities of water, produces an effluent that is below regulatory concern, permits the recovery of magnesium for sale, selectively concentrates the radioactive thorium and its daughters such that the radioactivity is separated from the magnesium, and reduces the volume and weight of radioactive solids for disposal as radioactive waste. The process may be run as either a batch or continuous process. The reagents employed, water and carbon dioxide, are easily brought to a site and can be recycled in the process. To more clearly indicate the process of the present invention, the following reaction scheme is provided. In the Reaction Scheme, FIG. 1, Step I involves the digging of the crude magnesium slag (Mg Slag, crude) from the site location and separating the debris (for example, parts of trees or brush, large waste items such as tires), and grinding the crude magnesium slag to provide the refined magnesium slag. Step II of the reaction adds water to the refined magnesium slag to give a magnesium slag slurry. The ratio of the refined magnesium slag to water is such that it permits adequate mixing of the slurry (e.g. stirring). The ratio of water to magnesium slag is preferably at least about 1:1, more preferably from about 1:1 to about 10:1, most preferably about 1:1 to about 5:1, and especially preferred at about 3:1. The magnesium slag slurry is then reacted with carbon dioxide (CO.sub.2). CO.sub.2 can be introduced by sparging at atmospheric pressure (approximately 14.7 psi). However, higher yields of magnesium, as Mg(HCO.sub.3).sub.2, can be extracted if the reaction is carried out in a vessel pressurized with CO.sub.2 gas. Pressures of CO.sub.2 can be as high as 1,000 psig (about 7,000 kPa). Alternatively, the refined magnesium slag material may be heated to liberate CO.sub.2 prior to contacting the slag with water and CO.sub.2. The time of the reaction is not critical but must be sufficient so that some of the magnesium forms magnesium bicarbonate, usually from about 1 minute to about 24 hours, preferably from about 5 minutes to about 4 hours. The temperature of the reaction is not critical but appears to be most commercially suitable if it is from about -10.degree. to about 70.degree. C., with from about 4.degree. to about 35.degree. C. preferred. In Step IV the CO.sub.2 magnesium slag slurry is filtered to separate the radioactive solids (Th solids) from the Mg(HCO.sub.3).sub.2 liquor. The solids contain the radioactive .sup.232 Th and .sup.230 Th with Th daughters and processed slag. The liquor contains the soluble components, including Mg(HCO.sub.3).sub.2. The radioactive solids, which are now of a reduced volume can be treated by several processes. In Step V the radioactive solids may be disposed of in a radioactive burial site. Alternatively, the radioactive solids may be compacted by conventional means in Step VI to further reduce their volume for disposal in a radioactive burial site. Alternatively, the solids may be heated and/or compacted to further reduce their volume. The radioactive solids may be recycled in Step VII. The liquor containing the Mg(HCO.sub.3).sub.2 is radioactively below regulatory concern (i.e. the extracted magnesium is essentially void of radioactivity) and may be disposed of in Step VIII in any acceptable way. Alternatively, the liquor may be treated in Step IX by removing the CO.sub.2 by conventional methods such as by reducing the pressure, agitating, aerating, heating or combinations of these methods (CO.sub.2 may be recycled in Step XI). The resulting liquor can then be filtered to obtain MgCO.sub.3, which may be sold or converted to other products such as MgO, MgCl.sub.2, and MgSO.sub.4 in Step X. The water may optionally be recycled in Step XII. To ensure that the MgCO.sub.3 is not radioactive in Step IX, it is desirable that barium sulfate be added to precipitate the .sup.232 Th/.sup.230 Th daughters with the radioactive solids. BaSO.sub.4 may be added preferably at Step II or III, or following Step IV (the filtration step). If the addition takes place after Step IV an additional filtration step is required to remove the BaSO.sub.4 /Th-daughter coprecipitate. Alternatively, BaSO.sub.4 can be formed in situ by adding BaCl.sub.2 and Na.sub.2 SO.sub.4 to the Mg(HCO.sub.3).sub.2 liquor following Step IV. This addition must be done following the CO.sub.2 removal (by conventional methods) or after the solution has been acidified with HCl or H.sub.2 SO.sub.4. Using this scheme also requires an additional filtration step to remove the BaSO.sub.4 /Th-daughter coprecipitate. The present process utilizes carbon dioxide (CO.sub.2) and water to react with Th containing magnesium slag. The magnesium slag is placed under CO.sub.2 pressure. The CO.sub.2 reacts with the water insoluble magnesium compounds present in the slag to form Mg(HCO.sub.3).sub.2, which is soluble in the carbonated water. The Mg(HCO.sub.3).sub.2 liquor is separated from the remaining solids by filtration and the excess CO.sub.2 removed to precipitate MgCO.sub.3. With repeated extractions (preferably from 2 to 20 times), volume and weight reductions of the radioactive material for disposal as radioactive waste of at least 50%, preferably from about 50 to 90%, by weight and volume can be attained. No Th is extracted into the Mg(HCO.sub.3).sub.2 liquor, however very small amounts of the .sup.232 Th daughters are extracted. These .sup.232 Th daughters precipitate with the MgCO.sub.3 causing the resulting MgCO.sub.3 to contain radioactive isotopes and be considered radioactive waste. A small amount of barium sulfate (BaSO.sub.4) is added to cause coprecipitation with any solubilized .sup.232 Th daughters which are then removed by filtration. In the presence of excess sulfate, solubilized barium is converted into BaSO.sub.4, which is non-leachable. When the process is completed the concentration of radioactivity has increased by at least about 200%, preferably from at least about 200% to about 1,000%, from that present in the magnesium slag. The invention will be further clarified by a consideration of the following examples, which are intended to be purely exemplary of the present invention. EXAMPLE 1 A 240 G sample (approximately 325 mL) of dried, ground, radioactive magnesium slag, where the average thorium-232 (.sup.232 Th) daughter activity of the slag was 259 pCi/G (7,000 Bq/G) by gamma analysis using a germanium (Ge) detector, was added to a two liter Parr bomb reactor, followed by the addition of 1,200 mL of deionized water. The reactor was charged with 145 psig (1,100 kPa) of carbon dioxide, placed in an ice/water bath and stirred for two hours. The slurry was removed and the solids separated by vacuum filtration. To the filtrate [liquor, Mg(HCO.sub.3).sub.2 ] was added HCl (18% by weight) until the pH was between 0 and 1. Metal analysis and radioactivity measurements were performed on both the remaining solids and liquor. The remaining solids following extraction and separation were dried and weighed. The dried solids were then re-extracted by returning them to the reactor, adding 1,200 mL of deionized water, and recharging the reactor with CO.sub.2. The process was repeated ten more times (a total of 12 extractions). Following 12 extractions, the final weight of the remaining solids was 43.7 G. A final weight reduction of 83% of the slag was realized. The extraction was highly selective for magnesium. The total amount of magnesium recovered after 12 extractions was 44.4 G (calculated on a magnesium metal basis). The remaining slag increased in radioactivity by approximately 6 times (600%). Metal analysis (by atomic emission spectroscopy using an inductively coupled plasma, "ICP"), isotopic Th (by alpha analysis), and gamma analysis for Th-daughters all demonstrated the absence of Th in the Mg(HCO.sub.3).sub.2 liquor. A small amount, about 0.6 pCi/mL (16 Bq/mL) of the .sup.232 Th daughters were extracted into the liquor. EXAMPLE 2 Dried, radioactive magnesium slag [249 G, 259 pCi/G (7,000 Bq/G)] was treated as in Example 1 for one extraction. After filtering the remaining solids, the Mg(HCO.sub.3).sub.2 liquor contained 0.6 pCi/mL (16 Bq/mL) average .sup.232 Th daughters. The liquor was aerated to remove the CO.sub.2 and precipitate MgCO.sub.3. The MgCO.sub.3 precipitate was analyzed for radioactivity and found to have 20 pCi/G (541 Bq/G) average .sup.232 Th daughters. EXAMPLE 3 The acidified Mg(HCO.sub.3).sub.2 liquor from Example 1, extraction 3, was analyzed for radioactivity by gamma analysis. The average .sup.232 Th daughter activity was 0.44 pCi/mL (12 Bq/mL). Approximately ten drops of concentrated H.sub.2 SO.sub.4 was added to the liquor, followed by 0.5 G of BaCl.sub.2.2H.sub.2 O in 10.5 G of deionized water. The liquor was then re-filtered and the filtrate analyzed for radioactivity. The average .sup.232 Th daughters activity was 0.035 pCi/mL (0.9 Bq/G). Upon reanalysis of the filtrate three days later, no detectable levels of .sup.232 Th daughters were found. EXAMPLE A COMPARATIVE Dried, radioactive magnesium slag, 251 G, was treated as in Example 1 for one extraction. However, after filtering the liquor was not acidified. The liquor was then treated with 0.5 G of BaCl.sub.2.10H.sub.2 O and 0.8 G of Na.sub.2 SO.sub.4. The liquor was filtered again and aerated to remove CO.sub.2. The white MgCO.sub.3 precipitate was analyzed for radioactivity and found to have 62 pCi/G (1676 Bq/G) average of .sup.232 Th daughters. The liquor following both precipitations was found to have 0.01 pCi/mL (0.3 Bq/mL) average of .sup.232 Th daughters. EXAMPLE 4 Dried, radioactive refined magnesium slag [253 G, 259 pCi/G (7,000 Bq/G)] was treated as in Example 1 for one extraction. After filtering the remaining solids, 10 mL of a BaSO.sub.4 suspension (prepared from 0.225 G of BaCl.sub.2.2H.sub.2 O, 12 G of K.sub.2 SO.sub.4, 6 mL (1.1 G) of H.sub.2 SO.sub.4 and 100 mL of water) was added to the Mg(HCO.sub.3).sub.2 filtrate. The liquor was then filtered to remove the BaSO.sub.4 coprecipitated with the .sup.232 Th daughters. The filtrate was analyzed for radioactivity and the average .sup.232 Th daughter activity was 0.145 pCi/G (3.9 Bq/G). The MgCO.sub.3 precipitate activity was approximately 4.4 pCi/G (119 Bq/G). Three days later no .sup.232 Th daughters were found in the MgCO.sub.3 precipitate. EXAMPLE 5 RECYCLE OF BaSO.sub.4 Dried, radioactive magnesium slag, 274 G, was treated as in Example 1 for one extraction. In addition, 10 mL of a BaSO.sub.4 suspension (prepared as in Example 4) was added. The remaining solids were filtered and Mg(HCO.sub.3).sub.2 liquor aerated to remove the CO.sub.2 and precipitate MgCO.sub.3. The MgCO.sub.3 was analyzed for radioactivity and found to have 1.8 pCi/G (49 Bq/G) average of .sup.232 Th daughters. The remaining slag solids were then recycled as in Example 1 for one extraction without adding more BaSO.sub.4. The precipitated MgCO.sub.3 was analyzed for radioactivity and found to have 4.4 pCi/G (119 Bq/G) average of .sup.232 Th daughters. After a third extraction (by the procedure of Example 1), the precipitated MgCO.sub.3 was found to have 14 pCi/G (378 Bq/G) average of .sup.232 Th daughters. EXAMPLE 6 NON-RADIOACTIVE MAGNESIUM SLAG A. CO.sub.2 Process A 240 G sample (325 mL) of dried, ground, non-radioactive crude magnesium slag was extracted 10 times in the manner described for Example 1. A metal analysis by ICP was performed on the acidified Mg(HCO.sub.3).sub.2 liquors. The remaining solids following extraction and separation were dried and weighed. Following 10 extractions the remaining solids weighed 54.1 G. A final weight reduction of 77.5% of the slag was realized. The process was highly selective for magnesium and the total amount of magnesium recovered after ten extractions was 46 G (calculated on a magnesium metal basis). B. Compaction Process A 49 G portion of the remaining dried slag material after the ten extractions was mixed with 20 mL of deionized water and then compacted to 41 mL using a Harvard miniature compactor [40 lb. (1.8 kg) spring]. Thus the 54 G sample could be compacted to 45 mL. Combining both the extraction process and the compaction method permitted an 86% volume reduction. Other embodiments of the invention will be apparent to those skilled in the art from a consideration of this specification or practice of the invention disclosed herein. It is intended that the specification and examples be considered as exemplary only, with the true scope and spirit of the invention being indicated by the following claims. |
062917366 | summary | FIELD OF THE INVENTION The present invention pertains to the field of chemical fixation of hazardous waste materials, including metal-bearing materials and radionuclides and radioactive substances, in debris, soils, solid materials, sludges and materials precipitated or filtered from liquids, rendering such hazardous waste materials within a stabilized, insoluble, non-leachable, non-zeolitic and pH stable form suitable for safe and ecologically-acceptable disposal; typically regulated by the U.S. Department of Energy, the U.S. Environmental Protection Agency ("USEPA"), and others. The ecologically safe state of the treated materials is not altered by exposure of the treated materials to acidic leachate, acid rain, or radioactive groundwater. In addition, the safe state of the treated materials is not altered by exposure to changing weather conditions; including rain, heat, freeze and thaw. BACKGROUND OF THE INVENTION Various forms of hazardous wastes pose a serious threat to the environment and safe and cost efficient methods for treating and disposing of these wastes has become increasingly important. Hazardous wastes containing excessive amounts of leachable lead are banned from land disposal. The regulatory threshold limit under Resource Conservation and Recovery Act is 5 mg/l of leachable lead as measured by TCLP (toxicity characteristic leaching procedure) test criteria, United States Environmental Protection Agency (USEPA) method 1311 (SW-846). Waste materials containing TCLP lead levels in excess of 5 mg/l are defined as lead-toxic hazardous waste and are as such restricted from land-filling under current land ban regulations. The cost of disposing lead toxic hazardous waste materials is in excess of $200.00 per ton plus the cost of transporting the hazardous material to landfills for hazardous wastes, which do not exist in every state. This makes the disposal of lead toxic hazardous waste material very expensive. Therefore, treating the lead-bearing process materials and waste streams to render them non-hazardous by RCRA definition would cut down the costs of transportation and disposal tremendously. Conventional treatment methods for radionuclides and other radioactive substances can be categorized into three groups: 1) separation; 2) structural containment; and 3) physical stabilization/solidification. These treatment methods are complex, costly, expand volumes, and are only temporary solutions. Various conventional methods have been tried to remove leachable, mobile radionuclides and radioactive substances from soils and other materials. Removal of contamination from soils and solid materials by leaching, centrifugation, extraction and/or washing procedures is extremely expensive and cost-prohibitive because these methods generate vast quantities of contaminated liquid wastes which require further treatment and disposal. Conventional solidification methods based on cementation technology require up to twenty-eight (28) days curing time, increase the waste volume and may raise the pH above 12.5. USEPA defines a pH above 12.5 as hazardous. Hardened cementitious material is not conducive to retreatment in the event treatment fails obligatory confirmation testing. Solidification methods utilizing lime kiln dust, calcium carbonate and/or powdered lime for fixation are, at best, temporary solutions. Furthermore, these methods increase the waste volume and mass. A primary concern is that cementitious methods dilute the parameters of concern in the final waste matrix. In the past, radionuclide and radioactive wastes have been temporarily stored; frequently as a liquid, a sludge, or a contaminated fine-grained solid in conjunction with contaminated soils. The art has recognized that a means must be provided for permanent disposal of these wastes, preferably as non-leachable solids, containing non-migratory radionuclides. Such solids must have certain characteristics which make the solids safe and economical for the long term (10.sup.3 to 10.sup.6 years) retention of radioactive isotopes. SUMMARY OF THE INVENTION The present invention discloses a method of treating hazardous waste materials, including metal-bearing materials and radionuclides and radioactive substances. One embodiment of the present invention relates to a chemical treatment technology for immobilizing leachable lead in contaminated soils and solid waste materials. According to the present invention, a process for treating lead-toxic solid wastes in order to stabilize the leachable lead is disclosed, comprising the steps of: (i) mixing the solid waste with a sulfate compound, such as calcium sulfate dihydrate (gypsum powder) or sulfuric acid, having at least one sulfate ion for contacting waste particles and reacting with said leachable lead to produce a substantially insoluble lead composition, such as anglesite and/or calcium-substituted anglesite; and, (ii) mixing said solid waste and sulfate compound with a phosphate reagent, such as phosphoric acid, having at least one phosphate ion for reacting with said leachable lead to produce a substantially insoluble lead composition. The treated material from this process is substantially solid in form and passes the Paint Filter Test while satisfying the regulatory standard for TCLP lead. In all instances, application of this process has been found very reliable in meeting the treatment objectives and in consistently decreasing waste volume. It is an object of the present invention to provide a technology for treatment of lead-containing solid waste and soil that produces an acceptably low level of leachable lead in the final product to comply with the statutory requirements of TCLP and to pass the Paint Filter Test. Another object of the invention is to provide such a process while producing no wastewater or secondary waste streams during said process. Still another object of the invention is to provide such a process which also causes the solid waste material to undergo a volume reduction as a result of treatment. Yet another object of the invention is to cause fixation of the leachable lead in the solid waste that is permanent under both ordinary and extreme environmental conditions. The present invention relates to treatment methods employed to chemically convert leachable metal in metal-bearing solid and liquid waste materials to a non-leachable form by mixing the material with one or a combination of components, for example, lime or gypsum and phosphoric acid. The solid and liquid waste materials include contaminated sludges, slurries, soils, waste waters, spent carbon, sand, wire chips, plastic fluff, cracked battery casings, bird and buck shots and tetraethyl lead contaminated organic peat and muck. The metal-bearing materials referred to herein which the present invention treats include those materials having leachable lead, aluminum, arsenic (III), barium, bismuth, cadmium, chromium (III), cooper, iron, nickel, selenium, silver and zinc. The present invention discloses a process comprising a single step mixing of one or more treatment additives, and a process comprising a two step mixing wherein the sequence of performing the steps may be reversible. The present invention provides a novel way of treating a plurality of metal-contaminated materials at a wide range of pH. The method works under acidic, alkaline and neutral conditions. The processes of the present invention provide reactions that convert leachable metals, especially lead, into a non-leachable form which is geochemically stable for indefinite periods and is expected to withstand acid rain impacts as well as the conditions of a landfill environment. A first group of treatment chemicals for use in the processes of the present invention includes lime, gypsum, alum, halites, Portland cement, and other similar products that can supply sulfates, halites, hydroxides and/or silicates. A second group consists of treatment chemicals which can supply phosphate ions. This group includes products such as phosphoric acid, pyrophosphates, triple super phosphate (TSP), trisodium phosphate, potassium phosphates, ammonium phosphates and/or others capable of supplying phosphate anion when mixed with a metal-bearing process material or with a metal-toxic hazardous waste. Depending on the process material or waste (i) matrix (solid, liquid or mixture thereof), (ii) category (RCRA or Superfund/CERCLIS), (iii) chemical composition (TCLP lead, total lead level, pH) and (iv) size and form (wire fluff, shots, sand, peat, sludge, slurry, clay, gravel, soil, broken battery casings, carbon with lead dross, etc.) the metal-bearing material is mixed with one or more treatment chemicals in sufficient quantity so as to render the metal substantially non-leachable, that is, to levels below the regulatory threshold limit under the TCLP test criteria of the USEPA. For lead-bearing materials, the treatment additives render the lead below the regulatory threshold limit of 5 mg/l by the TCLP test criteria of the USEPA. The disposal of lead-hazardous and other metal-hazardous waste materials in landfills is precluded under land ban regulations. It is an object of the present invention to provide a method of treating metal-bearing materials, contaminated soils and waste effluent, and solid wastes containing hazardous levels of leachable metal. It is a further object to provide a method which decreases the leaching of lead in lead-bearing materials to levels below the regulatory limit of 5 mg/l by TCLP test criteria. It is another object of the present invention to provide a method to immobilize lead to leachable levels below 5 mg/l by TCLP test criteria, through the use of inexpensive, readily accessible treatment chemicals. With this method, the leachability of lead is diminished, usually allowing municipal landfill disposal which would not otherwise be permitted. Yet another object of the present invention is to provide a treatment method for metal-bearing wastes, particularly lead-bearing wastes, which comprises a single step mixing process or a two-step process wherein the sequence of the two steps may be reversed. Another object of the present invention is to provide a method of treating a wide variety of lead bearing process materials, wire fluff and chips, cracked battery plastics, carbon with lead dross, foundry sand, lead base paint, leaded gasoline contaminated soils, peat and muck, sludges and slurries, lagoon sediment, and bird and buck shots, in order to render the material non-hazardous by RCRA definition, and pass the EPTOX, MEP, ANS Calvet and DI Water Extract tests. Another object of the present invention is to extend the scope for broad application in-situ as well as ex-situ on small as well as large quantities of metal-bearing process materials or generated waste streams. The present invention provides a method which converts metal-toxic process materials and hazardous wastes into a material which has a lower leachability of metal as determined by EPA's TCLP test. Such treated waste material can then be interned in a licensed landfill, a method of disposal only possible when the leachability of metal is diminished/reduced to levels below the regulatory threshold limit by TCLP test criteria, e.g., lead below 5 mg/l. Another embodiment of the present invention relates to a chemical treatment process that reduces the leachability and solubility of radionuclides and other radioactive substances contained in debris, soils, sludges and solid materials ("the host material" or "the host matrix"). The process for treating radionuclides and other radioactive substances employs the same methods and treatment chemicals used for treating metal-bearing hazardous waste materials. The process comprises contacting radionuclides and other radioactive substances in the host matrix with the first and second group treatment chemicals to promote recrystallization of the host material into Apatitic-structured end-products. Preferred reactants are comprised of at least one phosphate group and create mineral species of Apatitic geometric structures with reduced nuclide leachability and solubility. In the preferred embodiment, technical grade phosphoric acid (TGPA) is used in a one step process. TGPA contains sulfate as an impurity in addition to a phosphate anion source. The Apatite-structure ((AB).sub.5 (PO.sub.4).sub.3 Z) is preferred since the anion Z position is usually a halogen or a hydroxyl, both active scavengers of cations. The unique properties of the Apatitic-structure, (AB).sub.5 (XO.sub.4).sub.3 Z, are key to this invention. Just as low-temperature Apatite is nature's ion-prison in the biological/biosphere environment and high-temperature Apatite is natures ion-prison in the pegmatites/igneous lithosphere environment, Apatites can do the same in man-made (unnatural/synthetic) radioactive environments. The supplementary problem of metamict lattice disruptions, from the generation of excess heat and ion-cannon recoil damage by radioactive decay, is also self-resolved in Apatites. Both low-temperature and high-temperature Apatitic-structures are self-healing and non-leaching. In one embodiment of the present invention, the flow of normal groundwater through the treated material should be encouraged since the groundwater will disperse the build-up of heat and eliminate the requirement for costly cooling of monolithic encasement structures. In another embodiment of the present invention, treated material contacted with groundwater contaminated with radionuclides and radioactive substances reduces the radioactive level of the ground water. Natural scavenging of Lanthanides and Actinides by Apatitic-structure phosphate-complexing phases is well-documented from research conducted in connection with the mining of oceanic deposits throughout the world to produce phosphate products. To date, more than 300 Apatite mineral species have been classified by geologists. Substitution within Apatites are extremely complex. Many require a charge-compensating mechanism that can be grossly estimated from ionic radii and coordination numbers. Common substitution mechanisms noted are as follows: 1) simple within-site substitutions; 2) coupled substitutions involving chemically similar cations; 3) substitutions involving large cations, such as Cs, with smaller cations; 4) substitutions involving cation vacancies; 5) substitutions coupling specific cations with specific anions; 6) substitutions involving anions; 7) substitutions involving anion vacancies; and 8) substitutions involving a change in valence. From the structural and compositional studies of natural and synthetic Apatites, it is known that Apatites are complex geological structures. The present invention has found that Apatites can sustain a great variety of substitutions following the general formula (AB).sub.5 (XO.sub.4).sub.3 Z, [sometimes written, (AB).sub.10 (XO.sub.4).sub.6 Z.sub.2 ], wherein: A = Coordination Number 7 thru 12, most commonly 9. Cations smaller than Mn.sup.+2 are to small for an 8 coordination number, unless combined with a larger cation. = Ca, Sr, Mn, Pb, Mg, Ba, Zn, Cd, Fe, Ni, Co, Sn, Eu, Cu, and Be among divalent elements. = Na, K, Rb, Ag, Cs and possibly Li among monovalent elements. = Al, Fe, Y, rare earth elements (REE) except Eu and Ce, Bi and possibly Nb, Sb and Ti among trivalent elements. = U, Pb, Th, Zr, Ce, Transuranics and possibly Tl among quadrivalent elements. = [] minor lattice vacancies. B = Coordination Number 6 thru 9, most commonly 8. Cations smaller then W.sup.+6 are small for 6 coordination number and those larger than Zr.sup.+4 are too large. = Ca, Sr, Mn, Pb, Mg, Ba, Zn, Cd, Fe, Ni, Co, Sn, Cu, and Be among divalent elements. = Na, K, Rb, Ag, Li possibly among monovalent elements. = Al, Fe, Sc, Sb, Y, Eu and Ce REE, Nb, Bi and possibly Ta among trivalent elements. = Si, Mn, Ti, Mo, W, Sn, U, Th, Zr, C among quadrivalent elements. = Actinide ion species conforming to Metal.O.sub.2 (especially UO.sub.2). = [] minor lattice vacancies. XO.sub.4 = PO.sub.4, SiO.sub.4, SO.sub.4, AsO.sub.4, VO.sub.4, CrO.sub.4, BeO.sub.4, MoO.sub.4, CO.sub.3, CO.sub.3 F, WO.sub.4, MnO.sub.4, CO.sub.3 OH, BO.sub.4, AlO.sub.4, Fe.sub.3 O.sub.4, possibly GeO.sub.4, and SeO.sub.4. Z = F, OH, Cl, Br, I, O and [] minor lattice vacancy in structure of defective Apatites. Element 43--Technetium is effected by the process with leachability greatly reduced; however, its positioning within the Apatitic-structure has not been determined with certainty. Additionally, the radius ratios among A, B and XO.sub.4 components, and their respective coordination number, can have a strong influence on the Apatite-structure. Problems occur when an element's ionic radius is small for A and large for B; therefore, a single site cannot be considered alone and a partitioning between A and B sites is proposed. The partitioning is extremely difficult to predict since the amounts involved may be very minor as well as promoting localized crystal disorder. In its simplest and most efficient form, the current invention provides for the addition of at least one member selected from a first group of treatment chemicals that can supply sulfates, halides, hydroxides and/or silicates and at least one member selected from a second group of treatment chemicals that can supply phosphate ions to material consisting of, or containing, radionuclides and other radioactive substances. Technical grade phosphoric acid ("TGPA") that contains up to 70% (by weight) phosphate (as P.sub.2 O.sub.5) and sulfate (SO.sub.4.sup.-2), typically as sulfuric acid in the range of 2.5% to 7% (by weight) as an impurity, is a source of both a sulfate ion and a phosphate ion and can, therefore, be used as a single reactant. The addition of water at any point in the process aids in the dispersion of the TGPA throughout the host matrix. As the TGPA disperses and permeates through the matrix and during the course of subsequent reactions, the leachability and solubility of radionuclides and other radioactive substances is reduced. Supplemental mechanical or physical mixing can also be employed to enhance the contact of the TGPA with the leachable species in the host material. As a true chemical process, an object of the invention relies on molecular bonding and crystal nucleation principles to reduce nuclide solubility and to create conditions suitable for matrix volume reduction resulting, in part, from the dehydration properties of the treatment chemicals. When TGPA is utilized, molecular rearrangement and minimized addition of treatment agents is characteristic of the invention and supplemental buffering agents or traditional strength enhancement physical-binding additives typical of physically stabilized mixtures are not required. The end-product of the invention is a material that contains no free liquids and produces no supernatant wastewater or secondary waste streams. A further loss of water weight is achieved by capillary drying and evaporation which also contribute to volume reduction. Some volume reduction can be attributed to acidic carbonate destruction, especially those not incorporated into the Apatitic structures. The end product is friable and can be handled with traditional earth-moving equipment, as it is not monolithic in form. Moreover, the end-product can be made to have enhanced geotechnical properties without compromising the chemistry of the nuclide leachability/solubility reduction. The addition of water, either to suppress dust or due to rainfall, and excavation or other material handling activities do not affect the nuclide leachability or solubility of the end-product. Another object of the invention is to increase the level of protection offered by disposal facility designs engineered specifically to control, isolate, or contain material characterized with leachable radionuclides; and to minimize the migration of radionuclides and other radioactive substances from material that is accessed by the percolation of rain and surface waters, and/or the intrusion and flow-through of groundwater or leachate that can act as an ion-carrier. When groundwater contaminated with radionuclides and radioactive substances are contacted with materials treated by the present invention, the radioactive level of the groundwater will be reduced. The radionuclides and radioactive substances in the groundwater react with the phosphate compounds and sulfate compounds in the treated materials to form geochemically stable Apatite-structures. A further object of the present invention is the addition of liquid or solid reagents to a solid material or sludge without creating secondary byproducts or separable streams. Another object of the present invention is to engage and employ preexisting carbonates, borates, sulfates, and/or silicates within the matrix at the time of phosphate anion addition so that they contribute to the formation of Apatitic-structures that reduce nuclide leachability and solubility and host matrix volume. An additional objective of the invention is the immediate initiation of process reactions upon the contacting of phosphate anion with the leachable or soluble species, without the separation of nuclides or other byproducts from the matrix. Another objective is the in situ or ex situ application of process reagents to nuclide material; wherein fixation of the nuclides is permanent under both ordinary and extreme environmental conditions. Still another object of the invention is the use of acidity to enhance dissassociation of semi-soluble species so that problematic nuclides are freed to nucleate within the Apatite crystals. These and other objects will be apparent from the detailed description of the invention set forth below. The invention may be more fully understood with reference to the accompanying drawings and the following description of the embodiments shown in those drawings. The invention is not limited to the exemplary embodiments but should be recognized as contemplating all modifications within the skill of an ordinary artisan. |
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description | This application is a divisional application and claims priority to U.S. patent application Ser. No. 13/742,392, filed Jan. 16, 2013, entitled METHOD AND APPARATUS FOR REFUELING A NUCLEAR REACTOR HAVING AN INSTRUMENTATION PENETRATION FLANGE, and is related to U.S. application Ser. No. 13/457,683, filed Apr. 27, 2012, entitled INSTRUMENTATION AND CONTROL PENETRATION FLANGE FOR PRESSURIZED WATER REACTOR. 1. Field This invention relates in general to nuclear reactor systems, and in particular to nuclear reactors with instrumentation penetrations through an upper portion of the reactor vessel, below the reactor closure head 2. Description of Related Art A pressurized water reactor has a large number of elongated fuel assemblies mounted within an upright reactor vessel. Pressurized coolant is circulated through the fuel assemblies to absorb heat generated by nuclear reactions in fissionable material contained in the fuel assemblies. The primary side of such a nuclear reactor power generating system which is cooled with water under pressure comprises an enclosed circuit which is isolated from and in heat exchange relationship with a secondary circuit for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internal structure that supports the plurality of fuel assemblies containing the fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. In conventional nuclear plants of that type each of the parts of the primary side comprising the steam generator, a pump and a system of pipes which are connected to the reactor vessel form a loop of the primary side. For the purpose of illustration, FIG. 1 shows a simplified conventional nuclear reactor primary system, including a generally cylindrical pressure vessel 10 having a closure head 12 enclosing a nuclear core 14. A liquid coolant, such as water or borated water, is pumped into the vessel 10 by pump 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18, typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump 16, completing the primary loop. Typically, a plurality of the above-described loops are connected to a single reactor 10 by reactor coolant piping 20. An exemplary conventional reactor design is shown in more detail in FIG. 2. In addition to the core 14 comprised of a plurality of parallel, vertical co-extending fuel assemblies 22, for the purpose of this description, the other vessel internal structure can be divided into the lower internals 24 and the upper internals 26. In conventional designs, the lower internals function to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity in FIG. 2), and support and guide instrumentation and components, such as control rods 28. In the exemplary reactor shown in FIG. 2, coolant enters the reactor vessel through one or more inlet nozzles 30, flows down through an annulus between the reactor vessel and the core barrel 32, is turned 1800 in a lower plenum 34, passes upwardly to a lower support plate 37 and a lower core plate 36 upon which the fuel assemblies are seated and through and about the fuel assemblies 22. In some designs, the lower support plate 37 and the lower core plate 36 are replaced by a single structure, a lower core support plate having the same elevation as 37. The coolant flow through the core and surrounding area 38 is typically large on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate 40. Coolant exiting the core 14 flows along the underside of the upper core plate and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially to one or more outlet nozzles 44. The upper internals 26 can be supported from the vessel or the vessel head and include an upper support assembly 46. Loads are transmitted between the upper support assembly 46 and the upper core plate 40, primarily by a plurality of support columns 48. A support column is aligned above a selected fuel assembly 22 and perforations 42 in the upper core plate 40. Rectilinearly moveable control rods 28 which typically include a drive shaft or drive rod 50 and spider assembly 52 of neutron poison rods, are guided through the upper internals 26 and into aligned fuel assemblies 22 by control rod guide tubes 54. The guide tubes are fixedly joined to the upper support assembly 46 and the top of the upper core plate 40. The support column 48 arrangement assists in retarding guide tube deformation under accident conditions which could detrimentally affect control rod insertion capability. To control the fission process, a number of control rods 28 are reciprocally moveable in guide thimbles located at predetermined positions in the fuel assemblies 22. Specifically, a control rod mechanism positioned above the top nozzle of the fuel assembly supports a plurality of control rods. The control rod mechanism (also known as a rod cluster control assembly) has an internally threaded cylindrical hub member with a plurality of radially extending flukes or arms that form the spider 52 previously noted with regard to FIG. 2. Each arm is interconnected to a control rod 28 such that the control rod assembly mechanism 72 is operable to move the control rods 28 vertically within guide thimbles within the fuel assemblies to thereby control the fission process in the fuel assembly 22, under the motive power of the control rod drive shaft 50 which is coupled to the control rod mechanism hub, all in a well-known manner. The upper internals 26 also have a number of in-core instrumentation that extend down through axial passages within the support columns 48 and into instrumentation thimbles generally, centrally located within the fuel assemblies. The in-core instrumentation typically includes a thermocouple for measuring the coolant core exit temperature and axially disposed neutron detectors for monitoring the axial and radial profile of neutron activity within the core. Nuclear power plants, which employ light water reactors require periodic outages for refueling of the reactor. New fuel assemblies are delivered to the plant and temporarily stored in a fuel storage building, along with used fuel assemblies which may have been previously removed from the reactor. During a refueling outage, a portion of the fuel assemblies in the reactor are removed from the reactor to the fuel storage building. A second portion of the fuel assemblies are moved from one support location in the reactor to another core support location in the reactor. New fuel assemblies are moved from the fuel storage building into the reactor to replace those fuel assemblies which were removed. These movements are done in accordance with a detailed sequence plan so that each fuel assembly is placed in a specific location in accordance with an overall refueling plan prepared by the reactor core designer. In conventional reactors, the removal of the reactor internal components necessary to access the fuel and the movement of new and old fuel between the reactor and the spent fuel pool in the fuel storage building is performed under water to shield the plant maintenance personnel. This is accomplished by raising the water level in a refueling cavity and canal that is integral to the plant's building structure. The water level of more than 20 feet provides shielding for the movement of the reactor internal structures and the fuel assemblies. Refueling activities are often on a critical path for returning the nuclear plant to power operation, therefore, the speed of these operations is an important economic consideration for the power plant owner. Furthermore, the plant equipment and fuel assemblies are expensive and care must be taken not to cause damage or unnecessary radiation exposure due to improper handling of the reactor components that have to be removed to access the fuel assemblies, the fuel assemblies or fuel transfer equipment. The precision of these operations is also important since the safe and economical operation of the reactor core depends upon each fuel assembly being in its proper location. A typical pressured water reactor needs to be refueled every eighteen to twenty-four months. Commercial power plants employing the conventional designs illustrated in FIGS. 1 and 2 are typically on the order of 1,100 megawatts or more. More recently, Westinghouse Electric Company LLC has proposed a small modular reactor in the 200 megawatt class. The small modular reactor is an integral pressurized water reactor with all primary loop components located inside the reactor vessel. The reactor vessel is surrounded by a compact, high pressure containment. Due to both the limited space within the containment and the low cost requirement for integral pressurized light water reactors, the overall number of auxiliary systems needs to be minimized without compromising safety or functionality. For example, the compact, high pressure containment associated with the design of some small modular reactors does not allow for the incorporation of a large floodable cavity above the reactor vessel in which the transferred components can be shielded. Furthermore, in most traditional pressurized water reactors, the in-core instrumentation is retracted from the core prior to refueling. This is done by breaking primary pressure boundary seals and pulling the instrumentation through a conduit tube. This procedure is straight forward in plants with bottom mounted instrumentation since the conduit just extends from the bottom of the reactor vessel to a seal table located in a room separated from the reactor. In plants with top mounted instrumentation, this procedure is much more challenging because of the upper internal structure. This is further complicated when top mounted instrumentation is considered for use in an integral pressurized water reactor of a small modular reactor. Top mounted instrumentation is preferred in plants that use a severe accident mitigation strategy commonly referred to as in-vessel retention. This strategy requires that there are no penetrations in the lower portion of the reactor vessel. Accordingly, it is an object of this invention to provide a method and apparatus for nuclear plants that employ top mounted instrumentation that will facilitate access to the core for refueling. It is a further object of this invention to provide such a method and apparatus that will facilitate removal of the top mounted instrumentation as an integral part of the upper internals structure within the reactor vessel. It is an additional object of this invention to provide such a method and apparatus wherein the instrumentation penetrations through the vessel are removed from the vessel as an integral part of the upper internals package. These and other objects are achieved by a nuclear reactor having an elongated reactor vessel enclosed at a lower end and having an open upper end on which an annular flange is formed and a central axis extending along an elongated dimension. The reactor vessel has a removable head having an annular portion on an underside of the head that is machined to form a sealing surface with the annular vessel flange. A first removable annular seal ring, sized to seat on the reactor vessel flange between the flange and the sealing surface on the underside of the reactor vessel head, with the seal ring being interposed between the sealing surface on the underside of the vessel head and the flange on the reactor vessel and having a thickness sized to sealably accommodate one or more radial passages through which one or more instrument conduits pass from outside of the reactor vessel to an interior thereof to communicate instrumentation signals out of the core of the nuclear reactor, the core having a plurality of fuel assemblies. An upper internals package supported above the core within the reactor vessel has a plurality of hollow support columns respectively having a substantially vertical passage therethrough, that extend through and between an upper core plate and an upper support plate of the upper internals package, the passage through the upper core plate being aligned with a corresponding instrument thimble within one of the fuel assemblies. A plurality of hollow tubes are fixedly connected to the first removable annular seal ring, with each of the hollow tubes being slidable mounted within the passage of one of the support columns with at least one of the instrument conduits extending axially through the hollow tube into the corresponding support column. Each of the hollow tubes is slidably mounted within the passage of one of the support columns and moveable between a fully inserted position and a fully extended position wherein in the fully inserted position, the instrument conduit enters the instrument thimble and in the fully extended position, the instrument conduit is withdrawn from the core. Preferably, a lower end of the hollow tube is captured within the passage of the support column. In one embodiment, at approximately a lower end of travel of the hollow tube within the support column passage, the walls of the support column passage thicken to provide a tighter fit than experienced between an intermediate axial extent within the support column passage between the fully inserted position and the fully extended position. Desirably, a lower end of the hollow tube is narrower than an intermediate axial portion of the hollow tube. In another embodiment, wherein the first removable annular seal ring extends radially between an approximate extent of an outer wall of the reactor vessel and a wall of the upper internals package, including a second removable annular seal ring positioned below the first removable annular seal ring between the first removable annular seal ring and the reactor vessel flange and having substantially the same radial extent as the first removable seal ring. A radially outwardly extending abutting surface on each of the first and second seal ring are sealed to each other by at least one o-ring and an axially directed primary coolant passage extends and is substantially aligned through each of the first and second seal ring. An inwardly extending abutting surface on each of the first and second seal ring extends on an opposite side of the primary coolant passage from the o-ring and is sealed by a “T” shaped ring with the web of the “T” extending between the inwardly abutting surfaces. Preferably, the web of the “T” is secured by a fastener attached to one of the inwardly extending abutting surfaces and passing through a clearance hole in the web. In one embodiment the clearance hole is slotted to permit thermal expansion. Desirably, the “T” shaped ring is constructed from a material that upon heat up of the nuclear reactor expands faster than the material the first and second removable annular seal rings are constructed from. Preferably, the first and second removable annular seal rings are constructed from carbon steel and the “T” shaped ring is constructed from stainless steel. In another embodiment, the first removable annular seal ring extends radially between approximately an extent of an outer wall of the reactor vessel and a wall of the upper internals package and the nuclear reactor further includes a second removable annular seal ring positioned below the first removable annular seal ring between the first removable annular seal ring and the reactor vessel flange and has substantially the same radial extent as the first removable annular seal ring. A radially outwardly extending abutting surface on each of the first and second seal ring is sealed to each other by at least two radially spaced O-rings and a first leak off channel extends from between the at least two radially spaced O-rings to a collection reservoir which is connected to a second leak-off channel extending from between abutting surfaces on the second removable annular seal ring and the reactor vessel flange. In still another embodiment, the hollow tube is supported by a substantially horizontally extending grid structure that is connected to the first removable annular seal ring. Preferably, the first removable annular seal ring is configured so that raising the first removable annular seal ring raises the grid structure and raises the hollow tube from within the corresponding support column. The invention also contemplates a method of refueling the nuclear reactor described above including the step of removing the reactor head from the first removable annular seal ring. The first removable annular seal ring is then raised to an elevation that withdraws the instrument conduit from the core. Next, the method withdraws the upper internals package, including the first removable annular seal ring in the raised position, as a single unit, out of the reactor vessel to a storage location. The method then refuels the core. Preferably, after the refueling step, the method maintains the first removable annular seal ring in the raised position and lowers the upper internals package into the reactor vessel. The upper internals package is then supported above the core and the first removable annular seal ring is lowered on top of the reactor vessel flange, simultaneously lowering the hollow tubes within the corresponding support columns to lower the instrument conduits into the corresponding instrument thimbles in the fuel assemblies. The reactor vessel head is then replaced on the reactor vessel flange. In a further embodiment, the first removable annular seal ring extends radially between an approximate extent of an outer wall of the reactor vessel and a wall of the upper internals package, including a second removable annular seal ring positioned below the first removable annular seal ring between the first removable annular seal ring and the reactor vessel flange. The second removable annular seal ring has substantially the same radial extent as the first removable annular seal ring and is fixedly connected to the wall of the upper internals package. In this latter embodiment of the method, the step of withdrawing the upper internals package includes the step of removing the second removable annular seal ring from the reactor vessel flange as part of the upper internals package. FIGS. 3 and 4 illustrate a small modular reactor design available from the Westinghouse Electric Company LLC, Cranberry Township, Pennsylvania, to which the apparatus and method concepts of this invention can be applied. Though it should be appreciated that the invention can also be applied to a conventional pressurized water reactor design such as the one illustrated in FIGS. 1 and 2, as will be further explained hereafter. FIG. 3 shows a perspective view of the reactor containment 11, partially cut away to show the pressure vessel 10 and its internal components. FIG. 4 is an enlarged view of the pressure vessel shown in FIG. 3. The pressurizer 58 is common to most pressurized water reactor designs, though not shown in FIG. 1, and is typically included in one loop to maintain the system's pressure. In the small modular reactor design illustrated in FIGS. 3 and 4, the pressurizer 58 is integrated into the upper portion of the reactor vessel head 12 and eliminates the need for a separate component. It should be appreciated that the same reference characters are employed for corresponding components among the several figures. A hot leg riser 60 directs primary coolant from the core 14 to a steam generator 18 which surrounds the hot leg riser 60. A number of cooling pumps 16 are circumferentially spaced around the reactor vessel 10 at an elevation near the upper end of the upper internals 26. The reactor coolant pumps 16 are horizontally mounted axial flow canned motor pumps. The reactor core 14 and the upper internals 26, except for their size, are substantially the same as the corresponding components previously described with regards to FIGS. 1 and 2. From the foregoing, it should be appreciated that employing the traditional refueling method by flooding the reactor well above the area of the vessel flange 64 and transferring the fuel assemblies under water to a spent fuel pool by way of a transfer canal 62 that extends through the containment would not be practical with this type of containment and compact design. Furthermore, because the compact design practices in-vessel retention, it requires top mounted instrumentation which cannot practically exit the reactor vessel through the vessel head. Since the in-core instruments, which are traditionally used to measure core power and/or core exit temperatures need to be withdrawn from the fuel assemblies before refueling can begin, an innovative design is required that will permit the signal leads from the in-core instrumentation to exit the reactor through a penetration flange such as that described in U.S. patent application Ser. No. 13/457,678, filed Apr. 27, 2012, entitled “Instrumentation and Control Penetration Flange for Pressurized Water Reactor,” while enabling the in-core instrumentation to be removed from the core before the upper internals package above the core is removed from the reactor vessel. For efficiency and to reduce radiation exposure, the instrumentation and control penetration flange structure is preferably removed as an integral part of the upper internals package. A further understanding of the operation of the small modular reactor illustrated in FIGS. 3 and 4 can be found in U.S. patent application Ser. No. 13/495,050, filed Jun. 13, 2012, entitled “Pressurized Water Reactor Compact Steam Generator.” The apparatus and method of the embodiments of this invention claimed hereafter require the introduction of a penetration flange 66 between the reactor vessel flange 64 and the reactor head flange 68 as shown for a traditional pressurized water reactor in FIG. 6. This may be an additional penetration as to that required in certain small reactor designs as described in U.S. patent application Ser. No. 13/457,683, cited above, with one flange substantially dedicated to control rod drive mechanism power and the other flange serving the in-core instrumentation. In the case of the traditional pressurized water reactor, a single penetration flange 66 would be required as illustrated in FIG. 6. As previously described, the in-core instruments which are traditionally used to measure core power and/or core exit temperatures need to be withdrawn from the fuel assemblies before refueling can begin. The instrument is housed within a thimble tube with connected cabling that extends from the penetration flange to the fuel assembly. This thimble tube forms the primary pressure boundary and protects the instrument from the reactor coolant. FIG. 7 shows an enlarged view of a portion of the upper internals package for the integral pressurized water reactor shown in FIG. 5. A further enlarged view of a cross-section of the upper and lower penetration flanges 70 and 66 is shown in FIG. 8. The lower penetration flange 66 in the integrated pressurized water reactor configuration is shown extending between the outer wall of the reactor vessel 10 and an inner wall 76 extending up vertically from the upper support assembly 46. The flange 66 carries the connector and electrical conduits through which the control rod drive system 78 is powered, the upper penetration flange 70 carries the connector and electrical conduit through which the in-core instrumentation cabling signals 80 are conveyed. During refueling, after removal of the closure head 12, but prior to the removal of the upper internals 26, the upper penetration flange 70 is raised. The instrument thimble 82 which extends down into the fuel assembly 22 is supported within the upper internals 26. Just above the core 14, the core support columns 48 have a central passage drilled axially through their center which provides the conduit for the instrument thimble's passage into the core. Above the upper support plate 46 a tube 84 is aligned with the central passage through the support column 48. The tube 84 is supported within the structure 86 of the upper internals 26. The columns 48 support the in-core instruments and prevent them from buckling while the penetration flange 70 is being lowered during reactor reassembly after refueling. The instrumentation tube 84 is connected to and supported by a grid structure 88 that is connected to the upper penetration flange 70 and is raised and lowered with the upper penetration flange. The tubes 84 slide inside the support columns 48 maintaining alignment of the two assemblies as the penetration flange 70 is being raised or lowered. The columns 48 and the tubes 84 maintain a relatively loose clearance until the penetration flange 70 is in its lowest position. At this point, narrowed sections 90 of the instrument tube 84 engage thickened sections that narrow the passage of the column 48 to provide a tighter fit. The tubes 84 in columns 48 remain engaged through the full range of motion of the in-core instrument penetration flange 70, as can be appreciated from FIGS. 5, 6 and 7 for an integral pressurized water reactor design. Preferably, the lower penetration flange 66 that carries the power cables to the control rod drive mechanism 78 remains stationary, being fixedly attached to a vertical section 76 of the upper support assembly 46, while the upper penetration flange 70 is raised. However, it should be appreciated that, alternately, the lower penetration flange 66 can be attached to the upper penetration flange 70 or the two penetration flanges can be constructed as a single unit and raised or lowered together. However, the latter arrangement is less desirable, because it would require considerable slack in the power cables to the control rod drive mechanisms which would have to be accommodated. FIGS. 9, 10, 11, 12 and 13 show the disassembly sequence for the integrated pressurized water reactor shown in FIG. 5. FIG. 9 shows the reactor fully assembled with the vessel head 12 sealed to the vessel 10 with the flange bolts 74 in place and the instrument thimble 82 fully inserted into the core. It should be noted that the upper internals design illustrated in FIGS. 5-18 differs slightly from that previously illustrated in FIG. 2 in that the design shown in FIGS. 5-18 includes an extended upper internals package that extends the support columns 48 and, in the conventional pressurized water reactor design, the guide tubes 54 between the upper support plate 46 and an upper support assembly support structure 86. FIG. 10 shows the first stage in the reactor disassembly process for an integral pressurized water reactor design in which the head 12 including the steam generator 18 is removed. FIG. 11 shows the upper penetration flange being raised which retracts the in-core instrumentation thimble 82 from the fuel assemblies 22. FIG. 12 shows the upper internals being raised and FIG. 13 shows the reactor vessel with the upper internals completely removed. FIGS. 14, 15, 16, 17 and 18 show the corresponding sequence for the disassembly of a modified conventional pressurized water reactor design. In a pressurized water reactor design, the control rod drive mechanism drive rod travel housing 56 extends through the reactor head 12 and the drive mechanism is positioned above the head so that the control rod mechanism cabling is never within the reactor vessel. As is currently practiced, in refueling conventional pressurized water reactors, the control rod drive rods are disconnected from the spider assembly at the hub before the head 12 is removed and the drive rods are removed with the head. FIG. 14 is a duplicate of FIG. 6 showing the pressurized water reactor design fully assembled with the head secured by the flange bolts 74. FIG. 15 shows the head 12 removed with the control rod drive rods. FIG. 16 shows the upper penetration flange 70 (though, in this configuration there is no lower penetration flange) raised and the in-core instrumentation thimble 82 retracted from the core 14. FIG. 17 shows the upper internals partially removed and FIG. 18 shows the internals completely removed and moved off to a storage location. The assembly of the reactor after refueling is simply the reverse of the process just described. During reassembly of the reactor vessel, the internals are placed in the vessel above the core then the penetration flange is lowered until it makes contact with the reactor vessel flange. The primary pressure boundary seal between each of the flanges, i.e., the head flange 68, the upper penetration flange 70, the lower penetration flange 66 and the reactor vessel flange 64, is maintained with a pair of O-rings 92 with leak-off lines 94 used to monitor the seal. Each flange connection uses a pair of O-rings 92. A hole 94 through the flange can be used to connect the voids between the pairs of O-rings 92 allowing for a single set of leak-off lines to be used. These lines would connect to the leak-off lines of the reactor vessel flange 64 which drain into a common leak-off reservoir (not shown) so that they can remain connected to the plant leak-off monitoring systems during refueling. As described above, in the case of an integral pressurized water reactor, two flanges 66 and 70 may be used to introduce penetrations through the reactor pressure boundary. In some designs, a potential for core by-pass flow between the main coolant return passage 102 and the upper internals exists through an inner gap 108 between the flanges 66 and 70. The preferred embodiment for this configuration includes a sealing device 96 that takes advantage of the difference in the rate of thermal expansion for carbon and stainless steels. The device is a ring which has a “T” shaped cross section and is attached to one of the flanges. The device 96 is fastened through the web 98 of the “T” with threaded fasteners 100. The clearance hole through the web is slotted to allow for the ring to expand during reactor heat-up. During reactor assembly there is a generous clearance between the seal and the flanges. As the plant heats up, the stainless steel ring expands faster than the reactor flanges, creating pressure between the components and providing the required seal. The pressure differential is relatively low with an example range of 10 to 20 psi (69 to 138 kPa). An extended sleeve 104 attached to one of the flanges may also limit bypass flow and reduce the pressure drop as primary coolant passes through the flanges in the primary fluid flow channel 102. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof. |
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description | The present invention relates to an electron emission method employing an element that is made of a material expressed by a general formula of BN and containing sp3 bonds, sp2 bonds or a mixture thereof and has a surface profile showing an excellent field electron emission characteristic that makes it capable of operating for field electron emission in the atmosphere. More particularly, the present invention relates to an electron emission member having a unique configuration and showing an exceptionally remarkable field electron emission characteristic (of a current density of more than 1,000 times relative to similar conventional members) that has been developed with an objective of finding applications in the field of lamp type light source devices and field emission type displays using a field electron emission source and also to a method of manufacturing such an electron emission member. Additionally, the present invention relates to an emission/display device that employs boron nitride expressed by a general formula of BN and having at least sp3 bonds as cold cathode type electron source for electron emission. More particularly, the present invention relates to an emission/display device of the above identified type in which the electron source comprises boron nitride having a profile with pointed protrusions and an excellent field electron emission characteristic so that the device shows a low electron emission threshold value, a high output level and a long service life. Liquid crystal displays and VFD (vacuum fluorescent displays) have been put to use in display sections of mobile phones and displays mounted in vehicles and electronic appliances in recent years. Research and development efforts have been and being paid for organic ELs as promising choice for such displays. However, they have respective disadvantages. More specifically, (1) since liquid crystal does not emit light spontaneously and requires a back light when used in a display, the display is inevitably complex and it is difficult to design an ultimately thin display and (2) VFDs intrinsically provide a low display resolution and hence can display only simple images, whereas (3) Organic ELs have not been commercialized because of the problem of service life that has not hitherto been dissolved yet and (4) LEDs are accompanied by a problem that they require to be bundled by a large number to form a structure when they are used as illumination/display devices and hence are not very convenient. Massive research and development efforts have been and are being paid for field electron emission type displays as alternative displays. Field electron emission type displays include FEDs (field emission displays) and SEDs (surface-conduction electron-emitter displays). It is expected that these devices and related systems will be putting on significance more and more in the future. As a matter of fact, intensive research efforts are being paid for the purpose of improving devices and systems of the type under consideration and developing new field electron emitting materials. Field electron emitting materials are required to show a low field electron emission threshold, a high withstand voltage and a high current density. Materials recently attracting attention as field electron emitting materials include carbon nano-tubes. However, carbon nano-tubes require improvement of the electron emitting performance and the current density when designing an electron emitting material on the basis of this material. Efforts are being paid for patterning nano-tubes in order to grow thin film out of them and processing them to produce a profile adapted to electron emission. However, the process of manufacturing carbon nano-tubes has not been perfectly established and the technological development for processing them is still under way. In short, it is very difficult to realize a field electron emission display on the basis of carbon nano-tubes. Additionally, the current density that can be achieved by way of a cumbersome process of treating carbon nano-tubes is in the order of mA/cm2 at most. Carbon nano-tubes face a limit in terms of operating field intensity and problems such as degradation of material and exfoliation arise beyond the limit to make them undurable under hard operating conditions including a high voltage and a long operation time. While some reports say that displays using carbon nano-tubes are on the stage of experimental manufacture, the development of such displays is basically still in a difficult situation. Field electron emission technologies are highly important. It will not be necessary to explain further that they influence not only specific technological fields but also the society at large and daily lives of ordinary people. Thus, more and more efforts will be paid for the development of field electron emission technologies. There is a strong demand for materials that can withstand a high field intensity and stably emit electrons with a high current density for a long period of time without degradation and damages. The inventors of the present invention also have paid intensive research efforts in order to meet the demand and looked into boron nitride that is attracting attention as a heat-resistant and abrasion-resistant material. As a result of studies on electron emitting materials based on the compound, the inventors of the present invention came to find that boron nitride film that is prepared under certain conditions shows a surface profile that is remarkably good for emitting field electrons and withstands a high field intensity. More specifically, the inventors of the present invention found that, in the process of producing and depositing boron nitride on a base (which may be flat plate-shaped, wiry, spherical or of some other shape) by way of a reaction from a gas phase, boron nitride of a certain bond type is formed as film on the base when ultraviolet rays are irradiated on and near the base with a high energy level and sp3 bond type boron nitride is produced with a pointed profile and grown in a self-organizing manner in the direction of irradiation of rays at appropriate intervals and that the obtained film easily emits electrons when an electric field is applied to it and can stably maintain its condition and performance without being degraded, damaged and exfoliated, maintaining an exceptionally large current density. The inventors have already applied for a patent for the achievement (see Patent Documents 1 and 2). Patent Document 1: Jpn. Pat. Appln. Laid-Open Publication No. 2004-35301 Patent Document 2: Jpn. Pat. Application No. 2003-209489 The present invention is an improvement to the conventional art disclosed in the above-cited patent documents. The problem to be solved by the present invention is to provide a field electron emission element that operate stably in the atmosphere, a method of manufacturing the same and a field electron emission method using such an element as well as an emission/display device employing a cold cathode type electron source having a surface profile showing an excellent field electron emission characteristic along with a low electron emission threshold value, a high output level and a long service life. More specifically, the present invention provides a field electron emission element that is formed in a self forming manner from a gas phase by way of a reaction, using a material expressed by a formula of BN that is produced according to the prior art so as to have a pointed profile and contain sp3 bonds, sp2 bonds or a mixture thereof, and operates well in the atmosphere to show an excellent electron emitting performance and an emission/display device comprising a cold cathode type electron source having a low electron emission threshold value, a high output level and a long service life. As a result of intensive research efforts for solving the above problem, the inventors of the present invention succeeded in developing an electron emission element that operates stably for electron emission in the atmosphere by utilizing boron nitride provided by the conventional art and having a specific physical surface profile so as to perform excellently for electron emission. The inventors also tried to use such boron nitride for FEDs and illuminations and as field electron emitting material that can be employed generally for emission/display devices. The inventors of the present invention kept on developing devices, expecting that emission/display devices showing a low electron emission threshold value, a high output level and a long service life can be realized by using such devices. As a result, the inventors of the present invention came to find that it is possible to prepare devices of an outstanding level if compared with the prior art. This invention is based on the finding and is defined as follows. (1) A field electron emission element characterized in that it comprises a boron nitride material containing crystal that is formed on an element substrate to show a pointed profile and expressed by BN and that it shows a stable electron emitting property in the atmosphere when a voltage is applied thereto.(2) The field electron emission element as defined in (1) above, characterized in that the boron nitride material containing crystal that has a pointed profile and is expressed by BN is formed in a self-forming manner on the element substrate at intervals and to a density suitable for electron emission.(3) The field electron emission element as defined in (1) or (2) above, characterized in that the boron nitride material containing crystal that has a pointed profile and is expressed by BN is made of an sp3 bond type BN, an sp2 bond type BN or a mixture thereof.(4) The field electron emission element as defined in one of (1) through (3) above, characterized in that the boron nitride material containing crystal that has a pointed profile and is expressed by BN is formed by a reaction from a gas phase when excited by ultraviolet rays.(5) The field electron emission element as defined in one of (1) through (4) above, characterized in that the field electron emission element is employed for an emission/display device.(6) The field electron emission element as defined in one of (1) through (4) above, characterized in that the field electron emission element is employed for an illumination device.(7) A method of manufacturing a field electron emission element adapted to emit electrons stably in the atmosphere when a voltage is applied thereto, characterized by causing a dilute material gas of rare gas such as argon and/or helium, hydrogen or a mixture gas thereof to react by irradiating ultraviolet rays onto an electron emission element substrate held to a temperature level between room temperature and 1,300° C. in an atmosphere where boron source material gas and nitride source material gas are introduced to 0.0001 to 100 volume % relative to the dilute material gas under pressure of 0.001 to 760 Torr, causing or without causing plasma to be generated, and a boron nitride material containing crystal that has a pointed profile and is expressed by BN to be formed on the element substrate in a self-forming manner.(8) The method of manufacturing a field electron emission element as defined in (7) above, characterized in that the boron nitride material containing crystal that has a pointed profile and is expressed by BN is made of an sp3 bond type BN, or a mixture of the sp3 bond type BN and an sp2 bond type BN.(9) An electron emission method, characterized by applying a voltage to the field electron emission element as defined in one of (1) through (6) above to make it emit electrons.(10) The electron emission method as defined in (9) above, characterized in that the electron emitting property of the field electron emission element is improved by bringing the field electron emission element into contact with an actuating atmosphere containing polar solvent gas when making it emit electrons by applying a voltage to the field electron emission element.(11) The electron emission method as defined in (10) above, characterized in that the polar solvent gas is water and/or alcohol.(12) A cold cathode type emission/display device, characterized in that it comprises a boron nitride material containing crystal that is formed on an element substrate to show a pointed profile and expressed by BN as a field electron emission source necessary for exciting phosphor to emit light.(13) The cold cathode type emission/display device as defined in (12) above, characterized in that the field electron emission source is a boron nitride material containing crystal that has a pointed profile and is expressed by EN and formed in a self-forming manner on the element substrate at intervals and to a density suitable for electron emission.(14) The cold cathode type emission/display device as defined in (12) or (13) above, characterized in that the boron nitride material containing crystal that has a pointed profile and is expressed by BN is made of an sp3 bond type BN, or a mixture of the SP3 bond type BN and an sp2 bond type BN.(15) The cold cathode type emission/display device as defined in one of (12) through (14) above, characterized in that the boron nitride material containing crystal that has a pointed profile and is expressed by BN is formed by a reaction from a gas phase when excited by ultraviolet rays.(16) The cold cathode type emission/display device as defined in one of (12) through (15) above, characterized in that the field electron emission source is arranged directly on, opposite to or separated from the phosphor in a container having a window and that light emitted from the phosphor is taken out from the window.(17) The cold cathode type emission/display device as defined in (16) above, characterized in that the container is a vacuum container in the inside of which vacuum prevails.(18) The cold cathode type emission/display device as defined in (16) or (17) above, characterized in that the phosphor is powdery or filmy.(19) The cold cathode type emission/display device as defined in one of (16) through (18) above, characterized in that the phosphor is applied to the window.(20) The cold cathode type emission/display device as defined in one of (16) through (19) above, characterized in that the phosphor is tricolor phosphor that emits RGB rays of light.(21) A method of manufacturing a cold cathode type emission/display device, characterized by comprising: causing a dilute material gas of rare gas such as argon and/or helium, hydrogen or a mixture gas thereof to react by irradiating ultraviolet rays onto an electron emission element substrate held to a temperature level between room temperature and 1,300° C. in an atmosphere where boron source material gas and nitride source material gas are introduced to 0.0001 to 100 volume % relative to the dilute material gas under pressure of 0.001 to 760 Torr, causing or without causing plasma to be generated, and a boron nitride material containing crystal that has a pointed profile and is expressed by BN to be formed on the element substrate in a self-forming manner; taking out the reaction product from the reaction vessel with the substrate after the end of the reaction; and assembling the cold cathode type emission/display device, using the reaction product as field electron emission source.(22) The method of manufacturing a cold cathode type emission/display device as defined in (21) above, characterized in that the boron nitride material containing crystal that has a pointed profile and is expressed by BN is made of an sp3 bond type BN, or a mixture of the an sp3 bond type BN and an sp2 bond type BN. For the surface profile of a field electron emission element showing an excellent field electron emission characteristic according to the present invention to be formed in a self-forming manner, it is necessary to be irradiated with ultraviolet rays at the time of the reaction from a gas phase. This is a fact that is made clear by the inventors of the present invention in the above-cited patent documents. As described in the above-cited patent documents, the inventors of the present invention believe that the following explanation holds true. As Ilya Prigogine (a Nobel Laureate) pointed out, surface morphosis by self-organization can be grasped as “Turing structure” that appears under certain conditions where a surface diffusion and a surface chemical reaction of a precursor substance take place conflictingly. In the case of the present invention, irradiation of ultraviolet rays participates in photo-chemically accelerating them and influences the regular distribution of initial cores. The growth reaction on the surface is accelerated by irradiation of ultraviolet rays. This means that the reaction speed is proportional to the intensity of irradiated ultraviolet rays. If the initial cores are assumed to be semispherical, the intensity of irradiated rays is high and the growth is accelerated at and near the apex, whereas the intensity of irradiated rays is low and the growth is retarded in the peripheral edge area. It may be safe to assume that this is one of the factors that produce a pointed surface profile. All in all, irradiation of ultraviolet rays takes a very important role and there will be no denying that it is very important. For the purpose of the present invention, the expression of “showing a stable electron emitting property in the atmosphere” does not mean that a field electron emission element according to the present invention is to be used limitedly in the atmosphere in terms of conditions and modes of operation. The above expression means that a field electron emission element according to the present invention can operate properly without being held in a vacuum container, whereas it is difficult for conventional field electron emission elements to operate stably in the atmosphere and hence are normally held in a vacuum container and driven to operate in vacuum. Thus, the above expression does not mean that a field electron emission element according to the present invention needs to be used limitedly in the atmosphere. In other words, the modes of operation of a field electron emission element according to the present invention include those in a vacuum container as in the case of conventional elements as well as those in the atmosphere. Thus, a field electron emission element according to the present invention operates satisfactorily in the stage where crystal having a pointed profile and expressed by BN is formed on the element substrate and the present invention covers a field electron emission element in that stage of course, the present invention also covers a field electron emission element where the element substrate on which a boron nitride material that contains the crystal is formed is integrally combined with some other means to produce a unit or a module. Additionally, the present invention covers a field electron emission element in a state of being integrally fitted to the inside of a container, in which the atmosphere and the pressure may or may not be adjusted to vacuum. Conventionally, the operation of drawing out electrons from a substance relies on the use of a field electron emitting material showing a high electron emission threshold value. In the case of cold cathode type devices, it is indispensable to apply a large voltage in vacuum, or in the case of thermal electron type, it is indispensable to heat the electron emitting material to a high temperature level not lower than 2,000° C. in vacuum. Devices that utilize electrons drawn out into a space require a costly special arrangement for containing the device in vacuum in a hermetically sealed condition. To the contrary, present invention provides a field electron emission element (field electron emitting material) that is a thin film formed on a substrate operating as an electronic member by irradiating it with ultraviolet rays, made of a material expressed by a general formula of BN and mainly containing sp3 bonds or a mixture with sp2 bonds and showing a pointed profile. A field electron emission element according to the present invention shows a remarkable property of having a low electron emission threshold value and being able to stably emit field electrons in the atmosphere simply by applying a voltage to it. Additionally, advantages of the present invention include the following. When such a material is used for the electron source of a cold cathode type emission/display device, it is energy saving because it can be driven to start operating with ease and additionally, it is not degraded if used for a long time in severe operating conditions to consequently prolong the service life of the device because BN itself is a stable compound. Furthermore, when a thin film is formed by the material in a self-forming manner and incorporated into a device as an electron emitter, it is possible to simplify the structure of the device and the process of preparing the device to a great advantage of cost. Since the thin film part including the emitter has only a thickness of several to tens of several micrometers, it is possible to produce ultra-thin devices. 1. reaction vessel (reactor) 2. gas inlet port 3. gas outlet port 4. boron nitride depositing substrate 5. optical window 6. excimer ultraviolet laser 7. plasma torch Now, the present invention will be described in greater detail by referring to the accompanying drawings and by way of examples. A CVD reaction vessel having a structure as shown in FIG. 1 can be used for obtaining a boron nitride showing an excellent field electron emission characteristic and containing sp3 bonds, or a mixture of the sp3 bonds and SP2 bonds. Referring to FIG. 1, the reaction vessel 1 is equipped with a gas inlet port 2 for introducing reaction gas and dilute gas thereof and a gas outlet port 3 for discharging the introduced reaction gas and so on to the outside of the vessel and connected to a vacuum pump so that the internal pressure of the vessel is reduced and maintained to a level lower than that of the atmospheric pressure. A boron nitride depositing substrate 4 is arranged on the gas flow route in the vessel. An optical window 5 is fitted to part of a wall of the reaction vessel facing the substrate and an excimer ultraviolet laser 6 is arranged so as to irradiate the substrate with ultraviolet rays by way of the window. The reaction gas introduced into the reaction vessel is excited by ultraviolet rays irradiated onto the surface of the substrate and the nitrogen source and the boron source in the reaction gas reacts with each other in a gas phase to produce boron nitride that is expressed by general formula BN and contains sp3 bonds or a mixture with sp bonds. The produced boron nitride grows to become film. It has been made clear as a result of experiments that the reaction can proceed within a wide pressure range in the reaction vessel between 0.001 and 760 Torr and also within a wide temperature range of the substrate arranged in the reaction space between room temperature and 1,300° C., although the internal pressure and the substrate temperature in the reaction vessel is preferably low and high respectively to obtain a highly pure target reaction product. In a mode of carrying out the present invention, plasma is irradiated with ultraviolet rays to the surface of the substrate and a surrounding space region. The plasma torch 7 in FIG. 1 indicates this mode of carrying out the invention. As shown in FIG. 1, the reaction gas inlet port and the plasma torch are integrally arranged and directed to the substrate so that both reaction gas and plasma may be shot toward the substrate. After the above-described synthetic reaction, the reaction product may be taken out with the substrate from the reaction vessel and arranged in an emission/display device so as to operate as electron emitter. The invention of this patent application is carried out in a reaction vessel as described above. This will be described in greater detail by way of specific examples and by referring to the accompanying drawings. However, the examples described below are disclosed only to make the present invention easily understandable and by no means limit the present invention. As pointed out above, the object of the present invention is to provide a field electron emission element having a surface profile that shows an excellent field electron emission characteristic and is formed in a self-forming manner so as to mainly contain sp3 bond type boron nitride or a mixture with sp2 bond and a method of manufacturing the same. In other words, the reaction conditions may be appropriately selected and altered so long as the above object is achieved. Another object of the present invention is to provide a cold cathode type emission/display device comprising an electron emission source formed by using a specific material and hence the reaction conditions may be appropriately selected and altered so long as the above object is achieved. Now, the present invention will be described in greater detail by way of examples. However, as pointed out above, the examples described below are disclosed only to make the present invention easily understandable and by no means limit the present invention. Diborane and ammonia were introduced with respective flow rates of 5 sccm and 10 sccm into a dilute gas flow of argon having a flow rate of 3 SLM. At the same time, excimer ultraviolet rays were irradiated onto a disk-shaped nickel substrate having a diameter of 25 mm and heated to a temperature level of 900° C. in an atmosphere where the pressure was reduced to 10 Torr by means of a pump (see FIG. 1). At this time, the gas was turned into plasma in an inductively coupled manner due to an electric field of 13.56 MHz (it is known that a similar morphosis takes place to produce an excellent field electron emission characteristic if the gas is not turned into plasma, although the growth rate may be affected by the extent of morphosis) The target substance was obtained after a synthesis time of sixty minutes. The crystal system of the specimen was a hexagonal system as determined by X-ray diffractometry and the specimen showed a 5H polygonal structure due to sp3 bonds. The lattice constants of the specimen were a=2.50 Å and c=10.40 Å. As a result, it was confirmed through a scanning electronic microscope (FIG. 2) that the thin film of the obtained substance showed a peculiar surface profile that had been formed in a self-forming manner and was covered by a structure having pointed conical projections (that were several to tens of several micrometers long) apt to produce a concentrated electric field. The field electron emission characteristic of the thin film was examined in the following way. A piece of ITO glass was selected as anode and the specimen (thin film) was used as cathode. The two electrodes were separated from each other with a gap of about 40 micrometers and a voltage was applied to the electrodes to observe the rate of electron emission in the atmosphere. The electrically conductive side of the ITO was made to face the specimen. FIG. 3 summarily shows the obtained results. As seen from FIG. 3, an electric current was discharged from the very beginning without any threshold and an electric current of 1 μA was observed in the atmosphere with a field intensity of about 10 V/μm. During 60 minutes of observation, the average electric current showed no decline, although the electric current fluctuated to a certain extent. A resistor of 1 MΩ was connected to the device to be observed in series in order to prevent the electric current from flowing to a large extent to the ITO electrode. Thus, the electric current can be adjusted by modifying the resistance value of the resistor. For the purpose of reference, a Fowler-Nordheim plot obtained as a result of conducting a similar experiment in vacuum is shown in FIG. 4. In FIG. 4, the horizontal axis indicates 1/V and the vertical axis indicates Log [I/V^2] (where V is the device voltage and I is the current value). It will be seen that the plotted points are substantially on a straight line to show that field electron emission took place in vacuum due to a quantum mechanical tunneling effect. ZnO:Zn fluorescent micro-particles were applied to the specimen (thin film) obtained in Example 1 to a thickness of about 10 μm and arranged vis-à-vis an anode of ITO glass with a gap of about 40 μm separating the surface of the specimen and the anode to prepare a field emission display (FED). A voltage was applied between the anode and the specimen, which was made to operate as cathode, and the rate of electron emission was observed in the atmosphere of 1 atmospheric pressure. Again, a resistor of 1 MΩ was connected to the device to be observed in series in order to prevent the electric current from flowing to a large extent to the ITO electrode. FIG. 5 summarily shows the obtained results. An emission of electrons that was as good as that of FIG. 4 was observed in the atmosphere. An experiment similar to that of Example 2 was conducted in the atmosphere of 1 atmospheric pressure. However, a piece of sponge that was soaked with water was placed in the observation chamber so as to adjust the relative humidity of the air in the observation chamber to about 90%. FIG. 6 summarily shows the obtained results. It will be seen that the rate of electron emission and the electric current rose to about 200 times of those of Examples 1 and 2 due to the humidity adjustment of the operating atmosphere. While the inventors of the present invention believe that this is because of a fall of the electron emission threshold due to the formation of a surface electric dipolar layer by the water adsorbed to the surface, although the phenomenon needs to be looked into thoroughly by studies in the future. However, it is an empirically and experimentally proven fact that the electron emission characteristic can be improved by adjusting the humidity. In the Example, it is confirmed by a tester or the like that insulation between the anode and cathode is maintained. An experiment similar to that of Example 2 was conducted in the atmosphere of 1 atmospheric pressure. However, a piece of sponge that was soaked with ethyl or methyl alcohol was placed in the sealed observation chamber so as to fill the inside of the observation chamber with alcohol-containing air. FIG. 7 summarily shows the obtained results. It will be seen that the rate of electron emission and the electric current rose to about 300 times of those of Examples 1 and 2 due to the addition of alcohol to the operating atmosphere. While the inventors of the present invention believe that the electron emission characteristic was improved because of fall of the electron emission threshold due to the formation of a surface electric dipolar layer by the water adsorbed to the surface. In other words, like water, alcohol tends to be polarized and shows a physical adsorption characteristic relative to the surface of BN so that the adsorption layer forms a surface charged double layer to facilitate electron emission, although the phenomenon needs to be looked into thoroughly by studies in the future. However, it is an empirically and experimentally proven fact that the electron emission characteristic can be improved by adding alcohol to the operating atmosphere. Fluorescent micro-particles for forming a fluorescent display tube (ZnO:Zn particles) were applied to the specimen (thin film) obtained in Example 1 to a thickness of about 10 μm. A device having a structure as illustrated in FIG. 8(a) was prepared in a manner as described below. Firstly, a 50 μm-thick piece of mica is arranged on the surface of the above-described thin film specimen coated with fluorescent micro-particles as an inter-electrode gap forming insulating spacer and a piece of ITO glass was placed thereon with the ITO surface facing the specimen. Thus, the ITO surface was made to operate as anode, whereas the specimen was made to operate as cathode. A gap of about 40 μm was provided between the surface of the fluorescent body directly applied to the cathode and the ITO surface that operated as anode. FIG. 9 illustrates the current-voltage characteristic of the device prepared in the above-described manner in vacuum. A resistor of 1 MΩ was connected to the device to be observed in series in order to protect the device. In FIG. 9, the vertical axis indicates the logarithm of the electric current and the horizontal axis indicates the device voltage. Emitted light was observed with a range of device voltage between 100 and 200 V. It was confirmed that the range corresponds to the region surrounded by a dotted line in FIG. 9. FIG. 10 is a Fowler-Nordheim plot of the data of FIG. 9. In FIG. 10, the horizontal axis indicates 1/V and the vertical axis indicates Log [I/V^2], where V is the device voltage and I is the device current. It will be seen that the plotted points are substantially on a straight line to show that field electron emission took place in vacuum due to a quantum mechanical tunneling effect. A specimen was prepared by using a substrate equivalent that of Example 5 in similar reaction conditions. Then, a piece of ITO glass was brought in and fluorescent micro-particles were applied to the ITO glass side. A light emission device was assembled from them with an insulating spacer of mica interposed between them and the specimen was made to operate as cathode, whereas the ITO glass was made to operate as anode in an experiment where the device was electrically energized under current-voltage conditions similar to those of Example 5 to find that the device similarly emitted light. An RGB element was designed and prepared by combining devices, each being equivalent to that of Example 5, where three colors of green, blue and red fluorescent fine particles were respectively used. When applied a voltage, the device emitted light in RGB. An RGB element was designed and prepared by combining devices, each being equivalent to that of Example 6, where three colors of green, blue and red fluorescent fine particles were respectively used. The device emitted light in RGB. The ITO glass of each of the above examples was replaced by a 0.5 mm copper mesh plate (electrode) and a similar light emission effect was obtained. As described above in detail, the present invention provides a field electron emission element having a surface profile that shows an excellent field electron emission characteristic and is made of a material formed in a self-forming manner so as to mainly contain sp3 bond type BN or a mixture with sp2 bond type BN and a method of manufacturing the same as well as an electron emission method using such an element. Thus, the present invention made it possible to provide a field electron emission element showing a low electron emission threshold value, a high current density and a long service life and hence has a great technological significance. The present invention also provides an emission/display device using the above-described material as field electron emission source and a method of manufacturing the same. Thus, the present invention can expectedly reduce the thickness and the weight of such devices to a great extent in the future. The inventors of the present invention found a particular phenomenon that a thin film grows to show a peculiar profile in a self-organizing manner when irradiated with rays of light. The present invention is based on this finding. If the grown thin film itself is not processed (as grown), it shows a surface profile having a remarkable effect of accelerating the field electron emission performance. Additionally, such a thin film is practically not damaged by the electric discharge of the thin film material itself and maintains a high current density due to the physical characteristics of the material so that it shows a practically permanent service life in such an application. Thus, if compared with the prior art that requires processes for forming a profile and a pattern suitable for field electron emission, the significance of the present invention is not limited to the difference of process and the present invention provides a technology that intrinsically differs from the prior art. According to the present invention, there are provided a thin film that can emit field electrons with a constant current density of 1,000 times of the prior art, or of the order of A/cm2, and is highly durable and a method of manufacturing the same as well as a broad scope of application due to the synergetic effect of the self-forming effect of the surface profile and the outstanding physical characteristics of the material itself. Thus, the present invention is a major breakthrough to the technological status quo and hence really epoch making. As described above in detail, the present invention provides a field electron emission element having (a) a low field electron emission threshold value, (b) a high current density and (c) a long service life of electron emission. When it is used as an electron source in a cold cathode type emission/display device, it provides advantages including an easy startup, reduced weight and thickness of the device, a simplified assembly process and low cost in addition to the above identified advantages. In other words, the present invention can find a broad scope of application in the field of designing devices of the type under consideration. More specifically, the startup operation of such a device is satisfactory in the atmosphere that makes the performance of the device incomparable relative to the prior art. Particularly, since a field electron emission element according to the present invention is outstanding in terms of (b) and (c) above (a current density more than 1,000 times higher than that of the prior art and a strong and durable structure specific to BN) and hence provides a major technological breakthrough, it can find applications in various lamp type light source devices and field emission type displays that are required to show a high luminance level and to be free from material degradation if used for a long time in hostile operating conditions. Conceivable applications of the present invention include ultra-high luminance and high efficiency lighting systems realized to emit electron rays with a current density of more than 1,000 times of the prior art, ultra-high definition displays realized by using the characteristic that a sufficient current value can be obtained with a micro-electron emission area (which will by turn find applications in portable phones, wearable computers and so on), formation of peculiar electron emission patterns by utilizing the electron emission characteristic that only the surface irradiated with ultraviolet rays during the growth period shows, ultra-high luminance nano electron sources and ultra-compact electron beam sources. The scope of application in the field of electronic devices and other technical fields is expected to further expand. Consequently, the present invention will pave the way to technological innovations to various electric appliances and devices that are ubiquitous in our modern daily lives. In short, the scope of applicability of the present invention is very broad and may cover all the areas of human life. Thus, the technological and economic effects of the present invention are global and huge. |
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050671448 | claims | 1. A system for examining an object with penetrating radiation, comprising: a source/modulator which generates a fan of penetrating radiation scanning an object position and comprises wedge-shaped pins of a radiation attenuating material movable into the fan in the scanning direction and arranged in at least one row extending in a direction transverse to both the scanning direction and the propagation direction of said radiation; each of said pins having a ridge pointing to or away from the origin of the fan and each pin having sections of different areas in different planes normal to the scanning direction; and a control circuit moving said pins individually to respective degrees into the scanning fan. a source/modulator which generates penetrating radiation scanning an object position and comprises attenuating portions which are individually movable into the scanning radiation along respective axes; each attenuating portion having sections in planes transverse to the scanning direction whose areas change substantially monotonically with distance along the scanning direction over substantially the entire part of the attenuation portion that is movable into the scanning radiation to intercept the radiation, each section having a dimension in a direction transverse to both the scanning direction and the propagation direction of the radiation which changes monotonically along said propagation direction; and a control circuit which moves said attenuating portions to respective degrees into the penetrating radiation scanning the object position. scanning an object position with a fan of penetrating radiation and moving, together with the scanning fan, attenuating elements which are arranged in at least one row extending in a direction transverse both to the scanning direction and to the propagation direction of the fan and are individually movable into the fan in the scanning direction; each attenuating element when moved into the fan having a ridge which faces toward or away from the origin of the fan and having different areas in sections defined by different planes which are parallel to each other but transverse to the scanning direction; and controlling the respective degrees of movement of the attenuating elements into the fan as a function of the spatial distribution of attenuating material at the object position. 2. A system as in claim 1 in which said pins are in at least two rows spaced from each other in the propagation direction of the fan and are shaped and positioned to cause most rays in the fan to pass through a pin from each row when the pins are all the way into the fan. 3. A system as in claim 2 in which said rows of pins extend along respective arcs centered at the origin of the fan. 4. A system as in claim 3 in which said ridge is rounded. 5. A system as in claim 3 in which said ridge is truncated. 6. A system as in claim 2 in which said sections are generally triangular, each having an apex at said ridge and a base opposite the apex. 7. A system as in claim 6 in which the base does not vary in size as between different sections and as between pins. 8. A system as in claim 1 in which said sections are generally triangular, having bases which are the same in size in all sections and are spaced from each other by a distance much smaller than the size of a base. 9. A system as in claim 8 in which the pins are in at least two rows spaced from each other in the propagation direction of the fan. 10. A system as in claim 9 including a detector/imager which receives the fan exiting said object position and utilizes the received radiation to form a radiographic image of the object position and any object thereat and to generate a signal controlling said control circuit. 11. A system as in claim 1 in which the pins are in two rows spaced along the propagation direction of the fan and wherein when the pins are all the way into the fan a fan ray which passes between two adjacent pins in one row passes through the ridge of a pin in the other row. 12. A system as in claim 11 in which when all the pins are all the way into the fan, the path length of fan rays through the pins in any one of a plurality of planes normal to the scan direction does not vary substantially with angle in the fan. 13. A system as in claim 1 in which the source/modulator comprises an x-ray tube generating said radiation, the distance between the focal spot of the tube and the pins is in the range of about 7 to 15 inches and the largest dimension of a pin in a plane normal to the scanning direction is in the range of about two to four times the size of the focal spot. 14. A system as in claim 13 in which the size of the focal spot is about 1 mm and said largest dimension of a pin is about 3 mm. 15. A system for examining an object with penetrating radiation comprising: 16. A system as in claim 15 in which said sections are generally triangular. 17. A system as in claim 15 in which said portions comprise pins a radiation-attenuating material which individually move into the penetrating radiation scanning the object positon. 18. A system as in claim 15 in which said sections are trapezoidal. 19. A system as in claim 15 in which said portions comprise portions of a diaphragm of a radiationattenuating material and pins which individually move in the scanning direction. 20. A method comprising the steps of: |
abstract | A method of improving the performance of charged beam apparatus. The method including: providing the apparatus, the apparatus comprising: a chamber having an interior surface; a pump port for evacuating the chamber; a substrate holder within the chamber; and a charged particle beam within the chamber, the charged beam generated by a source and the charged particle beam striking the substrate; and positioning one or more liners in contact with one or more different regions of the interior surface of the chamber, the liners preventing material generated by interaction of the charged beam and the substrate from coating the one or more different regions of the interior surface of the chamber. |
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056129866 | claims | 1. A method of forming a desired X-ray image, having minimum image linewidths no greater than s=0.25 .mu.m, on a selected surface of X-ray-sensitive material, the method comprising the steps of: providing a source of partially spatially coherent X-rays, having a predetermined wavelength .lambda. and having a flux at least equal to a predetermined source flux; providing a layer of X-ray-sensitive material with a selected surface to receive an X-ray image thereat; positioning a membrane that is at least partly transparent to X-rays of wavelength .lambda. and that holds a holographic pattern thereat, between the X-ray source and the X-ray-sensitive material so that X-rays produced by the X-ray source are transmitted through the substrate toward the selected surface; determining a hologram, having variable thickness and variable associated transmissivity and optical phase angle, for transmission of X-rays of wavelength .lambda. through the hologram so that, when the hologram is irradiated by the X-ray source, the X-ray image produced at the selected surface by X-rays diffracted by the hologram is the desired X-ray image, with X-ray image minimum linewidths no greater than 0.25 .mu.m, said hologram having a finite number of discrete transmissivity values, where this number is at least two; positioning the hologram on the membrane; and irradiating the selected surface with X-rays received by transmission through the hologram to produce the desired X-ray image. providing a source of partially spatially coherent X-rays, having a predetermined wavelength .lambda. and having a flux at least equal to a predetermined source flux; providing a layer of X-ray-sensitive material with a selected surface to receive an X-ray image thereat; positioning a membrane that is at least partly transparent to X-rays of wavelength .lambda. and that holds a holographic pattern thereat, between the X-ray source and the X-ray-sensitive material so that X-rays produced by the X-ray source are transmitted through the substrate toward the selected surface; determining a hologram, having variable thickness and variable associated transmissivity and optical phase angle, for transmission of X-rays of wavelength .lambda. through the hologram so that, when the hologram is irradiated by the X-ray source, the X-ray image produced at the selected surface by X-rays diffracted by the hologram is the desired X-ray image, with X-ray image minimum linewidths no greater than 0.25 .mu.m; positioning the hologram on the membrane, wherein the presence of a contaminant or defect of diameter no greater than d' at said hologram is compensated for by the additional steps of: irradiating the selected surface with X-rays received by transmission through the hologram to produce the desired X-ray image. providing a source of partially spatially coherent X-rays, having a predetermined wavelength .lambda. and having a flux at least equal to a predetermined source flux; providing a layer of X-ray-sensitive material with a selected surface to receive an X-ray image thereat; positioning a membrane that is at least partly transparent to X-rays of wavelength .lambda. and that holds a holographic pattern thereat, between the X-ray source and the X-ray-sensitive material so that X-rays produced by the X-ray source are transmitted through the substrate toward the selected surface; determining a hologram, having variable thickness h and variable associated transmissivity and optical phase angle, for transmission of X-rays of wavelength .lambda. through the hologram so that, when the hologram is irradiated by the X-ray source, the X-ray image produced at the selected surface by X-rays diffracted by the hologram is the desired X-ray image, with X-ray image minimum linewidths no greater than 0.25 .mu.m; positioning the hologram on the membrane; irradiating the selected surface with X-rays received by transmission through the hologram to produce the desired X-ray image; and further comprising the steps of: providing a source of partially spatially coherent X-rays, having a predetermined wavelength .lambda. and having a flux at least equal to a predetermined source flux; providing a layer of X-ray-sensitive material with a selected surface to receive an X-ray image thereat; positioning a membrane that is at least partly transparent to X-rays of wavelength .lambda. and that holds a holographic pattern thereat, between the X-ray source and the X-ray-sensitive material so that X-rays produced by the X-ray source are transmitted through the substrate toward the selected surface; determining a hologram, having variable thickness and variable associated transmissivity and optical phase angle, for transmission of X-rays of wavelength .lambda. through the hologram so that, when the hologram is irradiated by the X-ray source, the X-ray image produced at the selected surface by X-rays diffracted by the hologram is the desired X-ray image, with X-ray image minimum linewidths no greater than 0.25 .mu.m; positioning the hologram on the membrane; irradiating the selected surface with X-rays received by transmission through the hologram to produce the desired X-ray image; and wherein the step of determining said hologram further comprises the steps of: providing a source of plane wave, approximately spatially coherent X-rays, having a predetermined wavelength .lambda. and having a flux at least equal to a predetermined source brightness; providing a layer of the X-ray-sensitive material with a selected surface to receive an X-ray image thereat; positioning a substrate that holds a holographic pattern thereat, approximately between the X-ray source and the X-ray sensitive material so that X-rays produced by the X-ray source and received by the substrate are reflected and diffracted toward and received by the selected surface; determining a hologram having variable associated optical phase angle, for reflection of X-rays of wavelength .lambda. from an X-ray-receiving surface of the hologram, so that, when the hologram is irradiated by the X-ray source, the X-ray image produced at the selected surface by diffraction of X-rays by the hologram is the desired X-ray image, with X-ray image minimum linewidths no greater than 0.25 .mu.m; positioning the hologram on the substrate; wherein said steps of providing and positioning said hologram comprises the step of providing each of a plurality of selected regions on said substrate with a transmissivity value chosen from a finite number of discrete transmissivity values, where this number is at least two; and irradiating the selected surface with X-rays reflected from the hologram. providing a source of plane wave, approximately spatially coherent X-rays, having a predetermined wavelength .lambda. and having a flux at least equal to a predetermined source brightness; providing a layer of the X-ray-sensitive material with a selected surface to receive an X-ray image thereat; positioning a substrate that holds a holographic pattern thereat, approximately between the X-ray source and the X-ray sensitive material so that X-rays produced by the X-ray source and received by the substrate are reflected and diffracted toward and received by the selected surface; determining a hologram having variable associated optical phase angle, for reflection of X-rays of wavelength .lambda. from an X-ray-receiving surface of the hologram, so that, when the hologram is irradiated by the X-ray source, the X-ray image produced at the selected surface by diffraction of X-rays by the hologram is the desired X-ray image with X-ray image minimum linewidths no greater than 0.25 .mu.m; positioning the hologram on the substrate, wherein the presence of a contaminant or defect of diameter no greater than d' at said hologram is compensated for by the additional steps of: irradiating the selected surface with X-rays reflected from the hologram. providing a source of plane wave, approximately spatially coherent X-rays, having a predetermined wavelength .lambda. and having a flux at least equal to a predetermined source brightness; providing a layer of the X-ray-sensitive material with a selected surface to receive an X-ray image thereat; positioning a substrate that holds a holographic pattern thereat, approximately between the X-ray source and the X-ray-sensitive material so that X-rays produced by the X-ray source and received by the substrate are reflected and diffracted toward and received by the selected surface; determining a hologram having variable associated optical phase angle, for reflection of X-rays of wavelength .lambda. from an X-ray-receiving surface of the hologram, so that, when the hologram is irradiated by the X-ray source, the X-ray image produced at the selected surface by diffraction of X-rays by the hologram is the desired X-ray image, with X-ray image minimum linewidths no greater than 0.25 .mu.m; positioning the hologram on the substrate; wherein the step of determining said hologram further comprises the steps of: irradiating the selected surface with X-rays reflected from the hologram. 2. The method of claim 1, further comprising the step of providing said discrete transmissivity values by providing said hologram with a finite number of discrete thicknesses of material. 3. A method of forming a desired X-ray image, having minimum image linewidths no greater than s=0.25 .mu.m, on a selected surface of X-ray-sensitive material, the method comprising the steps of: 4. A method of forming a desired X-ray image, having minimum image linewidths no greater than s=0.25 .mu.m, on a selected surface of X-ray-sensitive material, the method comprising the steps of: 5. A method of forming a desired X-ray image, having minimum image linewidths no greater than s=0.25 .mu.m, on a selected surface of X-ray-sensitive material, the method comprising the steps of: 6. A method of forming a desired X-ray image, having image minimum linewidths no greater than s=0.25 .mu.m, on a selected surface of X-ray-sensitive material, the method comprising the steps of: 7. A method of forming a desired X-ray image, having image minimum linewidths no greater than s=0.25 .mu.m, on a selected surface of X-ray-sensitive material, the method comprising the steps of: 8. A method of forming a desired X-ray image, having image minimum linewidths no greater than s=0.25 .mu.m, on a selected surface of X-ray-sensitive material, the method comprising the steps of: |
abstract | A nuclear fuel storage cask includes an outer shell having a length extending from a first end to a second end of the outer shell, the outer shell defining an inner cavity circumscribed by the outer shell, an outer perimeter extending around the outer shell, an inner perimeter positioned inward from the outer perimeter, and a cooling circuit extending along the length of the outer shell, the cooling circuit including an inner passage, and an outer passage, a coolant positioned within the cooling circuit, where the coolant is configured to move through the inner passage, absorbing heat from the inner cavity of the outer shell, and the coolant is configured to move through the outer passage, dissipating heat through the outer perimeter of the outer shell, and a lid coupled the outer shell, where the lid covers the inner cavity of the outer shell. |
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summary | ||
055132324 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT A preferred embodiment of a cask 10 for transportation and short-term storage of spent nuclear fuel is shown in FIG. 1. The cask 10 includes a body 12 constructed from a tubular structural shell 14 having an upper shell portion 16 and a lower shell portion 18. The lower shell portion 18 is sealed by a bottom closure plate 20 that has a central access aperture 22 that is sealed with an access cover plate 24. The upper shell portion 16 is sealed with a top closure plate 26. The exterior of the structural shell 14 is shielded with a neutron absorbing shield jacket 28. Two diametrically opposed upper trunnions 30 (only one shown) are secured within upper trunnion mounting sleeves 32 to the exterior of the upper shell portion 16. Two lower trunnions 34 are secured in diametric opposition to the lower shell portion 18 within lower trunnion mounting sleeves 36. As used herein throughout, "bottom" and "lower" refer to the end of the cask 10 and its components closest in proximity to the bottom closure plate 20, while the words "top" and "upper" refer to the opposite end proximate the top closure plate 26. A dry storage canister 38 for spent nuclear fuel is shown installed within the interior cavity 40 of the cask 10. The construction of the dry storage canister 38 is fully described in a U.S. patent application filed on Oct. 8, 1993, in the name of inventors R. A. Lehnert, R. D. Quinn, S. E. Sisley, and B. D. Thomas, entitled CONTAINERS FOR TRANSPORTATION AND STORAGE OF SPENT NUCLEAR FUEL, the disclosure of which is hereby expressly incorporated by reference. A plurality of lugs 42 are secured to the structural shell 14 and to the annular ends of the shield jacket 28 about the circumference of the cask 10, on both the lower and upper (not shown) ends of the shield jacket 28. The purpose of the lugs 42 is to enable mating of the cask 10 with impact limiters during transport. Impact limiters and transportation skids suitable for use in transporting the cask 10 are fully disclosed in a United States patent application filed on Oct. 8, 1993 in the name of inventors R. A. Johnson, I. D. McInnes, R. D. Quinn, and C. J. Temus, entitled IMPACT LIMITER FOR SPENT NUCLEAR FUEL TRANSPORTATION CASK, the disclosure of which is hereby expressly incorporated by reference. Referring now to FIG. 2, the construction of the body 12 shall be described. The body 12 has an overall cylindrical configuration and includes the structural shell 14. The structural shell 14 has a tubular configuration and defines a central longitudinal axis 44 that is aligned with the central longitudinal axes of the other annular components of the body 12, as shall be described. The lower shell portion 18 of the structural shell 14 also has a tubular configuration, defining a circumferential bottom edge 46 and a circumferential top edge 48. The length of the lower shell portion 18 is approximately two-thirds the length of the structural shell 14. The upper shell portion 16 extends the remaining one-third of the length, and defines a circumferential lower edge 50 and a circumferential top edge 52. The upper edge 48 of the lower shell portion 18 abuts and is welded to the lower edge 50 of the upper shell portion 16, using a full penetration weld around the entire circumference of the structural shell 14. The upper shell portion 16 and lower shell portion 18 each have a central axis that is aligned with the longitudinal axis 44, and cooperatively define a right cylinder. The lower shell portion 18 is formed from a rigid material, preferably a corrosion resistant metal, and most preferably a stainless steel, such as ASME SA-240 type 304 austentitic stainless steel. However, the upper shell portion 16 is preferably formed from a material having a higher load bearing strength, also preferably a stainless steel, such as ASME SA-240 type XM-19 high alloy stainless steel. Type XM 19 stainless steel is also austentitic, but has approximately twice the load bearing strength of type 304. As shown in FIG. 1, the upper trunnions 30 are secured to the upper shell portion 16. The upper trunnions 30 are intended to be used for hoisting and lifting the cask 10, both when empty and when loaded with a full canister 38. Thus, the upper trunnions 30 in use transmit significant shear and tensile loads to the upper shell portion 16. The lower shell portion 18 carries the lower trunnions 34, which are used to upright and stabilize the cask 10 during transport, as shall be described subsequently, and thus are subjected to lower loading. Because type XM-19 stainless steel is more costly than type 304 stainless steel, the cost of manufacture is reduced by utilizing the XM-19 for the load bearing portions of the cask 10. Both portions of the structural shell 14 can be formed and welded from rolled plate. Referring to FIG. 2, coaxially installed within the structural shell 14 is an inner shell 54, which also may be formed from type 304 stainless steel or other suitable corrosion resistant structural materials. The inner shell 54 is slightly smaller in external diameter than the interior of the structural shell 14, and thus defines an annular space therebetween. This annular space is filled with a gamma radiation absorbing material 56, such as ASTM B-29 chemical lead. The steel contained in the structural shell 14 and the inner shell 54, as well as the bottom closure plate 20 and top closure plate 26, also serve to absorb gamma radiation. The shield jacket 28 has a tubular configuration and is installed over and surrounds the majority of the length of the structural shell 14. The shield jacket 28 is formed from a tubular outer skin 58. The internal diameter of the outer skin 58 is greater than the external diameter of the shell 14, thus defining an annular space that is filled with a neutron radiation absorbing shield material 60. One suitable neutron radiation absorbing shield material 60 is a hydrogenous solid neutron absorbing material, such as a cementious castable neutron absorbing material. The upper and lower ends of the shield jacket 28 are closed by upper and lower annular support rings 62 and 63, respectively, welded to the exterior of the structural shell 14 and the edges of the outer skin 58. The lower annular support right 63 includes stainless steel rupture discs which prevent over pressurization of the shield jacket 28. A pair of elongate rails 64 are secured by welding or other means to the interior of the inner shell 54 of the body 12. The rails 64 are oriented parallel to the central longitudinal axis 44 of the cask 10, and extend the length of the inner shell 54. Each rail 64, also shown in FIG. 6, is formed from a strip of flat sheet. The rails 64 are spaced radially apart from each other within the same radial quadrant of the inner shell 54. The rails 64 are positioned on the side of the cask body 12 that rests on the trailer or other support surface when the cask 10 is laid down horizontally. Each rail 64 is preferably formed from a material that is harder than the material used to construct the inner shell 54, such as a hardened stainless steel, which provides a non-gouging, low-friction surface for the canister 38 to slide on during installation or removal of the canister 38 from the cask 10. One suitable material is nitronic 60, cold reduced sheet, ASTM A-240, grade UN5, 521800, RC29.35 stainless steel. The bottom edge 46 of the lower shell portion 18 and the bottom edge of the inner shell 54 are each welded to the bottom closure plate 20, thereby sealing the bottom end of the body 12, as shall be described in more detail subsequently. The top edge 52 of the upper shell portion 16 and the top edge of the inner shell 54 are each welded to an annular sealing ring 66. The top closure plate 26 can be secured to the annular sealing ring 66 to selectively close the top end of the body 12. Reference will now be had to FIGS. 2 and 3A to describe the configuration of the annular sealing ring 66. The sealing ring 66 has a main body portion having an essentially rectangular cross section. An annular lower flange 70 extends downwardly from the lower surface of the body portion 68 adjacent the inner edge of the ring 66. The lower flange 70 has an internal diameter substantially equal to the internal diameter of the internal shell 54. The top edge 52 of the upper shell portion 16 is welded to the main body portion 68 of the annular sealing ring 66, while the lower edge of the lower flange 70 is welded to the top edge of the inner shell 54. Both welds are full penetration welds extending around the full circumference of the annular sealing ring 66. The top surface of the body portion 68 defines an annular abutment surface 74. An annular upper flange 76 projects upwardly from the abutment surface 74 along the outer perimeter of the annular sealing ring 66. A hardened sealing surface is formed on the abutment surface 74 by an annular hardened metal inlay 78. The inlay 78 is preferably formed by weld overlay of a hard metal onto the base metal of the annular sealing ring 66. The annular sealing ring 66 is preferably formed from a machined ring forging of type 304 stainless steel. The inlay 78 is preferably formed of inconel alloy. The inlay 78 wraps the inner upper corner of the body portion 68 of the annular sealing ring 66, so that it provides a hard polished surface on both the inner portion of the abutment surface 74 and the upper portion of the internal diameter of the body portion 68. The hard surface provided by the inlay 78 is highly resistant to permanent deformation upon impact of the joint area of the cask 10. Referring still to FIGS. 2 and 3A, the top closure plate 26 is configured as a solid disk. The top plate 26 has an annular recess formed about its perimeter in its bottom side that defines an annular sealing surface 80. The annular sealing surface 80 corresponds in dimension substantially to the abutment surface 74 of the annular sealing ring 66. As shown in FIG. 3A, the top closure plate 26 is installed on the body 12 by sliding the top closure plate 26 within the annular upper flange 76 of the annular sealing ring 66. When so installed, the sealing surface 80 of the top closure plate 26 abuts the abutment surface 74 of the annular sealing ring 76. A non-recessed center portion 82 of the bottom side of the top closure plate 26 is received within the inside diameter of the body portion 68 of the annular sealing ring 66. The inlay 78 provides the sealing surface for the annular sealing ring 66. Two annular grooves 84 are formed in the portion of the sealing surface 80 of the top closure plate 26 that overlies the inlay 78. As shown in FIG. 3A, a seal 86 is received within each of the grooves 84. The seals may be either deformable metal seals, or elastomeric seals, e.g., O-rings, or alternately configured elastomeric seals. The seals 86 are deformed between the top closure plate 26 and the annular sealing ring 66, and retained within the grooves 84. Referring to FIG. 3B, each of the grooves 84 defines a half-dovetail cross-section, having a bottom surface 88, a first orthogonal side surface 90, and a second, inwardly angled side surface 92. The half-dovetail configuration of the grooves 84 ensures that the seals 86 are retained within the grooves 84 when the body 12 is positioned either horizontally or vertically and the top closure plate 26 is removed. The weld between the annular sealing ring 68 and the inner shell 54 is airtight. The weld between the annual sealing ring 68 and the upper shell portion 16 is also believed to be airtight, but is not tested for that characteristic. Likewise, the seal joint formed by the sealing surface 80, abutment surface 74, and seal 86 is also airtight. The top closure plate 26 is selectively secured to the annular sealing ring 66 by installing a plurality of bolts 94 through recessed apertures 96 formed at evenly-spaced intervals about the perimeter of the top closure plate 26 into correspondingly located threaded passages 98 formed in the abutment surface 74 of the annular sealing ring 66. Drain holes (not shown) are provided at the base of each threaded passage 98. Referring to FIG. 6, two monitoring ports 100 are formed in the top closure plate and are selectively sealed by plugs 102. Attention is now directed to FIGS. 2 and 4 to describe the airtight joints formed between the bottom closure plate 20 and the structural shell 14 and inner shell 54. The bottom shell 20 is also configured as a solid disk. An annular flange 104 projects upwardly from the top (i.e., inner) surface of the bottom closure plate 20, at a location spaced radially inwardly from the outer perimeter of the top closure plate 20. When the bottom closure plate 20 is placed over the bottom end of the body 12, an upper edge of the flange 104 abuts the lower edge of the inner shell 54. The upper edge of the flange 104 is welded to the lower edge of the inner shell 54. The bottom edge 46 of the lower shell portion 18 is welded to the bottom closure plate 20. Both welds are full penetration welds formed about the full circumference of the bottom closure plate 20, and the weld between the inner shell 54 and the flange 104 is airtight. The weld between the lower shell portion 18 and the flange 104 is also believed to be airtight, but is not tested for that characteristic. A drain port 106 is formed through the bottom closure plate 20, from the top (inner) surface of the plate to the plate's outer circumference, and is sealed with a threaded bolt 108 capped by a threaded plug 110. The threaded bolt 108 and threaded plug 110 each include a seal (not shown) that is leak tight. The port 106 permits drainage of liquids from the interior cavity 40 of the cask 10. The drain port 106 may be located at any orientation on the bottom of the cask. Referring to FIGS. 2 and 5A, the central access aperture 22 is formed centrally through the bottom closure plate 20. An annular recess 112 is formed in the bottom (i.e., outer) side of the bottom closure plate 20, effectively enlarging the diameter of the bottom portion of the central access aperture 22. The recess 112 defines an annular abutment surface 114. A hardened inlay 116, which may be formed by weld overlay of a hard metal, such as inconel, is formed angularly around the innermost portion of the abutment surface 114 adjoining the access aperture 22. The inlay 116 is polished to define a sealing surface. The access cover plate 24 is configured as a solid disk having an outer diameter that is sized to be received within the recess 112. An annular recess is formed in the top (i.e., inner) side of the access cover plate 24 about the plate's perimeter to define a sealing surface 11B. A non-recessed center portion 120 is bordered by the sealing surface 118. When the access cover plate 24 is assembled to the bottom closure plate 20, the access cover plate 24 is received within the recess 112 of the bottom closure plate 20, with the center portion 120 of the access plate 24 being received within the central access aperture 22. The sealing surface 118 overlies the inlay 116 in this installed configuration. Referring to FIGS. 5A and 5B, two half-dovetailed annular grooves 122, configured similarly to the previously described grooves 84 in the top closure plate 26, are formed in the sealing surface 118. Again, seals (not shown) are received within the grooves 122 and are compressed between the sealing surface 118 and the inlay 116 to form an airtight seal between the ram closure plate 24 and the bottom closure plate 20. The ram closure plate 24 is retained in place by a plurality of bolts 124 inserted through recessed apertures 126 formed at spaced intervals about the periphery of the access cover plate 24 and received within threaded passages 128 formed at corresponding locations in the abutment surface 114 of the bottom closure plate 20. The bottom closure plate 20 is preferably formed from a machine forging, such as a type 304 stainless steel forging. The ram closure plate is preferably formed from a higher strength material, such as type XM-19 stainless steel. Referring to FIG. 6, the construction of the shield jacket 28 will now be described in greater detail. As noted previously, the outer skin 58 of the shield jacket 28 is larger than the external diameter of the upper shell portion 16 and lower shell portion 18. The annular space created therebetween is filled with neutron radiation absorbing shield material 60. Neutron radiation shielding material 60 is not a strong load bearing material, and thus a plurality of elongate reinforcing members 130 are embedded within the shield material 60. The elongate reinforcing members 130 are oriented so as to be parallel to the central axis 44 of the cask body 12. Each reinforcing member 130, which are also illustrated in FIGS. 8 and 9, is bent centrally along its length on two fold lines, such that each member 130 defines a flattened V-shaped cross section. Each member 130 thus has an elongate center portion 132 and first and second elongate leg portions 134 that project angularly outwardly from the center portion 132. The center portion 132 of each member 130 is welded to the interior of the outer skin 58 of the shield jacket 28. The projecting edges of each of the two leg portions 134 contacts and is welded to the outside of the structural shell 14. This gives a "corrugated" reinforcing effect to the structure of the shield jacket 28. The reinforcing members 130 transfer heat from the structural shell 14 through the shield jacket 28 to the exterior of the cask 10 to remove the decay heat of spent fuel contained within the cask 10, and also provide an integral structural system for supporting the cask during transport. Reference is now had to FIGS. 1, 7, and 8 to describe an additional feature of the cask 10. The cask 10 includes a tie-down key way structure 136. The key way structure 136 serves as an anchor point for a tie-down that secures the cask 10 to a transport skid for secure transportation. The key way structure 136 defines an elongate arcuate opening formed through the shield jacket 28 approximately mid-length of the body 12. The key way structure 136 has a radially oriented length and an axially oriented width, and is formed from four frame members that are welded directly to the structural shell 14. Referring now to FIGS. 7 and 8, the long sides of the key way structure 136 are formed by arcuate bearing blocks 138 that are mounted arcuately in spaced-apart disposition on the lower shell portion 18. The perimeter frame of the key way structure 136 is completed by two longitudinally oriented tie-bar members 140 welded across the opposing ends of the bearing blocks 138. Each of the bearing blocks 138 and tie-bars 140 is welded to the lower shell portion 18, and cooperatively define a rectangular frame. A recess 142 is formed in the outer surface of each of the bearing blocks 138 and tie-bars 140 about the inner perimeter of the frame defined thereby. The perimeter frame defined by the bearing blocks 138 and tie-bars 140 are further reinforced by an arcuate pad plate 144 that fits over the bearing blocks 138 and tie-bars 140. The pad plate 144 is disposed within the interior of the shield jacket 28 and is welded directly to the lower shell portion 18, as well as to the bearing blocks 138 and tie bars 140. The outer skin 58 of the shield jacket 28 is also welded to the tie bars 140 and bearing blocks 138. The pad plate 144, tie bars 140, and bearing blocks 138 are preferably formed from a high-strength metal, such as type XM-19 stainless, due to the stress imposed on them during use. Because it is desired that the key way structure 136 be sacrificed rather than the integrity of the structural shell 14 in the event of excessive loads applied to the key way structure 136, the welds between the key way structure 136 and the lower shell portion 18 and outer skin 58 of the shield jacket 28 are relatively small. This ensures that the key way structure 136 will give way prior to breakage of the structural shell 14 in the event of extreme loads on the key way structure 136. The construction of the upper trunnions 30 and lower trunnions 34 will now be described with reference to FIGS. 9 and 10, respectively. The upper trunnions 30 and lower trunnions 34 are similarly constructed except as noted. Thus, only the upper trunnion 30 will be described with it being understood that the same description applies to the lower trunnion 34. The upper trunnion 30 has a cylindrical body 146. An annular flange 148 is formed about the midsection of the body 146. A recess 150 is formed in one of the circular faces 152 of the body 146, and extends fully into the interior of the body 146 to define a cavity 154. The body 146 thus has a hollow configuration. The portion of the trunnion body 146 between the flange 148 and the first face 152 defines a cylindrical base 156. The interior cavity 154 is filled with neutron radiation absorbing shield material 60. The neutron shield material 60 is capped and retained by a circular back plate 158 that is received within the recess 150 and welded in position. The presence of the neutron shield material 60 reduces streaming of neutrons through the upper trunnions 30. A cylindrical bearing projection 160 projects from the second circular face 162 of the trunnion body 146. An annular flange 164 is formed about the end of the bearing projection 160. The bearing projection 160, flange 164, and second circular face 162 cooperatively define a bearing groove that can be grasped by a correspondingly contoured hook for transport of the cask 10. A plurality of apertures 166 are formed through the range 48 at spaced intervals about the perimeter of the upper trunnion 30, for purposes of securement to the cask 10 by bolts 168. The lower trunnions 34 are configured similarly to the upper trunnions 30, except that no cylindrical bearing projection 160 projects from the trunnion body 146. Additionally, the interior cavity 154 is not filled with a neutron shield material, and back plate 158 is also not included. The upper trunnion 30 can be selectively and releasably secured to the cask body 12 by engagement with the upper trunnion mounting sleeve 32. The upper trunnion mounting sleeve 32 consists of a tubular sleeve that projects through and is welded to the upper shell portion 16. A circular aperture 170 is formed through the upper shell portion 16 at the desired location for the upper trunnion mounting sleeve 32. A similarly oriented aperture is formed through the outer skin 58 of the shield jacket 28. The upper trunnion mounting sleeve 32 is installed through the shield jacket 28 and the upper shell portion 16 such that the central axis (not shown) of the upper trunnion mounting sleeve 32 is oriented radially relative to the longitudinal axis 44 of the cask body 12. The upper trunnion mounting sleeve 32 is welded fully about its perimeter to the upper shell portion 16. Additionally, a weld is formed between the outer skin 58 of the shield jacket 28 and the upper trunnion mounting sleeve 30. A circular trunnion filler plate 171 is installed within the upper trunnion mounting sleeve 32, and positioned within the radially inward end of the trunnion mounting sleeve 30 so as to be in line with the arc of the upper shell portion 16. The trunnion filler plate 170 is welded to the interior of the upper trunnion mounting sleeve 32 to seal the radially interior end of the upper trunnion mounting sleeve 32. An annular recess 172 is formed about the entry to the upper trunnion mounting sleeve 32. To secure the upper trunnion 30 in position on the cask 10, the circular base 156 of the upper trunnion 30 is slidably received within the interior passage 174 defined by the upper trunnion mounting sleeve 32, and the flange 148 of the upper trunnion 30 is received within the recess 172. The dimensional tolerances of the interior passage 174 of the upper trunnion mounting sleeve 30 and the recess 172, as well as the base 156 and flange 148 of the upper trunnion 30, are closely controlled such that a very close slip fit is formed between the upper trunnion 30 and the upper trunnion mounting sleeve 32. This ensures that the upper trunnion 30 cannot become cocked within the upper trunnion mounting sleeve 32. The bolts 168 are installed through the apertures 166 and the flange 148 of the upper trunnion 30 and into correspondingly arranged threaded passages 176 formed into the recess 172 of the upper trunnion mounting sleeve 32. Because of this two-piece mounting of the upper trunnion 30, utilizing the separate upper trunnion 30 and upper trunnion mounting sleeve 32, the upper trunnion 30 can be removed as desired when hoisting of the cask 10 is not required. Additionally, because the upper trunnion mounting sleeve 32 receives and captures the upper trunnion 30, the bolts 168 are substantially isolated from shear and tensile loads, which instead are transmitted from the upper trunnion 30 to the upper trunnion mounting sleeve 32 and then to the structural shell 14. This construction helps to ensure that the upper trunnions 30 are not torn off of the structural shell 14 when the upper trunnions 30 are grasped to hoist the weight of the cask 10 and the contents therein. The upper trunnion mounting sleeve 32 and upper trunnion 30 are preferably formed from a high strength metal, such as type XM-19 stainless steel. The trunnion backing plate 158 can be formed from type 304 stainless steel or other suitable metals. Referring to FIG. 10, the lower trunnion mounting sleeve 36 is identically constructed and secured to the lower shell portion 18, as was the upper trunnion sleeve 32 constructed and secured to the upper shell portion 16, except as noted herein. Because the stresses imposed on the lower trunnions 34 are not as great as those imposed on the upper trunnions 30, a recess 172 is not formed in the outer face of the lower trunnion mounting sleeve 36 to receive the flange 148 of the lower trunnion 34. Instead, the axial length of the lower trunnion mounting sleeve 36 is correspondingly reduced, and the flange 148 of the lower trunnion 34 abuts the annular exterior face 178 of the lower trunnion mounting sleeve 36. Referring now to FIG. 11A, often when the cask 10 has been loaded with a canister 38, the cask 10 will be temporarily stationary on-site. During such times, it is not required to mount the upper trunnions 30 and lower trunnions 34 on the cask 10. In such instances, it is desired to further reduce neutron streaming past the trunnions 30 and gamma streaming past trunnions 34 by removing the trunnions 30 and 34, and capping the upper trunnion mounting sleeves 32 with trunnion shields 180 trunnion mounting sleeves 36 with trunnion shields 181. Trunnion shields 180 are metal disks that are filled with neutron shield material 60 (not shown) and bolted to the upper trunnion mounting sleeves 32. Trunnion shields 180 are solid metal disks bolted to the upper trunnion mounting sleeves. Additionally, when not in transport, the key way structure 136 is not being utilized. At such times, it is desirable to mount a key way shield 182 (also shown in FIG. 1) to cover the key way structure 136. The key way shield 182 is again filled with a neutron shield material 60 and is secured by bolting a top plate 184 to the frame of the key way structure 136. This again is to reduce neutron streaming through the key way structure 136. Finally, during unloading of the canister 38 from the cask 10, it is necessary to remove the access cover plate 24 from the bottom closure plate 20, as shall be described briefly below. During such times when the access cover plate 24 is removed from the cask 10 and it is not actually necessary to insert a ram, as shall be described, through the central access aperture 22, an access aperture shield assembly 186 is secured centrally to the bottom closure plate 20 to cover the access aperture 22. Referring to FIG. 11B, the access aperture shield assembly 186 consists of an annular first shield member 188 that is formed from two annular plates 190 that are secured together by an annular ring 192. An aperture 194 is formed centrally through the plates 190, and an internal ring 196 borders this aperture 194. The interior of the first shield member 188 is filled with a neutron absorbing shield material 60. A second similarly constructed shield member 198 is also utilized. Shield member 198 is also formed as a disk, but is a smaller diameter than shield member 188, and includes no central aperture. It also is filled with neutron absorbing shield material 60. Shield member 198 is supported by a plurality of hangers 200, extending outwardly from the first shield member 188 around aperture 194 in the shield member 188. When both the shield member 198 and shield member 188 are utilized, the complete area of the central access aperture 22 is shielded. Referring to FIGS. 12, 13 and 14, in another aspect, the present invention relates to a skid for supporting and protecting the transportation cask for spent nuclear fuel during transportation. FIG. 12 illustrates a conventional trailer 226 that includes a transportation cask enclosed by skid 220 formed in accordance with the present invention and a pair of impact limiters 222 formed in accordance with the invention described in the application entitled Impact Limiter For Spent Nuclear Fuel Transportation Cask. In FIG. 12, the transportation cask is not visible, as it is completely encased by skid 220 and impact limiters 222. Skid 220 is further enclosed by a curtain of expanded metal 224, which further obscures skid 220 and the transportation cask. The curtain of expanded metal 224 is provided around skid 220 in order to shield skid 220 and the transportation cask from sunlight. In FIG. 12, the longitudinal axis of the transportation cask is parallel to the length of trailer 226. Impact limiters 222 are positioned on opposite ends of the generally cylindrical transportation cask. Skid 220 supports the transportation cask along its length between impact limiters 222, as described below in more detail. Referring primarily to FIGS. 13 and 14, transportation skid 220 comprises a lower supporting member 290 and an upper retaining member 292. Lower supporting member 290 carries the vertical and lateral cask loads and includes a plurality of parallel spaced-apart plates 294 lying perpendicular to the longitudinal axis of the transportation cask. Plates 294 include an outer peripheral portion that is substantially square and in use rests on the bed of a transportation trailer. The inner periphery of plates 294 includes a trough which in the illustrated embodiment is semicircular and mates with a portion of the exterior surface of the transportation cask. At the bottom of the trough in supporting member 290 is a saddle 291 that comprises a plate extending lengthwise along the bottom of the trough and widthwise up the sides of the trough. In the illustrated embodiment, saddle 291 occupies approximately one-third of the bottom radius of the trough. At the bottom of the trough centrally located along the length of saddle 291 is an upward protruding rectangular block 296 that serves as a shear key for mating with tie-down keyway structure (136 in FIG. 1) on the transportation cask. Block 296 cooperates with the transportation cask in order to provide an independent means for carrying axial shear loads for the cask. Spaced-apart plates 294 of lower supporting member 290 are connected by a plurality of longitudinal fins 201 running parallel to the longitudinal axis of the transportation cask. In the illustrated embodiment, plates 294 are made from one-inch steel plates and fins 201 comprised of one-half-inch thick steel plates. Plates 294 provide support for the transportation cask for downward, vertical and transverse loads from the cask. Upper retaining member 292 carries vertical upward loads for the cask and includes a plurality of spaced-apart plates 298 lying perpendicular to the longitudinal axis of the transportation cask. In the illustrated embodiment, the inner periphery of plates 298 includes an inverted semi-circular trough that is a minor image of the trough in supporting member 290. The outer periphery of plates 298 is substantially concentric with its inner periphery. Upper retaining member 292 also includes a plurality of parallel longitudinal fins 202 that are positioned parallel to the longitudinal axis of the transportation cask. In the illustrated embodiment, plates 298 and fins 202 are made from metal, such as aluminum. Upper retaining member 292 and lower supporting member 290 mate with each other to define a cylindrical cavity which completely encases the neutron shielding material (60 in FIG. 2). As described above, the neutron radiation shielding material is not a strong load-bearing material, and accordingly, a plurality of elongate reinforcing members (130 in FIG. 6) are embedded within the shield material. The elongate reinforcing members are oriented so as to be parallel to the central axis of the cask body. The radial spacing between fins 201 and fins 202 is such that when the transportation cask is mated with rectangular block 296, the center portions 132 of the elongate reinforcing members 130 in the neutron radiation shielding material are aligned and rest along longitudinal fins 201 and 202. Accordingly, the neutron shielding material does not carry the load of the cask, but rather the elongate reinforcing members resting on the longitudinal fins serves to carry the load of the cask. Utilization of the cask 10 shall now be briefly described. When it is desired to install a canister 38 into the cask 10, the access cover plate 24 is secured to the bottom closure plate 20, while the top closure plate 26 is removed from the cask body 12. The canister 38 is installed into the interior cavity 40 of the cask body 12. These operations are performed inside pools or otherwise in accordance with industry practice. Transport of the open cask during this time is made by grasping the upper trunnions 30 to hoist the cask 10. After water is drained from the interior of the cask 10, and the cask 10 is dried in accordance with standard industry practice, the top closure plate 26 is secured to the cask body 12. The cask 10 is now hoisted by again hooking the upper trunnions 30 to move the cask 10 to a transport trailer. While being hoisted, the cask 10 is oriented vertically with the weight of the cask being supported by the upper trunnions 30. The cask 10 is then repositioned horizontally on a trailer, during which operation the lower trunnions 34 are utilized to stabilize and reposition the cask 10. The cask 10 can then be transported to the site where the canister 38 is to be installed in a horizontal storage module or other storage module. Once the cask 10 has arrived at the storage site, the top closure plate 26 is removed and the top end of the open cask body 12 is docked with the intended storage module. The access cover plate 24 can then be removed, and replaced with the access aperture shield assembly 186 to reduce neutron streaming. When it is time to transfer the canister 38 from the cask 10 to the storage module, the second shield member 198 is removed from the access aperture shield assembly 186. A ram can then be inserted through the remaining shield member 198 and the access aperture 22 into the interior cavity 40 of the cask body 12. The ram then pushes the canister 38, which slides on the rails 64 as it moves through the open end of the cask body 12, defined by the annular sealing ring 66. The canister 38 thus moves into the storage module. Once transfer of the canister 38 is completed, the cask 10 can be reassemble and reutilized. For transportation, the reverse operations to those described above are performed to retrieve the canister into the cask. The cask in then lifted from the trailer and placed on a suitable transportation skid, such as described above, with a shear key which engages the keyway structure 136. The trunnions 30 and 34 are removed and trunnion shields 180 and 181 are installed. While the present invention has been described above in terms of a preferred embodiment, it should be readily apparent to those of ordinary skill in the art that various alterations, modifications and substitutions may be made within the scope of the present invention. For example, materials other than those described can be utilized to form the components of the cask 10, provided that they meet the parameters set forth herein. It is thus intended that the scope of letters patent granted hereon be limited only by the definitions contained in the appended claims. |
claims | 1. A method for both imaging a tumor and treating the tumor of a patient using positively charged particles, comprising the steps of:providing a gantry support connected to a gantry, said gantry connected to at least a portion of a beam transport system;transporting the positively charged particle beam through said portion of said beam transport system;rotating said gantry and said portion of said beam transport system about a gantry rotation axis;translating a translatable imaging system parallel to an axis perpendicular to the gantry rotation axis;imaging the tumor using said translatable imaging system; andtreating the tumor using the positively charged particle beam. 2. The method of claim 1, further comprising the steps of:positioning a first rail and a second rail, of a set of rails, on opposite sides of the patient;positioning an element of said translatable imaging system on said first rail, said element comprising at least one of: a source and a first detector;positioning at least a portion of a detector system element of said translatable imaging system on said second rail; andsaid step of translating linearly moving said set of rails relative to the tumor. 3. The method of claim 2, further comprising the step of:monitoring gamma ray emissions using both: (1) said first detector and (2) a second detector of said detector system, said first detector and said second detector on the opposite sides of the patient. 4. The method of claim 3, further comprising the step of:generating a tomographic image of at least a portion of said tumor using output from both said first detector and said second detector. 5. The method of claim 4, further comprising the step of:linearly translating said translatable imaging system while said step of rotating rotates said gantry. 6. The method of claim 5, further comprising the steps of:extracting the positively charged particle beam from an ion source using a triode extraction system; andsubsequently accelerating the positively charged particle beam using an accelerator prior to said step of transporting. 7. The method of claim 2, said step of imaging further comprising the steps of:linearly translating said translatable imaging system while: (1) said step of rotating rotates said gantry and (2) said step of treating irradiates the tumor using the positively charged particle beam. 8. The method of claim 7, further comprising the step of:dynamically controlling said step of treating using output generated in said step of imaging. 9. The method of claim 8, further comprising the steps of:said step of imaging positioning a first gamma ray detector and a second gamma ray detector on opposite sides of the patient; andsaid step of translating said translatable imaging system: (1) moving said first gamma ray detector and said second gamma ray detector toward an input side of the positively charged particle beam as a function of decreasing energy of the positively charged particle beam and (2) moving said first gamma ray detector and said second gamma ray detector away from an input side of the positively charged particle beam as a function of increasing energy of the positively charged particle beam. 10. The method of claim 7, further comprising the step of:altering an axially cross-section shape of the positively charged particle beam using an insert retracted into an output nozzle of said portion of said beam transport system. 11. The method of claim 8, said step of imaging further comprising the step of:generating a positron emission tomogram using both: (1) output from said first detector mounted on said first rail on a first side of the patient and (2) output from a second detector mounted on said second rail on an opposite side of the patient. 12. The method of claim 1, further comprising:alternating between said step of imaging and said step of treating. 13. The method of claim 1, further comprising:providing a second rotation support system supporting at least a portion of an X-ray source system and a portion of an X-ray detector system of an X-ray system; andcircumferentially rotating all of: said second rotation support system, said portion of said X-ray source system, and said portion of said X-ray detector system about said gantry rotation axis independent from said step of rotating said gantry. 14. An apparatus for both imaging a tumor and treating the tumor of a patient, comprising:a gantry support connected to a gantry, said gantry positioning at least a portion of a beam transport system,wherein said gantry and said portion of said beam transport system rotate about a gantry rotation axis during use; anda translatable imaging system configured to translate parallel to an axis perpendicular to the gantry rotation axis,wherein said translatable imaging system images the tumor during use,wherein said positively charged particle beam, transported through said portion of said beam transport system, treats the tumor during use. 15. The apparatus of claim 14, said translatable imaging system further comprising:a first element of said imaging system mounted to a first extension position of a first extendable rail;a second element of said imaging system mounted to a first extension position of a second extendable rail, said first element and said second element extended to opposite sides of the tumor during use. 16. The apparatus of claim 15, said translatable imaging system further comprising:a source element of a second imaging system mounted to a second extension position of said first extendable rail;a detector element of said second imaging system mounted to a second extension position of said second extendable rail, said source element and said detector element extended to opposite sides of the tumor during use, said second imaging system using a distinct imaging technology from said first imaging system. 17. The apparatus of claim 14, said translatable imaging system configured to translate along a linear axis independent of rotation of said gantry about said gantry rotation axis. 18. The apparatus of claim 14, said gantry support further comprising:a rotatable gantry support. 19. A method for both imaging a portion of a sample and treating the sample using positively charged particles, comprising the steps of:providing a gantry support connected to a gantry, said gantry connected to at least a portion of a beam transport system;transporting the positively charged particle beam through said portion of said beam transport system;rotating said gantry and said portion of said beam transport system about a gantry rotation axis;translating a translatable imaging system parallel to an axis perpendicular to the gantry rotation axis;imaging the portion of the sample using said translatable imaging system; andtreating the sample using the positively charged particle beam. 20. The method of claim 19, at a common time of said step of rotating moving said portion of said beam transport system, said step of translating moving said imaging system. |
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description | The present invention relates to the technical field of large medical device, and in particular, to a multi-purpose radiation therapy system. Medical equipment for radiation therapy treats tumorous tissue with high energy radiation. Two types of radiation therapy methods are generally employed for radiation therapy, namely, stereotaxic multi-source focusing radiation therapy method and adaptive intensity modulated radiation therapy (IMRT) method. Regarding the stereotaxic multi-source focusing radiation therapy method, a plurality of radioactive rays are focused to one focal point (namely, the target region), so that high-dose irradiation is performed on the tumor which is in the target region. This multi-source focusing radiation therapy method may be adopted to perform high-dose irradiation for tumor tissues, while radiation damage for surrounding tissues is small. This multi-source focusing radiation therapy method, with a precise therapeutic property, has a very good therapeutic effect for intracranial tumors or head and neck tumors. However, for a body tumor that has a complicated shape or that is large, the foregoing multi-source focusing radiation therapy method has its limits, and the conformal knife radiation therapy method is required. The conformal knife radiation therapy method adopts a single radioactive source, which is conformally processed and enables a distribution shape of a radiation dose region to be identical with or the same as the shape of the tumor in three dimensions, thereby avoiding or decreasing irradiation for normal tissues. In addition, the radiation dose in the dose region is uniformly distributed. Currently, there is no radiation therapy device that can integrate the stereotaxic multi-source focusing radiation therapy method with the adaptive intensity modulated radiation therapy method. In other words, the current radiation therapy devices cannot implement both accurate multi-source focusing therapy and conformal therapy on one device. For patients, different therapy methods cannot be selected for different tumors or a same tumor on a same device. The present invention provides a medical device that integrates stereotaxic multi-source focusing radiation therapy with conformal radiation therapy, which can implement both accurate focusing therapy and adaptive intensity modulated therapy on one radiation therapy device. An embodiment of the present invention provides A multi-purpose radiation therapy system, comprising a base, a movable couch, a rotatable gantry, and a radiotherapeutic apparatus, the movable couch and the rotatable gantry being arranged on the base, and the radiotherapeutic apparatus being movably installed on the rotatable gantry, wherein, the radiotherapeutic apparatus comprises a multi-source focusing radiotherapeutic unit and an adaptive intensity modulated radiotherapeutic unit and the multi-source focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit are distributed at both sides of a rotatable gantry axis, and are connected via respective arc guide rail. Preferably, an angle from the multi-source focusing radiotherapeutic unit to the adaptive intensity modulated radiotherapeutic unit relative to the axis center of the rotatable gantry is continuously adjustable between 30 degrees and 180 degrees. Preferably, the multi-source focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit respectively swings around a focal point on a rotatable gantry axial plane, and a swinging angle is in a range of 0 to ±47.5 degrees. Preferably, the rotatable gantry is 360-degree rotatable around the rotatable gantry axis in a continuous or reciprocal manner, to drive the multi-source focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit that are connected to the rotatable gantry to continuously or reciprocally rotate 360 degrees around the rotatable gantry axis. Preferably, an incident angle of the multi-source focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit exceeds 2π. Preferably, the multi-purpose radiation therapy system further comprises a dynamic image guide system, a set of stereo imaging apparatus or two sets of stereo imaging apparatuses at a fixed angle are installed on the rotatable gantry with the same focal point. Preferably, an angle between two sets of imaging apparatuses in the stereo imaging apparatuses is in a range of 20 degrees to 160 degrees. Preferably, the multi-source focusing radiotherapeutic unit comprises a plurality of radioactive sources, a precollimator, and a movable collimator, and rays of the radioactive sources pass through the precollimator and the movable collimator, and are focused on one point, to form a focused field. Preferably, the movable collimator is provided with apertures in different size, and the movable collimator is configured to switch the apertures to change a size and a shape of the focused field. Preferably, the adaptive intensity modulated radiotherapeutic unit comprises a radioactive source, a precollimator, and a multi-leaf collimator. Preferably, the radioactive source of the adaptive intensity modulated radiotherapeutic unit is a single cobalt source or an X ray generator having an intensity greater than 4 mV. By adopting the radiation therapy system of the embodiment of the present invention that integrates the stereotaxic multi-source focusing therapy with the adaptive intensity modulated therapy, the radiation therapy device can implement both the accurate focusing therapy and the adaptive intensity modulated radiation therapy. To make the objective, technical solution, and advantages of the present invention more clear, the following section describes the technical solution of the present invention in combination with the accompanying drawings. It should be understood that the embodiment described here is only exemplary one for illustrating the present invention, and is not intended to limit the present invention. For a better understanding of the technical solution of the present invention, the applicant describes a multi-purpose radiation therapy system of the embodiment of the present invention by using detailed implementation manners of FIG. 1 and FIG. 2. FIG. 1 is a schematic diagram of a radiation therapy system that integrates stereotaxic multi-source focusing radiation therapy structure with adaptive intensity modulated radiation therapy structure according to an embodiment of the present invention. As shown in FIG. 1, the radiation therapy system comprises a base, a rotatable gantry, a radiotherapeutic apparatus, and a treatment couch. The base supports the whole radiation therapy system, and plays a role of bearing the whole radiation therapy system and a role of fixation. The treatment couch is arranged on the base, and is movably connected to the base, e.g. by using screws and pins. The treatment couch is used to support and position a patient, and can accurately deliver the patient to a specified position for treatment. The rotatable gantry is further arranged on the base, and is connected to the base by a rolling support. The rotatable gantry rotates around an axial line, which is defined as a rotatable gantry axis X, by means of, e.g. gear drive. The radiation therapy system further comprises a core portion, namely, the radiotherapeutic apparatus. In one embodiment of the present invention, the radiotherapeutic apparatus involves two types of radiotherapeutic apparatuses, namely, a focusing radiotherapeutic apparatus and an adaptive intensity modulated radiotherapeutic apparatus. More specifically, the radiotherapeutic apparatus includes a focusing radiotherapeutic unit and an adaptive intensity modulated radiotherapeutic unit. The focusing radiotherapeutic unit may perform Stereotaxic Radiosurgery (SRS) or Imaging Guide Radiation Therapy (IGRT). The adaptive intensity modulated radiotherapeutic unit may perform 3-Dimensional Conformal Radiation Therapy (3D-CRT), or Intensity Modulated Radiation Therapy (IMRT), or Stereotactic Body Radiation Therapy (SBRT), or Imaging Guide Radiation Therapy (IGRT). The focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit are distributed at both sides of the rotatable gantry axis X. Because the rotatable gantry rotates around the rotatable gantry axis X (i.e. the gyration center), the radiotherapeutic apparatuses are driven to continuously or reciprocally rotate 360 degrees around the rotatable gantry axis X. In addition, the focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit are connected to the rotatable gantry and movable along an axial direction of the rotatable gantry, via respective arc guide rail. In this way, the radiotherapeutic apparatuses may continuously swing around a focal point on a rotatable gantry axial plane, and a swinging angle is in a range of 0 to ±47.5 degrees, so as to implement non-coplanar focusing or conformal therapy with different incident angles, thereby carrying out tumor therapy more flexibly and effectively. Further, regarding the placement position of the focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit, an included angle from the focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit to the axis is continuously adjustable between 30 degrees and 180 degrees. Since the radiotherapeutic apparatuses can make a continuous incident angle change of maximum ±47.5 degrees and a central rotation of 360-degree winding, a treatment incident angle of the system may exceed 2π. The focusing radiotherapeutic unit further comprises a plurality of radioactive sources, a movable collimator, and a precollimator. In the embodiment of the present invention, the radioactive sources adopt cobalt-60, and gamma rays generated by the cobalt-60 pass through the precollimator and the movable collimator, and are focused on one focal point. As such, a focused field, namely, a high-dose region for therapy, is formed. The movable collimator is provided with a plurality of apertures in different size, and the movable collimator is moved while aligning with the theradioactive sources. The movement of the movable collimator is performed to switch the apertures, so as to change a size and a shape of the focused field. As such, the focusing radiotherapeutic unit can be used to implement accurate therapy with a small field size and a high dose. The adaptive intensity modulated radiotherapeutic unit comprises a radioactive source, a precollimator, and a multi-leaf collimator. In the embodiment of the present invention, the radioactive source may be a single cobalt source or an X ray generator having an intensity greater than 4 mV. The radioactive source cooperates with the a multi-leaf collimator to implement different field shapes on a treatment plane, so as to implement three-dimensional adaptive intensity modulated irradiation. The multi-leaf collimator is implement with generally used technology, and details will not be described in the embodiment of the present invention. In addition, the radiation therapy system of the present invention further comprises a dynamic image guide system (IGS), In this embodiment, the dynamic image guide system (IGS) is stereo imaging apparatus, and one or two sets of stereo imaging apparatuses are assembled on the rotating rotatable gantry, and focus to the same focal point of the focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit. Generally, each set of the stereo imaging apparatus includes an X-ray generator and an image detection and acquisition system. Accordingly, one or two sets of X-ray imaging apparatuses are installed on the rotatable gantry, to perform real-time detection of a body position and a focus space position of a patient. Space position compensation is performed for the treatment couch and the radiotherapeutic apparatuses according to a detection result, so as to ensure high-precision orientation during treatment and implement accurate radiation therapy. When two sets of X-ray imaging apparatuses are adopted, an included angle of two sets of imaging apparatuses is in a range of 20 degrees to 160 degrees. In the embodiment of the present invention, the multi-source focusing radiotherapeutic unit and the adaptive intensity modulated radiotherapeutic unit are integrated into one radiation therapy system, which has a great advantage for some special tumor focuses where two manners of multi-source focusing and intensity modulation are required simultaneously or separately for treatment. In the radiation therapy system, the adaptive intensity modulated radiotherapeutic unit and the multi-source focusing radiotherapeutic unit may be simultaneously or separately used for irradiation therapy with one positioning, to implement two types of combined radiation therapy, errors caused by multiple times of positioning are reduced, and radiation therapy precision and speed are improved, thereby improving quality and efficiency. The above descriptions are merely a preferred embodiment of the present invention, but are not intended to limit the present invention. Any modification, equivalent replacement, or improvement made without departing from the spirit and principle of the present invention should fall within the protection scope of the present invention. |
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summary | ||
060463740 | description | DETAILED DESCRIPTION a. Overview PA0 b. Grout Constituents The present invention provides a novel method for the encapsulation of radioactively "hot" reactor vessels and other components, in which a comparatively thin, inexpensive shell of steel or similar material is constructed around the component, and the gap between the shell and component is filled with a fluid, cellular cement grout material which is laden with a metallic constituent to provide the shielding function. The size of the gap, and therefore the thickness of the grout layer, is selected based on the shielding qualities of the grout material to provide an effective radiation barrier in combination with the comparatively thin metal shell, and the fluidity of the grout material enables it to flow quickly and reliably through the interior of the shell and into any openings or irregular cavities on the component itself. The present invention also provides a novel cellular cement fill material which is easily and quickly pumped in place to provide an effective radiation barrier which reduces levels of both gamma and neutron radiation, and apparatus for production and injection of the same. The grout mixture is sufficiently fluid that this can also be pumped into containment vessels using the original piping system, if desired. The pumpable cellular grout material employs a mixture of cement slurry and bentonite gel, finished foam, and a metallic gamma-radiation absorptive material, such as a barium compound. The constituents are mixed, either in a batch process or using continuous-generation equipment, to form a micro-bubble structure in which the individual bubbles are encased in a layer of the bentonite gel, which enables the bubble structure to support the comparatively heavy barium or other heavy-metal compound in particulate or granular form. The barium or other heavy metal constituents provide an effective shield for absorbing the gamma radiation, while the bentonite component provides effective shielding to reduce neutron radiation, in addition to producing the strong, stable bubble structure which supports the heavy metal component. FIG. 1 shows a batch-type system 10 for producing and injecting the cellular grout material described above. As can be seen, the constituents of the grout, i.e., cement slurry, bentonite gel, metallic component, and finished foam are combined in a tub 12 by a mixer 14 to produce the pumpable cement grout material described above. The material is fed into a pump 16, via line 18, is discharged under pressure through an injection hose 20. Hose 20 is connected to an injection fitting 22 which is in communication with the interior of the barrier shell 26. As was noted above, in the present invention the barrier shell 26 is constructed of much lighter (e.g., 1") steel plate or similar material, rather than the very thick steel plate which is employed in the conventional method. This enables the shell to be constructed in a far quicker, more economical manner. Owing to the fact that in most embodiments of the present invention the shielding which is provided per inch of thickness by the of the fill material using the granular metallic constituent will be somewhat less than that of a solid layer of the metal, however, the shell 26 is constructed so as to form a comparatively large spaced gap 27 around the containment vessel or other component 30. The size of this gap is select to provide a thickness of the cellular fill material which, in combination with the comparatively thin metal shell, will provide an effective radiation barrier to reduce external radiation levels to within acceptable limits. Owing to its cellular nature and the bentonite component the grout 28 is highly fluid, so that as the grout is pumped into the shell 26 this fills the cavity around the containment vessel 30 or other component, forming a "blanket" completely around the component which absorbs both gamma and neutron radiation. The high degree of fluidity of the grout material permits the shell to be filled safely in one or more lifts, with the grout flowing upwardly to the next injection port 29 in the direction indicated by the arrows in FIG. 1, without fear of the flow stopping or becoming blocked. Once solidified, the bentonite and suspended metallic constituent in the blanket of grout reduce the amount of radiation which reaches the barrier shell 26 so that the relatively this layer of solid steel or other metal is sufficient to reduce the intensity of external radioactive emissions to acceptable levels. The outer metal shell also protects the solidified cement grout from impact damage and abrasion during transportation and storage, and in some embodiments the impact resistance of the barrier may be supplemented by the increased modulus of elasticity which the cellular cement grout can offer in certain formulations. Furthermore, unlike a liquid or loose granular material, the solidified grout cannot accidently escape from the containment shell, and therefore provides a permanent enclosure which reduces the chances of a radiation leak at the ultimate storage/disposal site. Still further, the solidified cement grout is virtually impervious to corrosion, which is an important advantage for very long term storage. As was noted above, the principal constituents of the grout provided by the present invention are cement slurry, bentonite gel, a metallic constituent and finished foam. Exemplary processes for the production and mixing of the same will be described in the following paragraphs. The cement slurry may be formed using any suitable hydraulic cement. Preferably, this is be formed using ordinary portland cement dust (with or without materials such as fly ash, superplasticizers, and other substitutes/additives which are known to those skilled in the art), mixed with water in a suitable ratio to meet both the fluidity and compressive strength requirements for the product. For example, a water-to-cement ratio of approximately 0.5:1 may be suitable for many embodiments; however, a wide range of water-to-cement ratios may be employed in various embodiments, depending on the characteristics which are desired for the slurry. In order to provide a more stable bubble structure, the cement slurry is preferably colloidally mixed to produce a finer, more fluid slurry which will more evenly coat the individual bubbles of the finished foam. This colloidal mixing may be performed using a high speed, high-shear pump which circulates the slurry material through a holding tank, until the cement particles are finely divided and evenly hydrated by the water molecules. The finished foam component, in turn, may employ any of the suitable foam-generation materials which are known to those skilled in the art for use in the production of cellular cements and grouts. The aqueous foam materials which are preferred for use in the present invention typically consist of a foam concentrate material similar to a surfactant which is diluted with water to form a foam solution having a suitable concentration. Examples of suitable foam concentrate materials include "Mearl Geocel Foam Liquid", available from the Mearl Corporation, Roselle Park, N.J. The foam concentrate-water solution is combined with air and passed through a foam conditioner to produce a finished foam material having the microbubble structure. Preferably, the foam density should be selected so as to produce a relatively strong bubble structure; for example a 2.25-4.0 percent water-concentrate solution mixed with air to produce a foam weight in a range from about 2.25-4.0 pounds per cubic feet (pcf) may be suitable, although concentrations and weights well outside of these ranges may be suitable for some applications. The production of finished foam and the mixing thereof with cement slurry to produce a foamed cement grout is more fully described in U.S. Pat. No. 5,419,632, which is incorporated by reference herein in its entirety, the inventor of which is the same as in the present application. As was noted above, the metal component serves to provide the grout material with the ability to absorb gamma radiation. The metal constituent will in most embodiments be in granular form and may include one or more types of metals or metal alloys. Preferably, the metal or metals selected will have characteristics which are benign to both personnel and the environment; amongst the many known gamma-absorbing metal known to those skilled in the art which are suitable for use in the present invention, barium and barium compounds are perhaps best in terms of effective shielding, but iron/steel has the advantage of economy, even though more material may need to be used. The type and amount of residual radiation present will generally determine the amount (i.e., the ratio) of the metal constituent to be included in the grout matrix; for example, in some embodiments an iron or barium component may be present in an amount equal to twice the weight of the cement or more. Granular barium sulfate is suitable for the purpose, and may be re-ground (e.g. using a ball mill) prior to mixing with the other constituents. A more economical component for most applications is smooth tumbled/polished steel shot, such as size 2-4 bird shot or a similar material, for example. Much finer grades of shot are less desirable for most embodiments, in that the greater surface area per weight of the material will tend to require an excessive amount of water and cement to coat, and will therefore result in lighter mixes and less effective shielding because not as much steel particulate can be included while still keeping the material fluid and pumpable. On the other hand, much larger particles such as buckshot-sized steel shot, for example, will tend to render the mix coarse and less fluid and pumpable, to the point where its flowability within the barrier shell may be impaired and the water may be squeezed out of the mix when pumping; also, the increased weight will render it difficult for the cellular bubble structure to maintain the larger particles in suspension. Furthermore, smooth, generally spherical particles are generally preferable, and sharp, angular particulate materials such as slags and blast metals should generally be avoided. Although iron/steel and barium materials are readily available and advantageous for use in the present invention, various other suitable heavy element materials capable of absorbing gamma radiation will occur to those skilled in the art. For example, tin, iron, thallium and lead, and their various compounds, can be used in the present invention to absorb gamma radiation, although most lead compounds exhibit undesirable toxicities. Antimony and antimony-containing compounds can also be used to provide the gamma radiation shielding component, and these are generally less expensive than tin-containing compounds. Suitable elements/compounds may also be found from within the rare earths, particularly the lanthanum series. Finally, the bentonite constituent serves three purposes. Firstly, it forms a bubble structure which is sufficiently strong to support the heavy metal (gamma shield) constituent. For example, if a heavy metal particulate were to be added to a conventional foamed cement grout (such as the materials which are conventionally used for geotechnical fills), the weight of the particulate would cause the components to separate and result in collapse of the bubble structure; by way of illustration, it should be noted that barium materials are on average approximately seven times as heavy per volume as water. The bentonite gel, however, serves to encapsulate the bubbles within the grout material, so that these are able to absorb and carry the heavy metal constituent. This aspect of the present invention is illustrated somewhat schematically in FIGS. 2A-2C. FIG. 2A is a "macro" view showing the cellular matrix of the grout material 28. As can be seen in the enlarged view presented by FIG. 2B, the matrix includes a multiplicity of fine, sometimes microscopic bubbles 32 formed by the finished foam and containing entrained air 36, each of which is surrounded by a relatively thick encapsulating layer of the combination bentonite gel/cement slurry 34; as was noted above, this material is highly fluid and "slippery," and essentially provides a strong, resilient capsule which enhances the resistance of the bubble to collapse/rupture and also makes it easier to pump (i.e., it requires lower pumping pressures). The heavy metal particles 37, such as the steel shot described above, are interspersed in and supported by the tough, resilient bentonite/cement coated bubbles, which also cushion and surround the metal particles to render the matrix more fluid and pumpable. Moreover, although not shown in FIG. 2B, a layer or coating of the bentonite/cement slurry will also be formed on the grains of metal as well, further enhancing fluidity and flowability. Also, as can be seen in FIG. 3C, in some embodiments the heavy metal constituent (such as a barium sulfate material, for example) may enter solution or otherwise be dispersed so as to itself form a layer 38 which surrounds and is supported by the bentonite gel/cement layer without damage to the underlying bubble structure. The bentonite gel can be mixed in any suitable manner, following the supplier's specifications; a suitable amount for many applications may be in the range of about 5-15% of the total, by weight. Preferably, to produce the gel-like consistency, the bentonite is colloidally mixed in a manner similar to that described above, by recirculation using a high-speed, high-shear colloidal mixing pump. The colloidal mixing superwets the particles of the bentonite, producing a high-grade gel material. Examples of a suitable colloidal mixing pumps include Series A ANSI centrifugal process pumps available from Hayward Gordon Company, Buffalo, N.Y. The second purpose of the bentonite component is to increase the fluidity of the cellular fill material. The bentonite gel forms a very smooth, highly fluid coating on the bubbles and metal particles alike, so that the material pumps easily and will flow long distances within the barrier shell and into various cavities and irregularities on the vessel or other component which is encased therein. The third function served by the bentonite gel is neutron radiation shielding. Bentonite materials are naturally occurring clay materials (usu. sodium montmorillonite) containing complicated silicon structures and water; in particular, as relates to the present invention, bentonites contain large amounts of light metals which are effective neutron radiation barriers. Because the bentonite is a major constituent or the grout, both by weight and volume (for example, in some embodiments the mix may contain bentonite in an amount equal to or greater than the portland cement), the fill material achieves a very significant drop in neutron radiation intensity. Yet another advantage provided by the bentonite component is that this results in a water-impervious grout fill, which again benefits long-term storage under various environmental conditions. It should also be understood that other suitable, gel-forming materials may in some embodiments be used in conjunction with the bentonite, or possibly in place of the bentonite if the material can alone or in combination with other materials provide the functions described above. For example, the "Polyox" material available from Union Carbide Company may be suitable for use in some embodiments of the present invention. As was noted above, the cellular grout material may be prepared by batch or continuous process. In the continuous process, the components may be added from their respective sources (e.g. the bentonite gel from the colloidal mixing tub, the cement/barium from another tub, and so forth) with foam solution and air being added at metered rates and then passed through a conditioner to form the bubble structure. FIG. 3 is a schematic view showing the filling of an exemplary nuclear containment vessel 40 (in this illustration, a steam generation vessel) itself, rather than a separate casing or enclosure which surrounds the vessel as described above, using the cellular cement grout of the present invention and the existing (i.e. original) valving of the component. As can be seen, the grout generation apparatus 10 may be mounted on a skid or other structure for transportation to the site. The grout material is delivered to the containment vessel by pump 42, through fill lines 44A,44B which are attached to fittings 46A,46B on the vessel. Vent lines 48A,48B are connected to other pipe connections 50A,50B on the vessel, so as to release the air and excess grout from the vessel. The released material, which may be contaminated from the contents of the vessel, is vented to a collection container 52. Upon completed filling, the vessel can be removed from the structure (e.g. the reactor building or the ship's hull) for subsequent transportation and disposal. In this embodiment, the placement of grout inside the containment vessel provides a very quick and economical method for reducing radiation levels external to the vessel, both to create a safer work environment in which personnel stay times are increased, and to permit final shielding using a comparatively thin external shell, with or without filling the void between the two as previously described. Also, it is possible that in some circumstances the radiation barrier fill within the containment vessel may reduce radiation levels sufficiently that a separate external shell will not be required for storage/disposal of the component. It is to be recognized that various alterations, modifications, and/or additions may be introduced into the constructions and arrangements of parts described above without departing from the spirit or ambit of the present invention. |
description | The present invention relates to an electron beam source including a bias control circuit that controls an electron gun and its bias electrode electric potential, as well as an electron beam exposure apparatus employing the electron beam source, and more particularly, to a high-throughput, high-accuracy electron beam source employed in a multi-beam drawing apparatus for use in the lithography step in the process of producing a semiconductor device such as a dynamic random access memory (hereinafter “DRAM”) having a capacity of 4-plus gigabits. Conventionally, an electron beam exposure apparatus has long been used to produce masks that are the templates for semiconductor devices such as DRAMs and MPUs (microprocessing units). In the last several years, as advances in the resolutions that these electron beam exposure apparatuses are capable of achieving have led to ever-denser semiconductor production processes, such electron beam exposure apparatuses have been applied to exposure devices used in the lithographic part of the production process. Currently, a so-called direct draw-type electron beam exposure apparatus has been proposed as an apparatus capable of being adapted to the design rules of 4-gigabit DRAMs and more, in which electron beams discharged from an electron gun are concentrated and directed by a deflecting system, an electromagnetic lens or the like at the point of concentration onto a semiconductor substrate so as to draw directly on the substrate. However, there are several problems with attempting to adapt such an apparatus to the semiconductor device mass production process, of which the most important is the drawing speed, with a high throughput of from several tens to several hundreds of times that of a so-called mask drawing unit required. As one means of solving this problem there is a so-called multi-beam-type electron beam exposure apparatus, in which the electron beam discharged from the electron gun is divided into a plurality of beams, for example 1,000, and arranged in the form of a matrix, and the beams used simultaneously to draw on the substrate specimen. A multi-beam electron beam exposure apparatus's ability to draw patterns simultaneously across a wide area using a plurality of electron beams can achieve dramatic improvements in through-put. This type of apparatus, that is, an electron gun that draws directly on the substrate using a plurality of electron beams arrayed over a broad area, requires a certain electron beam intensity in order to draw a pattern directly with the divided electron beams. Moreover, because a single electron beam is divided into a plurality of beams over a wide area, the angular current density distribution of the beam must be flat and therefore the emittance must also be large. Ordinarily the intensity and the emittance are conservative levels whose values are determined by the electron gun that serves as the light source of the apparatus. The basic structure of the electron gun employed in the conventional electron beam exposure apparatus involves a cathode in which the tip is shaped into a projection or sharpened to a point in order to increase the intensity, a Wehnelt for concentrating the electrons emitted from the cathode, to which is applied an electric potential lower than the voltage applied to the cathode, and an anode having a ground electrode. This tripolar type of electron gun is simple and easy to operate, and is widely used. However, because this type of electron gun is geared to increasing the intensity of the beam, and therefore it is very difficult to satisfy the required flatness of angular current density distribution over a wide area as described above. In order to solve the foregoing problem, for example, Japanese Laid-Open Patent Publication (Kokai) No. 2000-285840 discloses a tripolar structure in which the cathode (the electron discharge surface of which is shaped into a hemisphere), the bias electrode and the anode are aligned on the optical axis, with a distance from a center of an aperture in the bias electrode to the tip of the electron discharge surface of the cathode being approximately equal to or slightly greater than a radius of the aperture of the bias electrode, and as a result achieving the desired flatness of angular current density distribution while maintaining relatively high intensity. FIG. 6 shows one example comparing the angular current density distributions of the electron gun described above and an electron gun having the typical tripolar configuration. In the diagram, 101 indicates the angular current density distribution of the typical tripolar configuration and 102 indicates the angular current density distribution of an electron gun employing a hemispherical cathode. As can be seen from the diagram, the electron gun employing the hemispherical cathode has a larger flat portion than that of the electron gun employing the typical tripolar configuration. By optically processing and then spatially dividing this flat portion, the electron gun employing the hemispherical cathode can provide the type of multiple high-intensity electron beams described above. It should be noted that similar technologies are disclosed in other publications, for example, Japanese Laid-Open Patent Publication (Kokai) No. 9-129166, Japanese Laid-Open Patent Publication (Kokai) No. 9-180663 and Japanese Laid-Open Patent Publication (Kokai) No. 9-260237. When laying particular emphasis on accuracy in actual lithography, it is necessary simultaneously to draw while controlling any drift in the characteristics of the apparatuses and to increase the resolving power of the parameters. The electron gun described above uses only the flat portion of the angular current density distribution, with the remaining portions being cut off at an appropriate location by an aperture. Such shielding of the current generates heat, and is cooled by a variety of methods to control temperature changes. At the same time, the drawing accuracy demanded of apparatuses of this generation does not permit fluctuations in the performance of the electro-optical system due to slight changes in temperature. For this reason, it is preferable that the absolute amount of the cut-off current be as small as possible. In addition, as patterns have become finer, so too, the importance of the accuracy of various fine adjustments during drawing, such as proximity effect correction, auxiliary exposure for form correction, etc., has increased. For example, it is necessary to fine-tune the exposure energy and draw the pattern and auxiliary pattern so as to further improve form accuracy. In order to achieve this objective, it is necessary to increase the control resolving power of the exposure energy amount of the drawing apparatus as much as possible. From this standpoint, it is preferable to be able to constantly change the intensity of the electron gun to suit the process and other such drawing conditions. However, in the electron gun of the tripolar structure described above, intensity adjustment is accomplished by fine adjustment of either the anode voltage or the bias voltage. Of these two methods, that of finely adjusting the anode voltage is difficult to adapt to lithographic apparatuses because the energy of the electrons ultimately obtained changes. In addition, with the bias voltage adjustment method as well, typically, as the intensity changes the angular current density distribution also changes. FIG. 7 shows one example of the relation between the bias voltage V1 and the angular current density distribution, in a tripolar electron gun having a rounded cathode. As can be understood from the graph in FIG. 7, the intensity changes as the bias voltage V1 changes, and at the same time the angular current density distribution also changes, changing the flatness. If the electron beam is split under these conditions, the strength of the individual beams obtained from the split can be uneven, thus degrading the drawing characteristics of the apparatus. The present invention has been conceived as a solution to the above-described drawbacks of the conventional art, and has as its object to provide an electron beam source capable of changing the intensity of the electron beam while maintaining the current density angular distribution characteristics of the electron gun unchanged, and further, to provide a variety of electron beam-adapting apparatuses employing such an electron beam source, in particular a high through-put electron beam exposure apparatus. The above-described object of the present invention is achieved by an electron beam source comprising: a cathode having a hemispherical electron discharge surface; an anode disposed opposite the cathode and having a first aperture on an optical axis; a first bias electrode disposed between the cathode and the anode, to which is applied an electric potential that is lower than an electric potential to the cathode, and having a second aperture on the optical axis that is larger than the electron discharge of the cathode; and a second bias electrode disposed adjacent to the cathode as seen from the first bias electrode, to which is applied an electric potential that is lower than an electric potential to the cathode, and having a third aperture on the optical axis that is larger than the electron discharge surface of the cathode; and a controller that controls the electric potential applied to the first bias electrode and the electric potential applied to the second bias electrode. According to the above-described invention, the first bias electrode and the adjacent bias electrode are each controlled independently, and therefore can provide an electron beam source capable of continuously changing intensity while maintaining the angular current density distribution. Preferably, the above-described electron beam source has a structure such that, in a case where a radius of the second aperture of the first bias electrode is R1, a radius of the third aperture of the second bias electrode, a distance from the tip of the electron discharge surface of the cathode to a center of the second aperture of the first bias electrode is D1, and a distance from the center of the second aperture of the first bias electrode to a center of the third aperture of the second bias electrode on the optical axis is D2, a relation between R1 and R2 is 0.8R1≦R2≦1.2R1 and a relation between D1 and D2 is 0.8D1≦D2≦1.2D1. According to the above-described invention, the ability to maintain the angular distribution characteristics of the electric current density during intensity adjustment can be further enhanced. Preferably, the above-described electron beam source has a structure such that the controller, in a case where the electric potential E1 applied to the first bias electrode is a relative electric potential relative to the electric potential of the cathode and the electric potential E2 applied to the second bias electrode is a relative electric potential relative to the electric potential of the cathode, controls the electric potential applied to the first bias electrode and the electric potential applied to the second bias electrode so as to maintain a relation between E1 and E2 such that 0.9k≦|E1+E2|≦1.1k, where k is a positive constant. According to the above-described invention, the ability to maintain the angular distribution characteristics of the electric current density during intensity adjustment can be enhanced and can be controlled more easily as well. Preferably, the above-described electron beam source has a structure such that one or more bias electrodes, to which is applied an electric potential lower than the electric potential applied to the cathode, and which have an aperture on the optical axis larger that the electron discharge surface of the cathode, are aligned between the first bias electrode and the second bias electrode. According to the above-described invention, adjustment of such characteristics as the location of the crossover and the diameter of the crossover can be simplified. In addition, the above-described object of the present invention is achieved by an electron beam exposure apparatus comprising: the electron beam source described above; a formation system for applying an exposure beam having desired characteristics to the electron beam from the electron beam source; a deflection system for controlling a position of the exposure beam; and a projection system for concentrating the exposure beam on a substrate that is a specimen and drawing a desired pattern thereon. Thus, as described above, good maintenance of the angular density distribution of the current density during adjustment of the intensity of the electron gun described above can be adapted to a variety of apparatuses that employ electron beams. Other features and advantages of the present invention will be apparent from the following description when taken in conjunction with the accompanying drawings, in which like reference characters designate the same or similar parts throughout the figures thereof. Preferred embodiments of the present invention will now be described in detail in accordance with the accompanying drawings. (First Embodiment) FIG. 1 is a diagram showing the structure of an electron beam source according to a first embodiment of the present invention. In the diagram, reference numeral 1 designates a cathode that releases electrons, 2 and 3 designate heaters that heat the cathode 1, and 4 and 5 designate sub-electrodes that hold the cathode 1 and the heaters 2 and 3 as well as apply an electric potential to the cathode 1 and release an electric current to the heaters 2 and 3. Reference numeral 6 designates a first bias electrode, 7 designates a second aperture, 8 designates a second bias electrode, 9 designates a third aperture, and 10 designates an anode for accelerating the electrodes released from the cathode 1 to a desired energy level. Reference numeral 11 designates a first aperture and reference numeral 12 designates a fourth aperture. In addition, reference numeral 13 designates an insulator for insulating the sub-electrodes 4 and 5 and the bias electrode 6. Reference numerals 14, 15, 16 and 17 designate bias control circuits for applying a desired electric potential to the cathode 1 and the two bias electrodes 6 and 8. Reference numeral 18 designates a high-voltage power source for applying an electric current to the cathode 1 via the bias control circuits 14, 15, 16 and 17. Further, reference numeral 19 designates a heating power source for supplying an electric current to the heaters 2 and 3 and heating the cathode 1. It should be noted that the cathode 1, the first bias electrode 6, the second bias electrode 8, the anode 10 and their associated apertures are coaxially disposed about a central axis that is also the optical axis of the electron beam (hereinafter the “optical axis”). The cathode 1 has an emitter made of a cylindrically shaped single crystal of lanthanum hexaboride (LaB6), a tip of which is fashioned into a hemisphere. The lateral surfaces of the cathode are provided with flat notches that hold the cylinder symmetrically about its axis. Graphite heaters 2 and 3 are fitted into these notches, and are held in place from their outer sides by the sub-electrodes 4 and 5. The cathode 1 is provided with a negative high electric potential by the high-voltage power source 18 via the bias control circuits 14, 15, 16 and 17, which in the present embodiment is −50V. The first bias electrode 6 and the second bias electrode 8 are constructed so as to be capable of separately applying an electric potential equal to or lower than the electric potential applied by the bias control circuits 14, 15, 16 and 17 to the cathode 1. The cathode 1 releases electrons when heated by the supply of electric current to the heaters 2 and 3 from the heating power source 19. A beam 27 of freed electrons is accelerated by the anode 10 and, at the same time, is concentrated by the lens effect of the electric field created by the first bias electrode 6 and the second bias electrode 8 so as to form a crossover 28 in the vicinity of the first aperture 11 of the anode 10. The fourth aperture 12 of the anode 10 acts as a restricting opening, cutting off the peripheral portion of the electron beam 27 so that only the flat angular current density distribution central portion of the beam passes through. FIG. 2 shows the geometrical disposition of the cathode 1, the first bias electrode 6, the second bias electrode 8 and the anode 10 of the electron beam source of FIG. 1. In the diagram, a radius of the second aperture 7 of the first bias electrode 6 is R1, a distance from the center of the second aperture to the tip of the electron discharge surface of the cathode 1 is D1, and the relation between R1 and D1 is such that D1 can be adjusted within a range 1.0R1≦D1≦1.5R1. In addition, particularly good characteristics can be obtained when the radius R2 of the third aperture 9 of the second bias electrode is such that 0.8R1≦R2≦1.2R1, and in this embodiment R1=R2. Preferably, a radius r of the hemispherical electron discharge surface is half or less the radius R1 of the second aperture 7 of the first bias electrode 6, and therefore in this embodiment r=R1/6. The position of the second bias electrode 8 on the optical axis can be adjusted within a range such that a distance D2 between the center of the second aperture 7 of the first bias electrode 6 and the center of the third aperture 9 of the second bias electrode 8 is 0.8D1≦D2≦1.2D1. Further, a radius R3 of the first aperture 11 of the anode 10 is smaller than the radius R1 of the second aperture 7 of the first bias electrode 6. In this embodiment R3=R1/2. Relative electric potentials E1 and E2 relative to the electric potential of the cathode 1 are applied to the first bias electrode 6 and the second bias electrode 8, respectively, by the bias control circuit 17. In other words, potential differences E1 and E2 are applied between the first bias electrode 6 and the second bias electrode 8 and the cathode 1. FIG. 3 is a graph showing the relation between electric current density and beam half size, that is, between E1 and E2, on the one hand, and intensity on the other, when maintaining a relation k between E1 and E2 such that k (measured in volts) is a positive constant where |E1+E2|=k. In this embodiment, k=1050 [V]. As can be understood from the diagram, by changing E1 and E2 while satisfying the equation |E1+E2|=k described above, the total intensity can be changed without affecting the angular distribution characteristics of the current density even as the intensity changes. In practice, there is a certain permissible range in the margin of deviation from |E1+E2|=k described above, but it can be seen that the desired characteristics can be still obtained provided that the relation 0.9k≦|E1+E2|≦1.1k is maintained. As can be appreciated by those of ordinary skill in the art, in the foregoing embodiment, when further a third bias electrode 20 having a fifth aperture 21 is added on the optical axis between the first bias electrode 6 and the second bias electrode 8 as shown in FIG. 4, and accordingly the bias control circuit 17 applies an electric potential E3 to the third bias electrode 20 and the electric potential E3 is changed, a crossover position that is formed by concentrating the electron beam emitted from the cathode 1 can be easily adjusted. In the present embodiment, the electric potential E3 of the third bias electrode 20 is a value intermediate between the applied electric potentials E1 and E2, and a radius R4 of the fifth aperture 21 is larger than the radius R1 of the second aperture 7 and the radius R2 of the third aperture 9. At this time, the practical movable range of the crossover position is approximately 1 mm. Further, the movable range can be further expanded by varying the value of the radius R4 of the fifth aperture 21 and the position along the optical axis of the third bias electrode 20, although caution is required to ensure that such expansion does not affect the current density angular distribution characteristics. Moreover, by changing the cross-sectional shape of the third bias electrode 20 symmetrical about the axis, for example by making the shape a trapezoid, the characteristics of the electron beam thus formed can be fine-tuned. However, the radii of the apertures of the first and second bias electrodes as well as their relative positions and their electric potentials E1, E2 and E3 are all interrelated, and therefore must be determined separately for each such cross-sectional shape of the third bias electrode 20. It should be noted that although in the present embodiment a single crystal of lanthanum hexaboride (LaB6) is used as the source of the electrons, the same effect can be achieved with other electron beam sources as well, for example a metal with a high melting point such as tungsten (W). (Second Embodiment) A description is now given of an electron beam exposure apparatus employing the electron beam source of FIG. 1 according to a second embodiment of the present invention, with reference to FIG. 5. FIG. 5 is a diagram showing an electron beam exposure apparatus that directly draws a desired pattern on a sample using as a light source the electron beam source of FIG. 1. In the diagram, reference numeral 1 designates the electron beam source of the first embodiment described above that is the light source of the drawing apparatus of this second embodiment, and controls the intensity of the electron beam using the method described above, according to the operating state of the drawing apparatus or the process conditions of the work object. The location of the light source is virtually the same position as the location of the crossover 28 in FIG. 1. The electron beam discharged from this light source is condensed into a substantially parallel electron beam by a condenser lens 30 whose forward focal position is the light source location. The substantially parallel electron beam is then directed into a formation system 31 having a plurality of element electronic optic systems 32, and the parallel beam is divided into a plurality of beams. The element electron optic systems 32, which have blanking electrodes, form a plurality of light source intermediate images 33 in the vicinity of a blanking aperture 35, and at the same time can individually shield each of the plurality of electron beams at the blanking aperture 35 by activating each of the plurality of blanking electrodes. Next, magnetic field lenses 37 and 38 are arranged in a symmetrical magnetic doublet to form a projection system, with a restricting aperture 42. A distance between the magnetic field lenses 37 and 38 is equal to the sum of the focal lengths of the individual lenses, with the intermediate images 33 of the light source formed at the focal position of the magnetic field lens 37 and their images formed at the focal position of the magnetic lens 38. At this time, the ratio of the focal lengths of the magnetic field lenses 37 and 38 is the projection magnification. Further, since the magnetic fields of the magnetic field lenses 37 and 38 are set so as to work in opposing directions, Seidel aberrations other than the so-called five aberrations (i.e., spherical aberration, isotropic astigmatism, isotropic coma, curvature of field and on-axis chromatic aberration), as well as chromatic aberrations concerning rotation and magnification, are cancelled out. Reference numeral 40 designates a magnetic field deflector that satisfies convergent magnetic field and MOL conditions and 41 designates an electrostatic deflector for performing deflection with a magnetic field. These two deflectors are used to deflect the electron beams from the plurality of intermediate images and move the images of the plurality of intermediate images 33 in a plane atop the specimen 43. The magnetic field deflector 40 and the electrostatic deflector 41 are used differently depending on the distances the light source images move. Reference numeral 39 designates a dynamic focus coil for correcting a shift in focal position of a light source image caused by the deflection aberration that occurs when the deflectors are actuated. Reference numeral 36 designates a dynamic stig coil for correcting the isotropic astigmatism that occurs by the same process. Reference numeral 43 designates an XYZ stage for moving the specimen 43 in three dimensions. A plurality of laser interferometers provides precision control of the position and movement speed of the XYZ stage 44. Actual drawing is accomplished by projecting the images of the plurality of intermediate images 33 onto the specimen 43 with the projection system based on pattern data and scanning the surface of the specimen 43 using the magnetic field deflector 40 and the electrostatic deflector 41 while turning the plurality of electron beams on and off by actualizing the blanking electrodes of the plurality of element electronic optic systems 32 and moving the XYZ stage 44 so as to obtain a desired exposure pattern. It should be noted that the foregoing example is but one embodiment of the present invention, and a variety of configurations can be made without departing from the scope of the present invention when drawing a desired pattern with high resolution over a wide area. Moreover, although the present embodiment adapts the electron beam source of the present invention to an electron beam exposure apparatus, when projection of a high-intensity, high-emittance, wide-area electron beam is required, as can be appreciated by those of ordinary skill in the art, the present embodiment can be immediately adapted to a so-called multi-beam apparatus, in which one of the electron beams is divided into a plurality of electron beams for further application. The electron beam source of the present invention described above, by aligning two or more bias electrodes in the electron gun and arranging in a desired disposition, and further by controlling the electric potential applied to the individual bias electrodes, can adjust the intensity of the electron beam while maintaining current density angular distribution characteristics. Further, the electron beam exposure apparatus employing the electron beam source of the present invention makes it possible to both change the intensity of the electron gun that is the light source according to the work object and process conditions without affecting the current density angular distribution characteristics, as well as to achieve both high through-put and high density in a so-called multi-beam drawing apparatus that discharges a plurality of electron beams arrayed over a wide area. <Embodiment of Semiconductor Production> Next described is a semiconductor device manufacturing process employing the above-described electron beam exposure apparatus FIG. 8 shows a flow of an overall manufacturing process of a semiconductor device. In step 1 (circuit design), a circuit of a semiconductor device is designed. In step 2 (generation of exposure control data), exposure control data of the exposure apparatus is generated based on the designed circuit pattern. Meanwhile, in step 3 (wafer production), a wafer is produced with a material such as silicon. In step 4 (wafer process), which is called a pre-process, an actual circuit is formed on the wafer by a lithography technique. In step 5 (assembly), which is called a post-process, a semiconductor chip is manufactured using the wafer produced in step 4. Step 5 includes an assembling process (dicing, bonding), a packaging process (chip embedding) and so on. In step 6 (inspection), the semiconductor device manufactured in step 5 is subjected to inspection such as an operation-check test, durability test and so on. Semiconductor devices are manufactured in the foregoing processes and shipped (step 7). FIG. 9 shows a flow of the aforementioned wafer process in detail. In step 11 (oxidization), the wafer surface is oxidized. In step 12 (CVD), an insulating film is deposited on the wafer surface. In step 13 (electrode forming), electrodes are deposited on the wafer. In step 14 (ion implantation), ion is implanted on the wafer. In step 15 (resist process), a photosensitive agent is coated on the wafer. In step 16 (exposure), the circuit pattern is rendered (exposed) on the wafer by the above-described charged particle beam exposure apparatus. In step 17 (development), the exposed wafer is developed. In step 18 (etching), portions other than the developed resist image are removed. In step 19 (resist separation), unnecessary resist after the etching process is removed. By repeating the foregoing steps, multiple circuit patterns are formed on the wafer. 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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044850697 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings in detail, there is shown a moisture separator reheater 1 comprising a generally horizontal elongated cylindrical shell portion 3 having a head 5 disposed on each end thereof (only one is shown) which form end closures for the shell 3. A motive steam inlet nozzle (not shown) is in fluid communication with a pair of distribution manifolds 7 which are disposed horizontally within the shell 3. A plurality of distribution slots 8 are disposed along the length of the distribution manifold 7 to distribute motive steam evenly throughout the shell 3. A pair of cylindrically shaped plates 9 are disposed longitudinally in the central portion of the shell 3. The plates 9 have generally horizontally disposed parallel upper and lower margins. A plurality of generally flat upper plates 11 and 13 extend from the upper margins of the cylindrical plates 9 to the upper portion of the shell 3. A plurality of generally flat lower plates 15 extend from the lower margins of the cylindrical plates 9 to the lower portion of the shell. The plates 9, 11, 13 and 15 cooperate with the shell in a sealed relationship to form within the shell a central chamber 17 flanked by two side chambers 19 and 21. The lower plates 15 have openings 23, placing the side chambers 19 and 21 in fluid communication with the central chamber 17. Disposed adjacent the openings 23 is a chevron-shaped moisture separator or other moisture separating means 25. The lower plates 15 may also end at the separator 25. A motive fluid outlet nozzle 27 is disposed in the upper portion of the shell 3 in fluid communication with the central chamber 17. A plurality of long tubes 31 are disposed on a triangular pitch in a generally parallel array to form a generally round tube bundle 33 with flat top and bottom portions. At least one end of the tubes 31 extend through holes 35 in a tubesheet 37. The ends of the tubes are preferably seal welded to the tubesheet 37. In the preferred embodiment, U-shaped tubes are utilized, however, it is understood that straight tubes and a floating head could be used. A hemispherical head 39 is attached to the tubesheet 37 by welding or other means and has a dividing plate 41 disposed therein to form two chambers 43 and 45 within the head 39. The chamber 43 has a heating steam inlet nozzle 47 in fluid communication therewith and the chamber 45 has a drain nozzle 49 in fluid communication therewith. A manifold 51 is disposed in the chamber 43 and is in fluid communication with a plurality of tubes 31 having smaller radius bends. Also in fluid communication with the manifold 51 are two intersecting holes 53 drilled at right angles in the tubesheet. The holes 53 form a drain for draining fluid from the manifold 51. The manifold 51 is held in a sealed relationship with the tubesheet by studs 55 which fasten to holes in the tubesheet. A plurality of tube supports 57 preferably formed from plates are spaced longitudinally along the length of the tube bundle 33, normal to the tubes 31 to maintain the spacing between the tubes 31. Disposed along each side of the tube bundle 33 and fastened to the tube support plates 57 by welding or other means are a pair of arcuate plates 59, which are generally disposed along the length of the tube bundle 33 adjacent the generally round portion thereof. The arcuate plates 59 form a wrapper which subtends only the generally round sides of the tube bundle 33 and add rigidity to the tube bundle and prevent it from damage as it is installed and removed from the shell 3. Tongue and groove or other sliding sealed junctures 61 extend generally the length of the tube bundle 33 and are disposed between the arcuate plates 59 and the cylindrically shaped plates 9. The tongue and groove junctures 61 are formed from three flat bars 62, 63 and 64 slidably disposed adjacent each other. The central bar 63 being attached to one of the pairs of plates 59 and the other two bars 62 and 64 being attached to the other pairs of plates 9 to form a tongue and groove juncture which allows the tube bundle to slide in and out of the moisture separator reheater and form a seal between the wrapper plate 59 and the cylindrically shaped plates 9. The moisture separator reheater hereinbefore described advantageously provides a generally round tube bundle 33, which replaces square or rectangular tube bundle arrangements and improves the heat rate of the moisture separator reheater by about 20 to 50 BTU's per kilowatt hour. The arcuate plates 59 form a wrapper significantly improving the structural rigidity of the tube bundle 33 to enhance bundle alignment within the moisture separator reheater. The tongue and groove juncture 61 between the wrapper 59 and the cylindrically shaped plates 9 in the central chamber 17 facilitate rapid removal and replacement of the round tube bundle 33 and provide a seal which is not affected by surges in pressure of the motive fluid. |
claims | 1. A miniature mechanical shutter assembly comprising:a chamber;a tube member having an aperture formed therethrough, said tube member being mounted to the chamber in a manner that allows for translation and rotation;a pair of cap members disposed on opposing ends of said tube member, said pair of cap members supporting said tube member during said translation; anda plurality of magnet members, a first of the plurality of magnet members being disposed in a first end of said tube member, a second of the plurality of magnet members being disposed in a second end of said tube member opposite said first end, and a third of the plurality of magnets members being disposed external to said chamber, at least one of said plurality of magnet members being responsive to an electrical impulse to translate said tube member between an opened position and a closed position. 2. The miniature mechanical shutter according to claim 1 wherein said pair of cap members each comprises a glass liner between a respective one of said cap members and said tube member. 3. The miniature mechanical shutter according to claim 1 wherein said pair of cap members each comprises a Teflon member between a respective one of said cap members and said tube member. 4. The miniature mechanical shutter according to claim 1 wherein said plurality of magnet members are neodymium magnets. 5. The miniature mechanical shutter according to claim 1, further comprising:a drive circuit for exerting said electrical impulse upon said at least one of said plurality of magnet members. 6. The miniature mechanical shutter according to claim 5 wherein said drive circuit is an H-bridge circuit. 7. A miniature mechanical shutter assembly comprising:a chamber;a shutter member having an aperture formed therethrough, said shutter member being mounted to the chamber in a manner that allows for axial translation and axial rotation;a pair of cap members disposed on opposing ends of said shutter member, said pair of cap members supporting said shutter member during said axial translation and said axial rotation; anda plurality of magnet members, a first of the plurality of magnet members being disposed in a first end of said shutter member, a second of the plurality of magnet members being disposed in a second end of said shutter member opposite said first end, and a third of the plurality of magnets members being disposed external to said chamber, at least one of said plurality of magnet members being responsive to an electrical impulse to axially rotate said shutter member between an opened position and a closed position. 8. The miniature mechanical shutter according to claim 7 wherein said pair of cap members each comprises a glass liner between a respective one of said cap members and said shutter member. 9. The miniature mechanical shutter according to claim 7 wherein said pair of cap members each comprises a Teflon member between a respective one of said cap members and said shutter member. 10. The miniature mechanical shutter according to claim 7 wherein said plurality of magnet members are neodymium magnets. 11. The miniature mechanical shutter according to claim 7, further comprising:a drive circuit for exerting said electrical impulse upon said at least one of said plurality of magnet members. 12. The miniature mechanical shutter according to claim 11 wherein said drive circuit is an H-bridge circuit. |
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abstract | A method of repairing a nuclear fuel cell wall and tools useful for performing that repair are described. A repair tool may be used to align a jack near a region of a bent or distorted structural component of nuclear fuel cell and that jack may be used to apply a force to that structural component. Application of such a force may serve to bend the structural component of a nuclear fuel cell in a way to restore the structural component to its position before damage occurred. The repair tool includes a way of mounting that tool to a fuel cell, positioning elements to align the tool near a structural deformation or bent element and a jack that may be use to apply a force to at least one structural component in a fuel cell. |
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claims | 1. An elution tool for a radiopharmaceutical elution system comprising:an elution tool body having a top, an opposing bottom, an opening in the top, a vial chamber extending from the opening in the top toward the bottom that is sized and shaped for receiving an elution vial therein through the opening in the top, and an access opening extending through the bottom to the vial chamber, the access opening being aligned with a septum of the elution vial when the elution vial is received in the vial chamber, wherein the elution tool body comprises at least one of depleted uranium, tungsten, tungsten impregnated plastic, and lead;an elution tool lid secured to the elution tool body by a hinged connection adjacent the top of the elution tool body, the elution tool lid being rotatable at the hinged connection and relative to the elution tool body between a occluded position, in which the elution tool lid occludes the opening in the top of the elution tool body, and an exposed position, in which the elution tool lid does not occlude the opening in the top of the elution tool body to allow the elution vial to be inserted into and removed from the vial chamber, the hinged connection configured to allow linear, transverse movement of the lid relative to the body when the lid is in the occluded position, wherein the lid comprises at least one of depleted uranium, tungsten, tungsten impregnated plastic, and lead; anda latching mechanism for selectively and releasably locking the lid in the occluded position. 2. The elution tool set forth in claim 1, wherein the hinged connection comprises a slot defined in one of the body and the lid, and a hinge pin on the other of the body and the lid and received in the slot. 3. The elution tool set forth in claim 1, wherein the latching mechanism comprises a latching member on the lid, and a latching groove defined in the body, wherein the latching member is configured to be slidably receivable and removable from the latching groove by moving the lid transversely relative to the body. 4. The elution tool set forth in claim 3, wherein the latching mechanism further comprises a detent which releasably engages the latching member as the latching member enters the latching groove to inhibit inadvertent removal of the latching member from the latching groove. 5. The elution tool set forth in claim 1, wherein the hinged connection and the latching mechanism are opposed to one another relative to the lid. 6. The elution tool set forth in claim 1, wherein the body includes a seat on which the lid seats when the lid is in the occluded position, wherein the seat has an oblong periphery with a major axis, and the lid has a generally circular periphery. 7. The elution tool set forth in claim 1, further comprising a dispensing cap removably secured to the bottom of the body of the elution tool, the dispensing cap comprising a body having an access opening that is aligned with the access opening of the body of the elution tool when the dispensing cap is secured to the body of the elution tool, and a dispensing lid rotatably secured to the body of the dispensing cap for selectively opening and closing the access opening, wherein the dispensing lid comprises at least one of depleted uranium, tungsten, tungsten impregnated plastic. 8. The elution tool set forth in claim 7, wherein the dispensing cap further comprises at least one magnetic coupler for releasably securing the dispensing cap to the body of the elution tool. 9. The elution tool set forth in claim 8, wherein the dispensing cap defines a socket for receiving the bottom of the body of the elution tool, wherein the at least one magnetic coupler at least partially surrounds the socket. 10. The elution tool set forth in claim 9, wherein the body of the elution tool defines an annular coupler surface that is magnetically attracted to the at least one magnetic coupler of the dispensing cap. 11. The elution tool set forth in claim 10, wherein the dispensing cap further comprises a locking pin that is receivable in a locking cavity defined in the annular coupler surface of the body of the elution tool to inhibit rotation of the dispensing cap about the bottom of the body of the elution tool. 12. The elution tool set forth in claim 1, further comprising a storage cap removably securable to the bottom of the elution tool body, wherein the storage cap comprises a body and a radiation shield secured to the body, wherein the radiation shield is aligned with the access opening in the elution tool body when the storage cap is secured to the elution tool body, the radiation shield comprising at least one of depleted uranium, tungsten, tungsten impregnated plastic, and lead. 13. The elution tool set forth in claim 1, wherein the elution body is sized and shaped to be held in one hand of a user. 14. An elution tool for a radiopharmaceutical elution system comprising:an elution tool body configured to be held in one hand of a user, the elution tool body having a top, an opposing bottom, an opening in the top, a vial chamber extending from the opening in the top toward the bottom which is sized and shaped for receiving an elution vial therein through the opening in the top, and an access opening extending through the bottom to the vial chamber, the access opening being aligned with a septum of the elution vial when the elution vial is received in the vial chamber; anda dispensing cap removably securable to the bottom of the elution tool body, the dispensing cap comprising a dispensing cap body having a dispensing access opening that is aligned with the access opening of the elution tool body when the dispensing cap is secured to the elution tool body, and a dispensing lid rotatably secured to the dispensing cap body, the dispensing lid configured to selectively occlude and expose the dispensing access opening by rotating the dispensing lid across the dispensing access opening, wherein the elution tool body and the dispensing lid comprise at least one of depleted uranium, tungsten, tungsten impregnated plastic, and lead. 15. The elution tool set forth in claim 14, wherein the dispensing cap further comprises at least one magnetic coupler for releasable securing the dispensing cap to the body of the elution tool. 16. The elution tool set forth in claim 15, wherein the dispensing cap defines a socket for receiving the bottom of the body of the elution tool, wherein the at least one magnetic coupler at least partially surround the socket. 17. The elution tool set forth in claim 16, wherein the body of the elution tool defines an annular coupler surface that is magnetically attracted to the at least one magnetic coupler of the dispensing cap. 18. The elution tool set forth in claim 17, wherein the dispensing cap further comprises a locking pin that is receivable in a locking cavity defined in the annular coupler surface of the body of the elution tool to inhibit rotation of the dispensing cap about the bottom of the body of the elution tool. 19. The elution tool set forth in claim 14, wherein the dispensing cap body has a density less than the density of the dispensing lid. 20. The elution tool set forth in claim 19, wherein the elution body is sized and shaped to be held in one hand of a user. 21. The elution tool set forth in claim 1, wherein the hinged connection comprises a slot defined in one of the body and the lid, and a hinge pin on the other of the body and the lid and received in the slot, the slot being elongate in a direction transverse to an axis about which the lid is rotatable. |
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046363637 | claims | 1. Installation for conditioning radioactive wastes with a binding agent to form a wastes/binding agent mixture for ultimate storage comprising, a premixer in which a binding agent and flowable wastes are mixed, a tank for the flowable wastes connected through a throughput measuring device to the premixer, a silo for the binding agent connected through a throughput measuring device to the premixer, a receiver main tank for liquid wastes connected to a throughput measuring device to a throughput-mixer downstream from the entrance of the premixer mixture into the throughput-mixer, for mixing the liquid wastes with the mixture containing the binding agent discharged from the premixer, a discharge transport downstream from the throughput-mixer for filling barrels with conditioned radioactive wastes from the throughput-mixer for ultimate storage; the combination therewith of (a) said throughput-mixer having a vertical throughput direction, (b) said premixer arranged at the upper end of the throughput-mixer and provided with a movable member, and that (c) movable parts of the throughput-mixer which are subject to wear are less than 700 mm long and less than 500 mm wide. 2. Installation according to claim 1, wherein the volumetric capacity of the throughput-mixer is less than 5 liters. 3. Installation according to claim 1, wherein the diameter of the throughput-mixer is less than 200 mm. 4. Installation according to claim 2, wherein the diameter of the throughput-mixer is less than 200 mm. 5. Installation according to claim 1, including a port at the input side of the discharge transport for the introduction of spent lubricants. 6. Installation according to claim 5, wherein the discharge transport is constructed for continuous throughput, and variable pressure. |
description | This application is a divisional application of U.S. patent application Ser. No. 12/638,138, filed Dec. 15, 2009, entitled “Upper Internals Arrangement for a Pressurized Water Reactor,” which application claims priority to provisional patent Application Ser. No. 61/138,155, filed Dec. 17, 2008. 1. Field of the Invention This invention relates to water cooled nuclear reactors, and more particularly, to pressurized water reactors having in-core instrumentation (in-core instrument thimble assemblies) that enter the reactor vessel through penetrations from the top of the reactor vessel and are used to monitor the neutron activities and coolant temperature within the core fuel assemblies. 2. Description of Related Art Many water cooled nuclear reactors utilize a core of vertically positioned filet assemblies within a reactor vessel. To monitor the neutron activities and coolant temperature within the core fuel assemblies, movable in-core instrumentation, such as movable neutrons detectors, conventionally enter the core from penetrations in the bottom of the vessel. In a few instances in the past, leakage occurred at the penetrations at the bottom of the vessel which presented significant repair problems. Accordingly, it would be desirable to have all of the in-core instrumentation access the core through penetrations from the top of the reactor vessel. Additionally, fixed in-core neutron detectors have been employed that reside in the fuel assemblies during reactor operation. In addition to fixed in-core instrumentation that enter through penetrations in the bottom of the vessel, there are fixed in-core instrumentation that enter through penetrations in the top of the vessel. In this latter configuration, each in-core instrument thimble assembly is totally enclosed in a guide path composed of tubing. The lower portion of this guide path extends down into the fuel assembly. However, even the fixed in-core neutron detectors have to be withdrawn from the fuel assemblies before the reactor core can be accessed for refueling operations. Thus, it is therefore necessary to provide structure which can satisfactorily guide and protect the in-core instrumentation entering from the top of the vessel and mitigate the potential for leakage. Guidance for the instrumentation is needed through the area above the upper core plate, which is just above the fuel assemblies, to an elevation above the upper support plate which is spaced from and sits above the upper core plate, so that the in-core instrumentation can be withdrawn so its lower most extremity is at least at or about the mid plane of the upper core plate. This is necessary so that the upper internals can be removed to access the core for servicing, such as refueling. The existing upper support columns are available in between the upper core plate and upper support plate assembly to provide such guidance. However, presently there is no support for the instrumentation above the upper support plate assembly through which the in-core instrumentation has to be withdrawn to clear the bottom of the upper core plate. Accordingly, a new structure is needed that will provide guidance and protection for the in-core instrumentation in an elevation above the upper support plate assembly without impeding coolant flow in the upper internals during reactor operation. This invention provides support for the in-core instrumentation above the upper support plate when the in-core instrumentation is withdrawn from the core. The design of this invention provides a support system for the upper internals in-core instrumentation. Furthermore, the design of this invention minimizes additional disassembly requirements to remove and install the upper internals guide tubes in the event maintenance of the guide tubes is required. As previously noted, it is desirable to route the in-core instrumentation through the upper reactor head rather than the bottom of the reactor vessel. The in-core instrumentation routed through penetrations in the reactor head have to pass through the upper internals package to gain access to the instrumentation tubes centrally located within the fuel assemblies within the core. The upper internals package includes: an upper core plate which sits over the fuel assemblies; an upper support plate which is spaced above and over the upper core plate and attached to either the reactor vessel or the head; and hollow support columns which extend between the upper core plate and the upper support plate and are aligned with holes in both the upper core plate and the upper support plate, with the holes in the upper core plate communicating with the instrumentation tubes within the fuel assemblies. In accordance with this invention, an axially slidable sleeve extends through an upper end in at least some of the support columns which are aligned with corresponding instrumentation tubes. The axially slidable sleeves are extendable from the corresponding support columns through openings in the upper support plate to an area above the upper support plate at an elevation that is sufficient to shield the in-core instrument thimble assemblies in their withdrawn position. Preferably, the upper internals package includes an instrumentation grid assembly positioned above the upper support plate, that extends over each of the slidable sleeves. The instrumentation grid assembly has openings through which the slidable sleeves extend at least partially through with an upper portion of the slidable sleeve attached to the instrumentation grid assembly. The instrumentation grid assembly is configured to be movable in an axial direction to slide each of the slidable sleeves within the corresponding support columns in unison. A plurality of guide studs axially extend from an upper surface of the upper support plate and through corresponding openings in the instrumentation grid assembly, for laterally supporting the instrumentation grid assembly as it moves axially. Preferably, at least some of the guide studs are spaced around the perimeter of the instrumentation grid assembly. In one embodiment, there are approximately four guide studs substantially equally spaced around the perimeter of the instrumentation grid assembly. In one preferred embodiment, the slidable sleeves comprise a plurality of concentric telescoping tubes that extend between the instrumentation grid assembly and the corresponding support column. Preferably, a spiral spring extends around an inner most one of the concentric telescoping tubes below an attachment of the slidable sleeve to the instrumentation grid assembly between the attachment of the slidable sleeve to the instrumentation grid assembly and another of the telescoping tubes. The spring provides a holddown force on the telescoping tubes when the instrumentation grid assembly is in a lower most position, to prevent vibration during reactor operation. Desirably, one end of the spring extends at least partially into the opening in the instrumentation grid assembly through which the slidable sleeve extends and another end of the spring extends axially below the opening in the instrumentation grid assembly. The lower end of the spring is preferably surrounded by a can housing that is slidably mounted within the instrumentation grid assembly opening. Desirably, an upper portion of the can housing is captured within the opening of the instrumentation grid assembly to restrain the spring between the instrumentation grid assembly opening and the bottom of the can housing. In one embodiment, a lower portion of an inner most telescoping member of the slidable sleeve is enlarged and restrained below a narrowed opening within an upper portion of a surrounding member of the slidable sleeve so that lower portion of the inner most member of the slidable sleeve is captured within the opening of the surrounding member. Preferably, the slidable sleeve extends axially to at least an elevation above the upper support plate that will support the in-core instrument thimble assembly when the in-core instrument thimble assembly is raised, to at least the mid plane of the upper core plate, without the in-core instrument thimble assembly extending above the slidable sleeve when the reactor is shut down and the core is to be accessed. Desirably, the slidable sleeve extends above the upper support plate for at least 15.4 feet (47 meters). Furthermore, the invention contemplates a nuclear electric power generating facility having a pressurized water reactor nuclear steam supply system of the type described above. Furthermore, the invention contemplates a method of accessing a nuclear reactor core having a plurality of elongated fuel assemblies enclosed within a pressure vessel of a pressurized water reactor, wherein at least some of the fuel assemblies have at least one instrumentation tube axially extending therethrough for housing in-core instrumentation and the core is covered by an upper internals package that is sealed within the pressure vessel by a removable head. The upper internals package includes an upper core plate positioned over the fuel assemblies and an upper support plate spaced above and positioned over the upper core plate with a plurality of support columns extending axially between the upper core plate and the upper support plate with at least some of the support columns aligned with a corresponding one of the instrumentation tubes; the support columns aligned with the instrumentation tubes having a slidable sleeve that is movable within the support columns and extendable above the upper support plate. The method for accessing the core comprises removing the removable head from the pressure vessel; raising the slidable sleeves so that an upper portion thereof extends above the upper support plate; withdrawing the in-core instrumentation from the instrumentation tubes in the fuel assemblies so a lower most extremity of the in-core instrumentation is approximately at or above a mid point in a width of the upper plate; and removing the upper internals package to access the core. Preferably, the step of raising the slidable sleeves raises the sleeves all at one time. In that regard, desirably the upper internals package includes an axially movable instrumentation grid assembly positioned above the upper support plate and attached to an upper end of each of the slidable sleeves wherein the step of raising the slidable sleeves involves raising the instrumentation grid assembly. Referring now to the drawings, FIG. 1 shows a simplified nuclear reactor primary system, including a generally cylindrical pressure vessel 10 having a closure head 12 enclosing a nuclear core 14. A liquid reactor coolant, such as water, is pumped into the vessel 10 by pump 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18, typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown) such as a steam driven turbine generator. The reactor coolant is then returned through pump 16, completing the primary loop. Typically, a plurality of the above described loops are connected to a sealed reactor vessel 10 by reactor coolant piping 20. A conventional reactor design is shown in more detail in FIG. 2. As previously mentioned, though not shown in FIG. 2, in a conventional pressurized water reactor design, the movable in-core neutron detectors enter the core from the bottom of the reactor through tubes that extend from penetrations in the vessel bottom to the lower core plate 36 where they mate with the instrumentation tubes within the fuel assemblies. Furthermore, in such a traditional reactor design, the thermocouples that measure core temperature enter the upper head 12 through a single penetration and are distributed by a yoke or cable conduit, such as is shown in U.S. Pat. No. 3,827,935, to individual support columns 48 and thereby to various fuel assemblies. In addition to the core 14, comprised of a plurality of parallel, vertical co-extending fuel assemblies 22, for purposes of this description, the other vessel internal structures can be divided into the lower internals 24 and the upper internal 26. In conventional designs, the lower internals function to support, align and guide core components and instrumentation, as well as to direct coolant flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity), and support and guide instrumentation and components such as control rods 28. In the exemplary reactor shown in FIG. 2, coolant enters the vessel 10 through one or more inlet nozzles 30, flows downward about a core barrel 32, is turned 180° in a lower plenum 34, passes upwardly through a lower core support plate 36 upon which the fuel assemblies 22 are seated, and through and about the assemblies. The coolant flow through the core and surrounding area 38 is typically large, in the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second (6.1 meters per second). The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate 40. Coolant exiting the core 14 flows along the under side of the upper core plate 40 and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially through one or more outlet nozzles 44. The upper internals 26 can be supported from the reactor vessel 10 or the vessel closure head 12 and includes an upper support assembly 16 which is also referred to as the upper support plate. Loads are transmitted between the upper support plate 46 and the upper core plate 40 primarily by a plurality of support columns 48. A support column is aligned above a selected fuel assembly 22 and perforation 42 in the upper core plate 40 to provide access to elongated axial instrumentation tubes centrally located within each fuel assembly with the instrumentation tubes being co-extensive with the fuel assemblies' control rod guide thimbles. Rectilinearly movable control rods 28, typically including a drive shaft 50 and a spider assembly of neutron absorbing rods, are guided through the upper internals 26 and into aligned fuel assemblies 22 by control rod guide tubes 54. The guide tubes are fixedly joined to the upper support assembly 46 and connected by a split pin force fed into the top of the upper core plate 40. FIG. 3 provides an enlarged view of the upper internals package from it can clearly be seen that the control rods, which extend from the head 12 through the upper internals package and into the core below the upper core plate 40, are guided substantially over the entire distance by the control rod guide tubes 54 and the control rod guide tube extensions 88. However, the in-core instrumentation which are guided through the support columns 48 only receive support above the elevation of the reactor core between the upper core plate 40 and the upper support assembly 46. A substantial distance remains between the upper support assembly 46 and the head 12 over which the in-core instrumentation is exposed once it is withdrawn from the core. In accordance with this invention, some or all of the instrumentation is routed through one or more penetrations 56 in the head 12. This invention provides a structural modification to provide support for the in-core instrument thimble assemblies 52 in their withdrawn position where they extend above the upper support plate 46. FIG. 4 shows the full path of insertion of the in-core instrument thimble assemblies 52. The in-core instrument thimble assemblies 52 are routed through the reactor head penetration 12 and extend through the area above the upper support plate 46 and into an upper opening in the support columns 48. The in-core instrument thimble assemblies 52 then proceed down through the center of the support columns 48, through the upper core plate 40, through the thimble plugging device 39, through the fuel assembly upper nozzle 64 and into the fuel assembly instrumentation tubes 50. As shown in FIGS. 5 and 6, in accordance with this invention, the support columns 48 are provided with a slidable sleeve 60 that is extendable from the upper portion 62 of the support columns 48 into the area above the upper support plate 46 to support the in-core instrument thimble assemblies 52 when they are withdrawn from the fuel assemblies 22 to gain access to the core. In reactors such as the AP1000 supplied by the Westinghouse Electric Company LLC, Pittsburgh, Pa., the length of withdrawal required to raise the in-core instrument thimble assemblies 52 to the mid plane of the upper core plate 40 is typically larger than the height of the support columns 48 which leaves the highly irradiated upper portion of the in-core instrument thimble assemblies 52 exposed above the upper support plate 46, unguided and potentially subject to damage. Typically, in the AP1000 design the in-core instrument thimble assemblies 52 need to be raised approximately 185 inches (470 cm). The slidable sleeves 60 are designed to extend to support the exposed area of the in-core instrument thimble assemblies 52 above the upper support plate 46. As shown in FIGS. 6-9, the slidable sleeve 60 extends through an opening 66 in an instrumentation grid assembly 53 that extends horizontally over a substantial width of the upper support assembly 46. The instrumentation grid assembly 53 is supported to move axially on a plurality of guide studs 58 (shown in FIGS. 5 and 8) that are anchored to and extend upwardly from the upper support plate 46. Preferably, four guide studs 58 are equally spaced around the periphery of the instrument grid assembly 53. A cross section of the upper internals showing the instrumentation grid assembly 53 in its lower most position with the slidable sleeves 60 fully retracted within the corresponding support columns 48 is shown in FIG. 5 with more detail of the slidable sleeve shown in FIGS. 6 and 7. The slidable sleeve 60 comprises two tubes; an outer telescoping sleeve 68 and a fixed inner instrument tube 70 in which the in-core instrument thimble assembly 52 passes through. The instrument tube 70 extends slightly above the instrumentation grid assembly 53 and is anchored to the top surface thereof by the holddown plate assembly 84. The lower portion of the instrument tube 70 is telescopically received within an opening in the outer sleeve 68 and has an enlarged lower end 72 that is captured within a narrowed opening 74 within the outer sleeve 68 so that the instrument tube 70 cannot readily separate from the outer sleeve 68. A spiral spring 76 surrounds an upper portion of the instrument tube 70 between the holddown plate assembly 84 and a can housing 78 that surrounds a lower portion of the spring 76. The can housing has an enlarged upper portion 80 that is slidably mounted and axially movable within the instrumentation grid assembly slidable sleeve opening 66. The enlarged upper portion 80 of the can housing 78 is captured within the opening 66 by a lower annular lip 82. The lower portion of the can housing 76 has a lower lip 86 that captures the spring and seats upon the upper portion 74 of the outer sleeve 68 when the instrument grid assembly 53 is in its lower most position. With the instrumentation grid assembly 53 in its lower most position, the spring 76 exerts a force of approximately 50 pounds on the outer sleeve which prevents the sleeve from vibrating. FIG. 8 shows the cross section of the upper internals package previously shown in FIG. 5 with the instrumentation grid assembly 53 in its fully elevated position and the telescoping sliding sleeve 60 fully extended. FIG. 9 provides a more detailed cross sectional view in foreshortened form of the support column 48, sliding telescoping sleeve 60 and instrumentation grid assembly 53 and FIG. 10 shows a more detailed view of the telescoping sleeve 60 in its fully extended position. As can be observed in FIGS. 9 and 10, the inner instrument tube 70 extends until the enlarged end 72 abuts the narrowed opening 74 in the outer sleeve 68. As the inner instrument tube 70 extends the spring 76 decompresses and the spring can housing 78 moves down the opening 66 until the enlarged end 80 is captured by the lower lip 82 on the opening 66. The lower lip 86 on the spring can housing 78 prevents the spring 76 from moving further down the inner instrument tube 70. In the AP1000 design there are 42 in-core instrument thimble assemblies 52 each with its own telescoping sliding sleeve 60 that shields the highly irradiated portion of the in-core instrument thimble assemblies when they are raised above the fuel assemblies to service the core. After the head of the reactor has been removed, the polar crane within the containment can be employed to raise the instrumentation grid assembly 53 to its fullest, axially extended position where it can be locked in position on the guide studs 58 employing a locking mechanism such as the swing clamp 90. Raising the instrumentation grid assembly 53 simultaneously raises the in-core instrument thimble assemblies from each of the fuel assembly instrumentation tubes 50 so that the upper internals can then be removed as a package to access the core. Thus, this invention provides a means to protect and support the highly irradiated portion of the in-core instrument thimble assemblies used in a pressurized water reactor in-core instrumentation system white the instrumentation grid assembly is withdrawn during core servicing operations. This invention thus prevents the highly irradiated portion of the in-core instrument thimble assemblies from buckling in the event one or more of the assemblies meets some minor obstruction while the instrumentation grid assembly is being lowered to reinsert the in-core instrument thimble assemblies back into the fuel assemblies following completion of the servicing activities. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the breath of the appended claims and any and all equivalents thereof. |
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claims | 1. An EUV photon source, comprising: a plasma chamber filled with a gas mixture; multiple electrodes within the plasma chamber defining a pinch region and a central axis; a power supply circuit connected to the electrodes for delivering a main pulse to the electrodes for energizing the plasma around the central axis to produce an EUV beam output; a preionizer for ionizing the gas mixture in preparing to form a dense plasma around the central axis upon application of the main pulse from the power supply circuit to the electrodes; an ionization unit positioned along a beam path of the EUV beam outside of the plasma region for ionizing contaminant particulates along the beam path; and an electrostatic particle filter for collecting the ionized particulates. 2. The EUV source of claim 1 , wherein said ionizing unit generates a corona discharge. claim 1 3. The EUV source of claim 1 , further comprising one or more baffles along the beam path outside of the pinch region. claim 1 4. The EUV source of claim 3 , the one or more baffles for diffusing gaseous and contaminant particulate flow emanating from the pinch region. claim 3 5. The EUV source of claim 4 , the one or more baffles further for absorbing or reflecting acoustic waves emanating from the pinch region away from the pinch region. claim 4 6. The EUV source of claim 3 , further comprising a clipping aperture along the beam path outside of the pinch region for at least partially defining an acceptance angle of the EUV beam. claim 3 7. The EUV source of claim 6 , wherein said aperture comprises ceramic. claim 6 8. The EUV source of claim 6 , wherein said aperture comprises Al 2 O 3 . claim 6 9. The EUV source of claim 6 , wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. claim 6 10. The EUV source of claim 9 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 9 11. The EUV source of claim 10 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 10 12. The EUV source of claim 10 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 10 13. The EUV source of claim 6 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 6 14. The EUV source of claim 13 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 13 15. The EUV source of claim 13 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 13 16. The EUV source of claim 3 , wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. claim 3 17. The EUV source of claim 16 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 16 18. The EUV source of claim 17 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 17 19. The EUV source of claim 17 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 17 20. The EUV source of claim 3 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 3 21. The EUV source of claim 20 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 20 22. The EUV source of claim 20 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 20 23. The EUV source of claim 1 , further comprising a clipping aperture along the beam path outside of the pinch region for at least partially defining an acceptance angle of the EUV beam. claim 1 24. The EUV source of claim 23 , wherein said aperture comprises ceramic. claim 23 25. The EUV source of claim 23 , wherein said aperture comprises Al 2 O 3 . claim 23 26. The EUV source of claim 23 , wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. claim 23 27. The EUV source of claim 26 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 26 28. The EUV source of claim 27 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 27 29. The EUV source of claim 27 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 27 30. The EUV source of claim 23 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 23 31. The EUV source of claim 30 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 30 32. The EUV source of claim 30 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 30 33. The EUV source of claim 1 , wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. claim 1 34. The EUV source of claim 33 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 33 35. The EUV source of claim 34 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 34 36. The EUV source of claim 34 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 34 37. The EUV source of claim 1 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 1 38. The EUV source of claim 37 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 37 39. The EUV source of claim 37 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 37 40. An EUV photon source, comprising: a plasma chamber filled with a gas mixture; multiple electrodes within the plasma chamber defining a pinch region and a central axis; a power supply circuit connected to the electrodes for delivering a main pulse to the electrodes for energizing the plasma around the central axis to produce an EUV beam output; a preionizer for ionizing the gas mixture in preparing to form a dense plasma around the central axis upon application of the main pulse from the power supply circuit to the electrodes; and one or more baffles along a beam path outside of the pinch region. 41. The EUV source of claim 40 , the one or more baffles for diffusing gaseous and contaminant particulate flow emanating from the pinch region. claim 40 42. The EUV source of claim 41 , the one or more baffles further for absorbing or reflecting acoustic waves emanating from the pinch region away from the pinch region. claim 41 43. The EUV source of claim 40 , further comprising a clipping aperture along the beam path outside of the pinch region for at least partially defining an acceptance angle of the EUV beam. claim 40 44. The EUV source of claim 43 , wherein said aperture comprises ceramic. claim 43 45. The EUV source of claim 43 , wherein said aperture comprises Al 2 O 3 . claim 43 46. The EUV source of claim 43 , wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. claim 43 47. The EUV source of claim 46 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 46 48. The EUV source of claim 47 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 47 49. The EUV source of claim 47 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 47 50. The EUV source of claim 43 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 43 51. The EUV source of claim 50 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 50 52. The EUV source of claim 50 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 50 53. The EUV source of claim 40 , wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. claim 40 54. The EUV source of claim 53 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 53 55. The EUV source of claim 54 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 54 56. The EUV source of claim 54 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 54 57. The EUV source of claim 40 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 40 58. The EUV source of claim 57 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 57 59. The EUV source of claim 57 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 57 60. An EUV photon source, comprising: a plasma chamber filled with a gas mixture; multiple electrodes within the plasma chamber defining a pinch region and a central axis; a power supply circuit connected to the electrodes for delivering a main pulse to the electrodes for energizing the plasma around the central axis to produce an EUV beam output; a preionizer for ionizing the gas mixture in preparing to form a dense plasma around the central axis upon application of the main pulse from the power supply circuit to the electrodes; and a clipping aperture along a beam path outside of the pinch region for at least partially defining an acceptance angle of the EUV beam. 61. The EUV source of claim 60 , wherein said aperture comprises ceramic. claim 60 62. The EUV source of claim 60 , wherein said aperture comprises Al 2 O 3 . claim 60 63. The EUV source of claim 60 , wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. claim 60 64. The EUV source of claim 63 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 63 65. The EUV source of claim 64 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 64 66. The EUV source of claim 64 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 64 67. The EUV source of claim 60 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along the beam path of the EUV beam. claim 60 68. The EUV source of claim 67 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 67 69. The EUV source of claim 67 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 67 70. An EUV photon source, comprising: a plasma chamber filled with a gas mixture; multiple electrodes within the plasma chamber defining a pinch region and a central axis; a power supply circuit connected to the electrodes for delivering a main pulse to the electrodes for energizing the plasma around the central axis to produce an EUV beam output; a preionizer for ionizing the gas mixture in preparing to form a dense plasma around the central axis upon application of the main pulse from the power supply circuit to the electrodes, and wherein said power supply circuit generates the main pulse and a relatively low energy prepulse before said main pulse for homogenizing the preionized plasma prior to the main pulse. 71. The EUV source of claim 70 , further comprising a multi-layer EUV mirror disposed opposite a beam output side of the pinch region for reflecting radiation in a direction of the beam output side for output along a beam path of the EUV beam. claim 70 72. The EUV source of claim 71 , wherein the EUV mirror has a curved contour for substantially collimating the reflected radiation. claim 71 73. The EUV source of claim 71 , wherein the EUV mirror has a curved contour for substantially focusing the reflected radiation. claim 71 |
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047023910 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT A container 1 enclosing solidified melted material with highly radioactive content is disposed in a hollow circular containment cylinder 2 which is closed at its bottom by a bottom plug structure 4 and at its top by a top plug structure 3. Together these parts form a containment. Both plugs 3 and 4 are mounted to the cylinder 2 by means of threads 9 and 10. The top plug structure 3 consists of a lid 5 and a cover plate 6; the bottom plug structure 4 consists of a bottom 7 and a bottom plate 8, the bottom plate 8 and the cover plate 6 having a diameter slightly larger than the bottom 7 and the lid 5, so that they both project slightly radially outwardly. Lid 5 and cover plate 6, as well as bottom 7 and bottom plate 8, are mounted together in a special way as shown in FIG. 2. The containment cylinder 2, the lid 5, the cover plate 6, the bottom 7 and the bottom plate 8 all consist of fine-grain construction steel. The cover plate 6 has a gripping cavity 11 formed therein for engagement thereof by a lifting mechanism. The containment cylinder 2 has disposed therein around the body 1 a heat conductive centering sleeve 12 adapted to improve the heat transfer from the body 1 to the containment cylinder walls. The containment cylinder 2 is provided, by explosion plating, with a liner 13, for example, of titanium-palladium alloy, providing a corrosion protective barrier which extends axially beyond its bottom and top ends such that it forms axial projections 14 of a length about equal the thickness thereof. This projection 14 serves as a centering means for the top and bottom plug structures 3 and 4. The bottom 7 and the lid 5 are also provided, by explosion plating, with liners of the titanium-palladium alloy. The liners 15, 16 have about the same thickness as the liner 13. They project radially slightly beyond the bottom 7 or the cover 5 and have a diameter essentially corresponding to the inner diameter of the liner 13. When the bottom and top plug structures are mounted, the liners' 15 and 16 circumferential faces 17 are disposed adjacent the inner circumference of the projections 14 of the cylinder's outer liner 13. Additional titanium-palladium layers 18, 19 of the same thickness are explosion welded onto the inner sides of the cover plate 6 and the bottom plate 8. As may be seen from FIG. 2, where the layer 18 of the bottom plate 8 is shown in greater detail, the radius of this layer 18 is larger than that of the liner 15 of the bottom 7 by about a layer's thickness. The projection 20 so formed is disposed adjacent the projection 14 of the cylinder liner 13 and is flush therewith when the bottom plug is mounted. The arrangement at the top end of the containment is essentially the same, that is, the layer arrangement is identical and so is the welding procedure for the layers to be described below. The bottom plate 8 and the bottom 7 and also the cover plate 6 and the lid 5 are welded together before the bottom and top plug structures are threaded into the containment cylinder 2. For this purpose bottom plate 8 and bottom 7 and also cover plate 6 and lid 5 are placed together such that liner 15 and layer 18 and also liner 16 and layer 19 are disposed adjacent one another. They are then welded together at the separating seam 21 by a circumferential weld 22, which is formed by electron beam welding and which extends between the plating liners and layers to a depth of up to 15 mm. When the top and bottom plugs 3, 4 are now threaded into the containment cylinder 2, the liner 13 with its projection 14 overlaps the liner 15 and abuts the layer 18 and forms an additional seam 23 between the projections 14 and 20. This seam 23 is now also welded by means of a circumferential weld 24 which is formed as a smooth cosmetic weld such that the lid 5, the bottom 7 and the containment cylinder 2 are completely surrounded by explosion welded layers and liners of the titanium-palladium alloy with sections of normal steel such as the bottom plate 8 and the cover plate 6 being disposed adjacent the corrosion protection layers below the bottom 7 and above the lid 5. The corrosion protection liners and layers 13, 15, 16, 18, 19 therefore completely surround the containment 2, 3, 4. At the top and bottom end faces there are the cover plates 6 and 8 which consist of a material different from that of which the corrosion protection liners and layers consist and which have a diameter slightly larger than that of the containment cylinder 2. The bottom and cover plates consist of the same fine-grain construction steel as the containment cylinder walls 2, the bottom 7 and the lid 5 onto which the corrosion protection liners and layers are explosion welded. At the separating seam 21 the corrosion protection liners and layers 15, 18 and 16, 19 of the lid 5 and cover plate 6 and of the bottom 7 and the bottom plate 8, respectively, are welded together radially from the circumference thereof. The explosion welded layers 18, 19 of the bottom and cover plates 8, 6 have a larger diameter than the explosion welded liners 15, 16 of the bottom 7 and the lid 5; the liner 13 which is explosion welded onto the containment cylinder 2 has projecting end portions 14 which overlap the liners 15, 16 of the bottom 7 and the lid 5 and which are tightly welded from the outside, that is, circumferentially to the larger diameter layers 18, 19 of the bottom and cover plates 8, 6. For the manufacture of the sealed containment according to the invention, the manufacturing steps are as follows: (a) Explosion plating of the outer surfaces of the containment cylinder 2 and of the outer surfaces of the lid 5 and the bottom 7 with liners or layers of a corrosion inhibiting material such as a titanium-palladium alloy, PA1 (b) Explosion plating of the inner surfaces of the bottom and cover plates 8, 6 with layers 18, 19 of the same material, PA1 (c) Fitting together of the bottom 7 and the bottom plate 8 and also of the lid 5 and cover plate 6 by placing the explosion welded liners and layers 15 and 18 and also 16 and 19 adjacent one another and welding along the separation seam at the circumference thereof, PA1 (d) Threading the bottom plate structure 4 into the lower end of the containment cylinder 2, PA1 (e) Welding the two explosion welded liner 15 and layer 18 between the bottom 78 and the bottom plate 8 along their circumference together with the projection 14 of the explosion welded liner 13 protruding at the lower end of the containment cylinder 2, PA1 (f) Installing a heat conductive centering sleeve 12 into the containment cylinder 2, PA1 (g) Inserting the radioactive body 1 into the centering sleeve 12, PA1 (h) Threading the top plug structure 3 consisting of lid 5 and cover plate 6 from the top into the containment cylinder 2, and PA1 (i) Welding the outer edges of the explosion welded liner 16 and layer 19 between the lid 5 and cover plate 6 to the upwardly projecting end of the explosion welded liner 13 of the containment cylinder 2. LISTING OF REFERENCE NUMERALS 1--Container with radioactive content PA0 2--Container cylinder PA0 3--Top plug structure PA0 4--Bottom plug structure PA0 5--Lid PA0 6--Cover plate PA0 7--Bottom PA0 8--Bottom plate PA0 9--Thread PA0 10--Thread PA0 11--Grapping cavity PA0 12--Heat conductive centering sleeve PA0 13--Titanium-palladium liner PA0 14--Projection PA0 15--Titanium-palladium liner PA0 16--Titanium-palladium liner PA0 17--Circumferential faces PA0 18--Titanium-palladium layer PA0 19--Titanium-palladium layer PA0 20--Projection PA0 21--Separating seam PA0 22--Circumferential weld PA0 23--Seam PA0 24--Circumferential weld |
description | The present invention relates generally to liquid crystals, and more particularly to an apparatus and method for forming a liquid-crystal alignment layer by ion beam irradiation. In a liquid crystal display (LCD), the orientation direction of a liquid crystal is controlled with an alignment layer disposed on a substrate. The alignment layer is formed in such a manner that a thin-film made of polyimide or an inorganic material is irradiated with ion beams such that the bonds between atoms in the thin-film are broken. If the surface of the thin-film is subjected to alignment processing, that is, alignment treatment, effective in aligning the liquid crystal by ion beam irradiation, the distribution of the irradiation direction of the ion beams corresponds to that of the orientation direction of the liquid crystal. Hence, it is preferable that the spread of the ion beams be small and the ion beams be aligned in a predetermined direction. According to a conventional technique, as shown in FIGS. 1(a) and 1(b), ion beams 28 are applied to a thin-film 26 while a substrate 24 is being moved close to or away from an ion source 12. Since the ion beams 28 spread as described below, a mask (a shielding plate) 20 is used to apply only some of the ion beams 28 that are useful in forming an alignment layer to the thin-film 26 through a slit 22. In order to subject the entire surface of the thin-film to alignment treatment uniformly, it is ideal that ions are applied to the thin-film 26 in such a manner that the ions form parallel beams with uniform ion density. The orientation direction of a liquid crystal needs to be uniform over the alignment layer. The misorientation of the liquid crystal causes brightness or color non-uniformity in liquid crystal panels. Therefore, in order to manufacture a high-image-quality liquid crystal panel, the liquid crystal needs to be more uniformly oriented. In order to uniformly orient the liquid crystal, the ion beams emitted from the ion source need to have uniform density. The density of the ion beams is rendered uniform by controlling the density of gas in a plasma-generating chamber and the density of generated free electrons. As shown in FIG. 1, the ion source for generating the ion beams 28 usually includes a plurality of sheet-shaped grids 11. Each grid 11 has a plurality of outlets for emitting the ions. The ion beams 28 emitted through the outlets spread. In order to allow the ion beams to have uniform intensity or density distribution, the outlets have different sizes depending on the ion density in the ion source as disclosed in U.S. Pat. No. 6,849,858. If a region extending in the X direction in FIG. 1B, that is, in the direction perpendicular to the traveling direction of the substrate, is irradiate with the ion beams with different intensities, misorientation occurs in the thin-film 26 subjected to alignment treatment. U.S. Pat. No. 7,057,692 discloses that the ion beams 28 are aligned in one direction in such a manner that the shape of an edge 32a of the mask 20 that defines the slit 22 is varied or corrected depending on the orientation direction of the liquid crystal. The above conventional techniques have the following problems: a problem that an apparatus needs to be precisely manufactured, for example, the grid needs to be precisely machined such that the ion beams emitted from the ion source have uniform density and/or the mask edge needs to be precisely shaped such that the ion beams are aligned in one direction and a problem that the process time required to apply the ion beams to the irradiation region becomes long because the irradiation region is narrow. The present invention provides an apparatus and method for forming a liquid-crystal alignment layer by ion beam irradiation. The apparatus need not be extremely precisely manufactured, for example, a grid need not be extremely precisely machined or a mask edge need not be extremely precisely shaped. The apparatus and the method are useful in rendering the orientation of the alignment layer uniform without narrowing a region irradiated with an ion beam. Since the orientation thereof is uniform, a liquid crystal display with a uniform image quality, having no brightness or color non-uniformity, can be obtained. In one embodiment, the present invention provides an apparatus and method for forming a liquid-crystal alignment layer by ion beam irradiation. This apparatus and this method are capable of adjusting the pre-tilt angle of liquid crystals to a desired value. Since the pre-tilt angle of liquid crystals can be adjusted to a desired value, a liquid crystal display can be improved in display quality including a viewing angle property. According to an aspect of the present invention, an alignment layer-forming apparatus includes an ion source for applying a plurality of ion beams to a thin-film disposed on a substrate and a mask (a shielding plate), disposed between the substrate and the ion source, having a reflective face for reflecting the ion beams toward the substrate. The alignment layer is formed in such a manner that the ion beams are reflected toward a thin-film disposed on the substrate by the reflective face in a predetermined direction, that is, an orientation direction, whereby the orientation direction of a liquid crystal is uniformly aligned. The predetermined direction is the same as the orientation direction of the liquid crystal and the mask has a reflective face for varying the incident angle of the ion beams with respect to the thin-film depending on the predetermined pre-tilt angle. That is, the incident angle of the ion beams with respect to the thin-film is varied by reflecting the ion beams with the reflective face depending on the desired pre-tilt angle. The reflective face of the mask that is directed to a surface of the substrate is flat and is tilted with respect to the substrate surface such that the ion beams are reflected forward. Furthermore, the reflective face of the mask is tilted with respect to the traveling direction of the substrate, whereby the propagation direction of the ion beams having a spread angle is aligned in the orientation direction. The orientation direction may be parallel to the traveling direction of the substrate. The reflective face of the mask may be tilted with respect to the traveling direction of the substrate such that the tilt angle of the reflective face corresponds to the offset angle of the ion beams with respect to the predetermined direction, that is, the orientation direction. When the alignment layer-forming apparatus further includes a mask tilt control section for controlling the tilt angle of the reflective face with respect to the substrate, that is, the spatial tilt of the reflective face, the position of the reflective face of the mask can be precisely fine adjusted with the mask tilt control section. When the mask has a sawtooth-shaped surface and the reflective face is disposed on the sawtooth-shaped surface and is flat or curved, the ion beams can be reflected in a desired direction. The reflective face of the mask and the thin-film on the substrate, processed into the alignment layer, may be made of the same material. The mask may have end portions which have a tilt angle with respect to the traveling direction of the substrate and which correspond to a region irradiated with the ion beams having a large spread angle. Furthermore, the tilt angle of the reflective face with respect to the traveling direction of the substrate may be increased stepwise from a center portion to end portions of the mask that correspond to the region irradiated with the ion beams, whereby the ion beams can be aligned in the orientation direction depending on the spread angle of the ion beams. According to another aspect of the present invention, a method for forming a liquid-crystal alignment layer by processing a thin-film disposed on a substrate includes preparing a substrate having a thin-film processed into the liquid-crystal alignment layer; placing a mask, having a reflective face directed to the substrate, between the substrate and an ion source; conveying the substrate; irradiating the thin-film on the substrate with a plurality of ion beams emitted from the ion source; and reflecting the ion beams with the reflective face in an orientation direction by reflecting the ion beams between the thin-film and the reflective face. The method further includes adjusting the tilt angle of the reflective face of the mask with respect to the substrate. The liquid-crystal alignment layer can be formed and the orientation direction of a liquid crystal can be uniformly aligned in such a manner that the ion beams are reflected by the reflective face in the direction toward the thin-film on the substrate and in the orientation direction of the liquid crystal. The propagation direction of the ion beams may be the same as the traveling direction of the substrate, that is, the substrate may be conveyed in the direction away from the ion source. In this case, the reflective face of the mask is formed such that the spread angle of the ion beams is corrected. A method for allowing a liquid-crystal alignment layer to have a predetermined pre-tilt angle includes reflecting a plurality of ion beams with a reflective face of a mask toward a thin-film disposed on a substrate such that the incident angle of the ion beams with respect to the thin-film on the substrate corresponds to the predetermined pre-tilt angle. This allows the pre-tilt angle to be freely varied depending on optical design. According to the present invention, the spread angle (a difference in incident angle) or incident angle of a plurality of ion beams applied to a thin-film disposed on a substrate is controlled by the secondary reflection (even-numbered order reflection) of the ion beams by a reflective face of a mask, whereby a uniform alignment layer can be formed or the pre-tilt angle of a liquid crystal can be controlled. This allows a high-image-quality liquid crystal display having no brightness or color non-uniformity to be manufactured. Apparatuses and methods for forming alignment layers according to embodiments of the present invention will now be described with reference to the attached drawings. The orientation direction of a liquid crystal is defined as the direction in which the longitudinal axes of molecules of the liquid crystal are aligned in the X-Y plane when the liquid crystal molecules are projected onto a substrate 24. The orientation direction of the liquid crystal depends on the density distribution of ion beams 28 and Angle φ in FIG. 2, that is, Angle φ (an azimuth angle) between the Y axis (actually a line parallel to the Y axis) and each ion beam 28 projected onto the X-Y plane (a plane parallel to the substrate 24 or a mask 20). With reference to FIG. 2, Angle α (an elevation angle) represents the incident angle of the ion beam 28 with respect to the X-Y plane. With reference to FIG. 3, the ion beams 28 are emitted from an ion source and usually have a spread distribution, that is, the ion beams 28 are offset from a desired or target irradiation direction. The ion source usually has a large number of outlets or exit apertures for emitting a plurality of the ion beams 28, which travel in different directions. For example, as shown in FIG. 3, Point B on a thin-film 26 is irradiated with the ion beams 28 traveling from different directions. That is, the ion beams have a spread (between Point B and Point B″) or a narrow (between Point B and Point B′) and do not necessarily travel in parallel to each other. If Angle φ varies depending on locations on the thin-film, the liquid crystal molecules are not aligned in one direction. This causes brightness non-uniformity and/or color non-uniformity in liquid crystal display. If Angle φ varies depending on Points A, B, and C, present on the thin-film 26, irradiated with the ion beams 28a, 28b, and 28c, respectively, the orientation direction of the liquid crystal is nonuniform. When the Y axis in FIG. 2 is a desired orientation direction, Angle φ is preferably zero (0) for every location on the thin-film 26, that is, the desired orientation direction is preferably parallel to the Y axis. Angle φ is hereinafter referred to as an offset angle φ. It has been found that the ion beams 28, finally applied to locations on the thin-film 26, and having energy sufficient to break the bonds between atoms in a surface domain of the thin-film that is involved in orientation, affect the orientation direction of the liquid crystal. That is, it is effective that the ion beams 28 finally applied to all locations on the thin-film 26 are aligned in one direction. The present invention is characterized in that a mask has a reflective face directed to a substrate, ion beams that spread or are offset from a desired orientation direction are reflected forward by the reflective face such that the offset of each ion beam from the desired orientation direction is corrected, and the ion beams are allowed to travel in parallel to the Y axis when the desired orientation direction is parallel to the Y axis. That is, the present invention provides an apparatus and method for forming an alignment layer, the apparatus and the method being capable of aligning the ion beams in the desired orientation direction by adjusting the tilt angle of the reflective face with respect to regions irradiated with the ion beams. An apparatus, according to an embodiment of the present invention, for forming an alignment layer using a mask having a reflective face will now be described. With reference to FIG. 4, the alignment layer-forming apparatus 10 includes an ion source 12 for generating ion beams 28 and one or more masks 20 disposed between a substrate 24 and the ion source 12. The masks 20 each have a reflective face 34 directed to the substrate. Two of the masks 20 are arranged in parallel to the substrate 24 in the same level as shown in FIG. 1A and edges 32a and 32b of the masks 20 define a slit 22. The two masks 20 are used herein and one monolithic mask may be used. The slit is preferably used and may be omitted. The reflective face 34 is directed to the substrate, the substrate 24 is mounted on a movable table (not shown), and the ion beams are then applied or irradiated to a thin-film 26 disposed on the substrate while the substrate is being moved usually at a constant speed in the direction away from the ion source, that is, in the traveling direction of the substrate as indicated by an arrow shown in FIG. 4. The ion beams 28 are reflected between the reflective face 34 and the thin-film 26 to be processed into the alignment layer and the alignment layer is finally formed by the last irradiation of the thin-film 26 with the ion beams. This is due to that the bonds between atoms in the thin-film are broken with the ion beams reflected by the reflective face 34. The ion source 12 includes a plasma-generating chamber 16, a gas inlet 14 for introducing gas into the plasma-generating chamber 16, an accelerating electrode 18 for accelerating ions generated in the plasma-generating chamber 16, and a sheet-shaped grid (or grids) 11 having a plurality of ion outlets for emitting the accelerated ions. The grid 11 has a rectangular shape and the ion outlets are uniformly arranged in the grid 11. The gas used is, for example, argon (Ar). If argon gas is used, argon ions (Ar+) are generated in the plasma-generating chamber 16. The gas used is not limited to argon and inert gas such as neon or xenon may be used. Alternatively, the following gas or gas mixture may be used: gas, such as hydrogen, nitrogen, methane, or acetylene, containing an element contained in the alignment layer or a gas mixture of some of these gases. A method for forming a liquid-crystal alignment layer using a reflective face of a mask will now be described. As shown in FIG. 4, the thin-film 26 on the substrate is irradiated with the ion beams 28 having an energy of 50 to 1000 eV while the substrate 24 having the thin-film 26 placed thereon is being conveyed at a speed of 1 to 100 mm/s in the direction away from the ion source 12 (in the same direction as that in which the ion beams 28 are emitted), that is, in the traveling direction (the Y direction) of the substrate shown in FIG. 4. The longitudinal direction of a narrow region irradiated with the ion beams is referred to as an ion beam-irradiated region direction (the X direction), which is perpendicular to the traveling direction of the substrate (the direction perpendicular to the plane of FIG. 4) and corresponds to the direction in which the ion outlets are arranged. As shown in FIG. 4, the ion beam-irradiated region direction, the traveling direction of the substrate, and the direction perpendicular to the X-Y plane are hereinafter referred to as the X direction, the Y direction, and the Z direction, respectively. The ion beams 28 incident onto the thin-film placed on the substrate 24 are primarily reflected by the surface of the thin-film, are secondarily reflected by the reflective face 34, and then reach the thin film surface again, whereby the bonds between atoms in a surface domain of the thin-film are broken, resulting in the conversion of the thin-film into the alignment layer or alignment film. The energy of the ion beams may be adjusted such that the bonds between atoms are broken by the quaternary and/or senary reflection of the ion beams. Alternatively, as shown in FIG. 4, a mask tilt control section 40 may be used and the tilting operation of the mask may be controlled with a controller 42, whereby the position of the mask is adjusted such that the tilt angle of the reflective face with respect to the substrate, that is, the spatial tilt of the reflective face is varied. This allows the tilt angle of the reflective face 34 to be varied to control the direction of the reflected ion beams. Furthermore, the following system may be used: a computer system which optically detect the offset angle of the ion beams applied to the thin-film and which automatically control the mask tilt control section in response to detected information. As described above, the present invention is based on that since the ion beams reflected by the reflective face are applied to the thin-film on the substrate, the ion beams have energy sufficient to break the bonds between the atoms in the thin-film and final reflected components of the ion beams reflected by the thin-film determine the liquid crystal orientation of the alignment layer. A conventional mask 20 has a flat face directed to a substrate and is arranged such that the flat face is parallel to the substrate. In general, this mask is used to define a region irradiated with ion beams emitted from an ion source. FIG. 5 shows an example of the propagation direction of the ion beam 28 presumed to be reflected by the flat face of the mask 20. FIG. 5A is a plan view of the mask, FIG. 5B is a front view showing the mask and substrate observed in the ion beam-irradiated region direction, and FIG. 5C is a side view showing the mask and substrate observed in the traveling direction of the substrate. As shown in FIG. 5A, if the ion beam 28 is reflected by the flat face of the mask 20, the ion beam 28 is reflected in one (or straight) direction and the reflection direction of the ion beam 28 is not varied because the mask 20 is placed in parallel to the substrate 24. The reflected ion beam 28 is applied to a thin-film 26 disposed on the substrate while the ion beam 28 is spreading without varying reflecting direction. In other words, since the mask 20 has a reflective face 34 which is flat and parallel to the substrate and which is directed to the substrate, the ion beam that is reflected in one direction and then secondarily reflected reaches the thin-film (the reflective face is not shown in FIG. 5). That is, the irradiation direction of the ion beam agrees with the orientation direction, an alignment layer can be formed with the ion beam secondarily reflected without varying the direction of the ion beam. On the other hand, that mask 20 shown in FIG. 4 has that reflective face 34 directed to that substrate and that reflective face 34 is tilted with respect to that substrate as shown in FIG. 6. That is, the ion beams primarily reflected upward by that thin-film disposed on that substrate are reflected forward depending on the tilt angle γ offset from the X-axis direction such that the resulting ion beams are directed toward that thin-film. The ion beams are directed in a desired direction (for example, the direction parallel to the Y-axis direction when the desired direction is the Y-axis direction that is the traveling direction of that substrate) depending on the tilt angle β offset from the Y-axis direction. FIG. 6A is a plan view showing a mask and a substrate, FIG. 6B is a front view showing this mask and substrate observed in the ion beam-irradiated region direction, and FIG. 6C is a side view showing this mask and substrate observed in the traveling direction of this substrate. The arrangement shown in FIG. 6 is hereinafter referred to as “the tilt arrangement of a flat reflective face”. An ion beam is reflected forward in a desired orientation direction in such a manner that the offset angle φ between the desired orientation direction and the direction of the ion beam primarily applied to a thin-film disposed on a substrate is corrected. This is particularly effective in the case where the ion beam is offset from the desired direction in one direction. The tilt angle of the reflective face with respect to the traveling direction of the substrate (the Y-axis direction) is hereinafter referred to as an orientation correction tilt angle β and the tilt angle of the reflective face with respect to the substrate is hereinafter referred to as a face tilt angle γ. The face tilt angle γ is preferably adjusted to retain the constant incident angle α of the ion beam applied to the thin-film on the substrate and is not limited to this angle. FIG. 7 shows how the angle φ offset from an orientation direction is corrected by varying the orientation correction tilt angle β using the tilt arrangement of a flat reflective face. The correction is explained using an incident ion beam of which the offset angle φ from the orientation direction is −1 (relative value). As shown in FIG. 7, an increase in the orientation correction tilt angle β from zero degree (that is, the flat reflective face is not tilted) to 15 degrees by tilting the flat reflective face increases the angle φ offset from the orientation direction of a liquid crystal from −1 to +0.3 (relative value). The angle φ offset from the orientation direction of the liquid crystal can be detected by optically measuring a liquid crystal panel that is prepared in such a manner that a panel prepared by stacking two substrates facing each alignment-treated thin-film on each substrate and then sealing the stacked substrates with a sealant and the liquid crystal is injected between the resulting substrates or empty cell. Alternatively, the angle offset from the orientation direction can be detected by optically measuring the surface profile of a thin-film disposed on a substrate. FIG. 7 illustrates that the angle φ offset from the orientation direction can be reduced to zero, that is, the ion beam can be aligned with a desired orientation direction by adjusting the orientation correction tilt angle β to about 13 degrees in this example. In other words, in the tilt arrangement of the flat reflective face, the orientation direction of the liquid crystal can be controlled in such a manner that the angle φ offset from the orientation direction is varied by varying the orientation correction tilt angle β. For the shape of a reflective face 34 of a mask 20, a mask used for ion beam irradiation in which the spread of beams is wide may have a sawtooth-shaped surface having reflective faces which are tilted with respect to a substrate and which are arranged at a small pitch p in the X-axis direction in FIG. 4. In this case, the distances (gaps) between the substrate and the reflective faces of the mask can be maintained substantially constant as compared to “the tilt arrangement of a flat reflective face”; hence, differences between the directions of ion beams reflected by the reflective faces can be reduced. FIG. 8 includes views for illustrating a monolithic mask 60 having a plurality of reflective faces arranged in a sawtooth pattern. FIG. 8A is a plan view of the monolithic mask 60, FIG. 8B is a front view showing the monolithic mask 60 observed in the ion beam-irradiated region direction, and FIG. 8C is a sectional view showing the monolithic mask 60 and substrate observed in the traveling direction of the substrate. With reference to FIG. 8, the reflective faces 34 arranged in such a sawtooth pattern are spaced at a pitch p in the X-axis direction and have a face tilt angle γ and an orientation correction tilt angle β. The monolithic mask 60 is placed so as to have an orientation correction tilt angle β with respect to the Y-axis direction (parallel to the traveling direction of the substrate). Ion beams 28 incident onto a thin-film 26 disposed on the substrate 24 are primarily reflected by the thin-film surface and secondarily reflected by the reflective faces. The resulting ion beams 28 then reach the thin-film surface to break the bonds between atoms in a surface domain of the thin-film, whereby an alignment layer is formed. Since the reflective faces are tilted (principally due to the setting of the orientation correction tilt angle β), the propagation directions of the ion beams secondarily reflected are corrected with the reflective faces 34 with respect to the Y-axis direction, that is, in the X-Y plane (the surface of the thin-film on the substrate), whereby the ion beams are aligned in a desired orientation direction, that is, in the Y-axis direction in FIG. 8 and then applied to the thin-film 26 on the substrate 24 (see FIG. 8A). The face tilt angle γ of the reflective faces may be set such that the incident angle α of the ion beams with respect to the thin-film surface is maintained (see FIG. 8C). In general, as shown in FIG. 1B, beams 28 emitted from outlets located close to both ends of a grid of an ion beam apparatus are likely to spread; hence, it is effective to set the orientation correction tilt angle of end portions of a mask (both ends of the mask observed the ion beam-irradiated region direction) that correspond to the ends of the grid to a large value. FIG. 9 shows the relationship between the propagation direction of ion beams and a mask with a sawtooth-shaped surface having reflective faces arranged in end regions of the surface. FIG. 9A is a plan view of the mask, FIG. 9B is a front view showing the mask and substrate observed in the ion beam-irradiated region direction, and FIG. 9C is a side view showing the mask and substrate observed in the traveling direction of the substrate. As shown in FIG. 9, the reflective faces located in a center portion of the mask are set in parallel to the substrate because the directions of the ion beams incident onto the center portion of the mask 20 are not offset from a desired orientation direction. In this case, the ion beams 28b are primarily and secondarily reflected straight by the center portions thereof and then applied to a thin-film 26 disposed on the substrate 24 in such a manner that the angle of each ion beam 28b with respect to the traveling direction of the substrate (the Y-axis direction) is maintained. In contrast, the ion beams 28a and 28c reflected by end portions of the mask are incident onto the thin-film 26 on the substrate 24 in such a manner that the ion beams 28a and 28c are offset from the desired orientation direction at an angle φ. Therefore, in order to render the orientation direction uniform over the thin-film 26, it is effective that the ion beams secondarily reflected are aligned in the traveling direction of the substrate (the Y-axis direction) by correcting the offset angle φ in the ion beams secondarily reflection such that the propagation directions of the ion beams reflected by the end portions are aligned with those of the ion beams reflected by the center portions. The offset angle φ of the ion beams is corrected by the secondary reflection using a mask (for example, the mask shown in FIG. 8) having tilted reflective faces arranged in end portions of the mask 20 in a sawtooth pattern as shown in an enlarged view in FIG. 9. The reflective faces arranged in such a sawtooth pattern are spaced at a pitch of 2 mm and have a face tilt angle γ of 45 degrees and the offset angle φ can be reduced by adjusting the orientation correction tilt angle β as shown in FIG. 7. That is, the directions of the ion beams finally applied to the substrate can be aligned in such a manner that the propagation directions of the ion beams, reflected by the center portion, slightly spreading are not varied and the spread angle (about one degree) of the ion beams, reflected by the end portions, relatively greatly spreading is reduced by varying the orientation correction tilt angle β. The reflective faces arranged in the sawtooth pattern can be formed in such a manner that a reflective film is deposited on a processable material (for example, plastic or metal such as aluminum) by a sputtering process, a chemical vapor deposition (CVD) process, or a dipping process or in such a manner that a reflective film is mechanically bonded to such a processable material. Alternatively, the reflective faces can be formed in such a manner that a graphite material that is hardly sputtered by ion beam irradiation is directly processed so as to have a sawtooth shape. Reflective faces of a mask may be curved although the reflective face shown in FIG. 4 is flat. The reflective faces of the mask, shown in FIG. 8 or 9, having the sawtooth surface may be flat or curved. Reflective faces 34 arranged in a sawtooth pattern as shown in FIG. 10A are flat and have a tilt angle γ. On the other hand, reflective faces 34 arranged in a sawtooth pattern as shown in FIG. 10B are curved. In order to avoid that atoms emitted from a surface layer of a reflective face of a mask by sputtering the reflective face by ion beam irradiation affect the surface of a thin-film which is disposed on a substrate and which is processed into an alignment layer, it is effective that the reflective face of the mask and the thin-film are made of the same material. Examples of a material for forming the reflective face and the thin-film include polyimide (PI) and diamond-like carbon (DLC). The reflective face may have a thickness of 1 to 1000 nm and the thin-film may have a thickness of 1 to 100 nm. Such a material for forming the reflective face and the thin-film is not limited to these materials and the thicknesses of the reflective face and the thin-film are not limited to the above values. For example, the following material may be used: an organic material such as polyvinyl alcohol (PVA) or an inorganic material that can be orientated by ion beam treatment. If the reflective face is made of a material that has small sputtering yield or is hardly sputtered by an ion beam, the mask has high durability. It is effective that deposits on the reflective face are removed as required such that the reflectivity of the reflective face is maintained. In general, as shown in FIG. 1B, ion beams applied to both ends of a substrate are likely to have a large spread angle. FIG. 11 shows the positional relationship between ion beams 28, a thin-film 26 disposed on a substrate 24, and masks 20 similar to the mask, shown in FIG. 9, for aligning the propagation directions of those ion beams in the desired orientation direction (the Y-axis direction in FIG. 9), these ion beams 28, this thin-film 26, and the masks 20 being present on the X-Y plane. Arrows shown in this figure indicate the propagation directions of these ion beams in the X-Y plane. The rightmost arrows indicate the propagation directions of these ion beams incident onto this thin-film 26, the central arrows indicate those of these ion beams primarily reflected, and the leftmost arrows indicate those of these ion beams secondarily reflected. These ion beams secondarily reflected are incident onto this thin-film 26 again. As shown in FIG. 11, the central one of these ion beams has an extremely small spread angle. Hence, reflective faces located in a center portion of each mask may be flat and parallel to this substrate such that the central ion beam propagates straight. Reflective faces located in end portions of the mask may be tilted in opposite directions such that the other ion beams offset from a desired orientation direction (the Y-axis direction) in opposite directions are aligned in the desired orientation direction. If, for example, a mask having a configuration shown in the enlarged view in FIG. 9 is used, the propagation directions of the other ion beams offset from the desired orientation direction in opposite directions can be aligned in the desired orientation direction in such a manner that the correction tilt angles β of reflective faces, arranged in a sawtooth pattern, located in end portions of the mask are set to β or negative β. When the absolute values of the angles of the ion beams, located outside, offset from the desired orientation direction are different from each other, it is effective to use a mask having reflective faces of which the correction tilt angles β are set depending on the offset angles. Furthermore, it is effective that the position of the mask shown in FIG. 11 is precisely adjusted using the mask tilt control section shown in FIG. 4 such that the tilt angle thereof is properly set. FIG. 12 shows an example using masks 20 having reflective faces 34 of which the orientation correction tilt angle β increases stepwise (varies gradually) toward end portions of each mask that correspond to both ends of a substrate. The orientation correction tilt angles β of the reflective faces reflecting ion beams having large offset angles are set to large values depending on the spread angles of the ion beams and those of the reflective faces reflecting ion beams having small offset angles are set to small values, whereby the propagation directions of the ion beams applied over a thin-film 26 disposed on the substrate 24 can be aligned in a desired orientation direction. Furthermore, it is effective that the position of the mask shown in FIG. 12 is precisely adjusted using the mask tilt control section shown in FIG. 4 such that the tilt angle thereof is properly set. In the above description, the spread angle of the beams, that is, the angle of the beams offset from the desired orientation direction increases toward outside. If a mask having reflective faces of which the orientation correction tilt angle is set to correspond to the offset angle of the ion beams is used, an effective correction can be obtained. As described above, a uniform orientation direction can be achieved by improving the configuration of reflective faces of a mask. The following apparatuses and methods are as described above: apparatuses and methods for determining the shapes and arrangements of reflective faces of masks that reflect ion beams such that the ion beams are aligned in the desired orientation direction of a liquid crystal. In another embodiment of the present invention, the pre-tilt angle of a liquid crystal can be controlled using ion beams reflected by reflective faces of a mask. If the shape and arrangement of the reflective faces are as shown in FIG. 13, the pre-tilt angle thereof can be desirably controlled by varying the incident angle of the ion beams that are reflected by the reflective faces and then applied to a thin-film disposed on a substrate. An ordinary liquid crystal display is known to be operable in a twisted nematic (TN) mode and the longitudinal axes of liquid crystal molecules sandwiched between an upper substrate and a lower substrate are twisted by 90 degrees with respect to the upper and lower substrates. It is used for switching that the liquid crystal molecules are untwisted and the longitudinal axes of liquid crystal molecules are then aligned perpendicularly with respect to the substrates by the application of a voltage. In order to enhance the response of the liquid crystal molecules to the applied voltage and in order to eliminate the misalignment or alignment defect of the liquid crystal molecules, the longitudinal axes of the liquid crystal molecules need to be aligned such that the longitudinal axes thereof are tilted with respect to the substrates (that is, alignment layers) at a few degrees. The angle of the longitudinal axes of the liquid crystal molecules (the initial arrangement of the liquid crystal molecules to which no voltage is applied) with respect to the alignment layers is usually referred to as a pre-tilt angle. For alignment treatment by ion beam irradiation, it is known that the pre-tilt angle of a liquid crystal usually depends on the incident angle α of an ion beam with respect to a thin-film processed into an alignment layer. In a liquid crystal display having an initial arrangement in a vertical alignment (VA mode) as a wide viewing angle technology, a liquid crystal usually has a pre-tilt angle of about 90 degrees and it is used for switching that molecules of the liquid crystal are laid down and then aligned horizontally by the application of a voltage. FIG. 13 shows a mask having reflective faces 34 arranged in a sawtooth pattern. The reflective faces 34 are spaced at a pitch p in parallel to the ion beam-irradiated region direction (the Y-axis direction) and have an appropriate face tilt angle γ. This allows the incident angle α of ion beams with respect to a thin-film to be set to a predetermined value depending on the pre-tilt angle of a liquid crystal. As shown in FIG. 13, the ion beams reflected by the thin-film 26 disposed on a substrate are secondarily reflected by the reflective faces repeatedly arranged; hence, the incident angle of the secondarily reflected ion beams with respect to the thin-film 26 is close to the right angle and corresponds to a pre-tilt angle. Thus, a pre-tilt angle close to 90 degrees can be achieved depending on the incident angle of the secondarily reflected ion beams. This provides a vertical alignment having good viewing angle properties. This is due to that the pre-tilt angle of the liquid crystal does not depend on the incident angle α of the ion beams primarily applied to the thin-film but the incident angle αp of the secondarily reflected ion beams (the ion beams which are finally applied to the thin-film and which can break the bonds between atoms in a surface domain of the thin-film). In this case, the pre-tilt angle of the liquid crystal can be varied without varying the orientation direction of the liquid crystal. Even if the incident angle α of the ion beams primarily applied to the thin-film is small due to constraints of an alignment layer-forming apparatus, the incident angle of the ion beams secondarily reflected can be set to substantially the right angle. The mask shown in FIG. 13 is an application for the liquid crystal, having a large pre-tilt angle, operable in a vertical alignment mode. The mask can be applied to the twisted nematic (TN) mode, an in-plane switching (IPS) mode in which liquid crystal molecules are rotated in a horizontal plane with an applied voltage, and the control of the incident angle of an ion beam used in a small pre-tilt angle (one to ten degrees) range by varying the reflective faces. The following mask may be used: a mask, having reflective faces repeatedly arranged in a sawtooth pattern, for controlling the pre-tilt angle. The face tilt angle γ of the reflective faces are adjusted such that ion beams are reflected toward a substrate to be incident onto the substrate. The reflective faces extend along a region (the direction perpendicular to the plane of FIG. 3, or the Y direction in FIG. 3) irradiated with the ion beams. In this case, it is the key to set the tilt angle γ of the reflective faces. The orientation correction tilt angle β may be zero, that is, the ion beams may be reflected straight in the X-Y plane. It is effective that the mask has end portions which have an orientation correction tilt angle β and which are tilted flat faces. The reflective faces may be flat or curved. The embodiments of the present invention are as described above. The present invention is not limited to the embodiments. Within the scope of the present invention, various improvements, modifications, and variations may be made on the basis of findings of those skilled in the art. |
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047643056 | description | In all the examples, use is made of a polymerizable mixture incorporating: an epoxy resin constituted by a bis-phenol A diglycidyl ether with an epoxy equivalent of approximately 190 diluted by neopentyl diglycidyl ether and marketed by CDF Chimie under the reference MN 201 T, a hardener constituted by the products sold under reference D6M5 by CDF Chimie and which consists of a cycloaliphatic polyamine having an amine equivalent of approximately 63 and an adduct of diaminodiphenyl methane and epoxy resin having an amine equivalent of approximately 130 and a pitch solution marketed under reference 730/30 by Desailly. EXAMPLE 1 This example illustrates the coating of solid waste constituted by metal filings and tools placed in a metal basket. 20% by volume of said waste is introduced into a 200.iota. container. 80% by volume of a mixture (40% pitch, 20% hardener and 40% resin) are then added and line the gaps left between the waste. The block obtained has a Shore D hardness of 54. EXAMPLE 2 This example illustrates the conditioning of a drainage oil constituted by an industrial lubricating oil. In this example, into a 200 liter barrel is introduced a mixture of pitch and hardener D6M5, epoxy resin MN 201 T and drainage oil in the following proportions: 10% by weight drainage oil, 30% by weight pitch, 22.5% by weight hardener, 37.5% by weight epoxy resin. The mixture is then allowed to harden at 20.degree. C. for 15 days and the Shore hardness of the block is determined and is approximately 67D. Tests performed by replacing part of the drainage oil by pitch and retaining the same resin and hardener proportions have revealed that the Shore hardness decreases with the oil content. These results are illustrated in the attached graph representing the Shore hardness as a function of the oil percentage of the oil--pitch mixture. Moreover, it has been found that on increasing the oil content of the oil--pitch mixture to above 50%, there is a decanting or settling of the oil, which does not make it possible to obtain a homogeneous block. EXAMPLE 3 This example illustrates the conditioning of waste constituted by tributyl phosphate. In this example, mixing takes place of the tributyl phosphate (TBP), pitch, hardener and epoxy resin in the following proportions: 4% by weight TBP, 36% by weight pitch, 22.5% by weight hardener, 37.5% by weight epoxy resin. After the mixing operation, hardening is allowed to take place at 20.degree. C. and the Shore hardness of the blocks obtained is measured after 15 days hardening and is approximately 52D. EXAMPLE 4 This example illustrates the conditioning of a scintillation liquid used for beta counting and constituted by 99.5% by weight solvent, mainly xylene and 0.5% by weight linked with scintillator. In this example, the scintillation liquid, pitch, hardener and epoxy resin are mixed in the following proportions: 4.8% by weight scintillation liquid, 35.2% by weight pitch, 22.5% by weight hardener, 37.5% by weight epoxy resin. As hereinbefore, the Shore hardness of the blocks obtained is measured after 15 days hardening at 20.degree. C. and exceeds 50D. |
047298707 | claims | 1. A nuclear fuel element comprising a tubular can having an internal diameter between a maximum diameter and a minimum diameter, which is sealed at at least one end by a plug having a cylindrical portion force fitted into said can, wherein the cylindrical portion has a diameter equal to the minimum diameter of the can and has at least three serrations oriented in accordance with the generatrixes of said cylindrical portion and which are regularly spaced, said serrations forming overhanging beads with respect to said cylindrical portion over a thickness equal to half the difference between the maximum diameter and the minimum diameter of the can. 2. A nuclear fuel element according to claim 1, wherein the cylindrical portion of the plug has three serrations. |
abstract | Apparatus and methods for compensating for the movement of a substrate in a lithographic apparatus during a pulse of radiation include providing a pivotable mirror configured to move a patterned projection beam incident on the substrate in synchronism with the substrate. |
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047387998 | summary | BACKGROUND OF THE INVENTION The present invention relates to permanent disposal of radioactive waste from nuclear reactors, particularly radioactive wastes from primary coolant fluid systems and steam generator blowdown sludge. In the operation of nuclear power plants, radioactive particular waste develops in the primary coolant fluid and in the sludge produced during steam generator blowdown. In the latter case, primary-to-secondary leakage in the steam generator requires that the blowdown sludge be treated as radioactive waste. All radioactive waste from a nuclear reactor must be processed for disposal in a manner which minimizes the exposure of operating personnel to radiation. The conventional filter cartridges utilized in the auxiliary systems of a nuclear power plant have been found to be difficult to handle and to create a significant radiation exposure hazard for the maintenance personnel. To help alleviate this hazard, replacement of conventional removable cartridge filters with backflushable filters has been proposed. Backflushable filters, however, while convenient, serve only as temporary collection and holding devices. Final removal of the radioactive particulate waste, or "crud", requires backflushing of these filters to convey the waste to a suitable disposal plant for packaging and burial. SUMMARY OF THE INVENTION It is an object of the present invention to prepare such waste for permanent disposal in a manner which minimizes the radiation exposure hazard to operating personnel. Another object of the invention is to transfer such radioactive waste to permanent storage containers by a fully remotely controllable system which enables the operating personnel to be safely isolated from regions where radiation is present. Another object of the invention is to permanently store radioactive particulate waste in relatively inexpensive and structurally simple containers. A further object of the invention is to permanently store such radioactive waste in pre-packaged. ferromagnetic filter matrices disposed within such containers. A further object of the invention is to permanently store such radioactive waste in the form of a uniform, dewatered dispersion of solid radioactive waste, encapsulated in an organic resin solidification system in a liquid-impervious disposal package. Another object of the invention is to permanently store such radioactive waste in a disposable electromagnetic filter cartridge in conjunction with a backflushable filtration system forming part of a nuclear plant fluid system. The above and other objects are achieved, according to the invention, by the provision of a noval cartridge for permanent disposal of radioactive particulate waste, comprising: a liquid impervious casing having an upper end cover, a lower end cover and a side wall extending between the covers, the casing enclosing a waste storage region; ferromagnetic fibrous material defining a waste retaining matrix and filling a major portion of the waste storage region; means defining an inlet conduit extending through the upper end cover and axially of the casing through the waste storage region, and opening into the waste storage region in the vicinity of the lower end cover; and means defining first and second outlet conduits extending through the upper end cover and opening into the waste storage region in the vicinity of the upper end cover. Preferably, the casing and the conduits are all made of a suitable plastic, such as fiberglass, with all components being securely bonded together, for example by means of a suitable adhesive, to form a rigid unit. In accordance with a particular novel feature of the invention, the ferromagnetic fibrous material is constituted by ordinary steel wool, preferably of a fine grade. The grade employed will be determined, at least to a substantial extent, by the size of the particles to be stored, as will be explained in detail below. Preferably, the cartridge further includes two annular, perforated retainer-distribution plates spaced apart in the axial direction of the casing and delimiting the portion of the waste storage region containing the matrix. Between each retainer-distribution plate and an associated cover for the casing, there is thus defined a space for the circulation of fluids between the portion of the waste storage region containing the matrix and the various conduits. Preferably, the upper end cover of the cartridge casing is formed to present a reservoir for holding any liquid which may spill during the introduction of liquid containing the radioactive waste of flushing water to the cartridge. Any spillage can be removed by means of a siphon tube introduced into that reservoir during the filling operations. In addition, the upper end cover is formed, at its top, to present a lateral flange, preferably an inwardly extending annular flange, which partly covers the reservoir and which is used to lock the cartridge to an associated filling system. In the use of this cartridge, a slurry containing the radioactive waste to be stored will be delivered via the inlet conduit, while the liquid filtrate contained in that slurry as well as subsequently delivered flushing water, are expelled via the first output conduit. After the flushing water has been circulated through the waste storage region, dewatering air is introduced via the inlet conduit and is expelled via the first outlet conduit. During this operation, the second outlet conduit will be blocked and the cartridge will be disposed within a magnetic field which acts to trap the particulate waste in the matrix. During the subsequent encapsulation operation, which also takes place while the cartridge is in the magnetic field, encapsulating material is injected via the inlet conduit, while the air previously trapped in the cartridge is expelled via the second outlet conduit, the first outlet conduit then being blocked. Introduction of the encapsulating material continues until the interior of the cartridge is completely filled. In order to monitor the filling of the cartridge with encapsulating material, the second outlet conduit is provided with a check valve which is oriented to be closed by the pressure exerted by the encapsulation material when it enters the second outlet conduit. Closing of the valve actuates a microswitch that is also disposed in the second outlet conduit in order to produce a signal indicating completion of filling with the encapsulating material. The encapsulating material can be of any suitable composition already known in the art, such as known resin-catalyst systems, or even cement. The ends of the conduits which project from the cartridge all project from the upper end thereof and preferably have conically tapered surfaces to define coupling components. While the conically tapered surfaces are preferably exterior surfaces which define male coupling elements, they may also be constituted by interior conduit surfaces to define female coupling components. Preferably, the bottom cover is provided, at its lower surface, with an alignment groove which will cooperate with a lug provided on an associated conveyor to assure that the cartridge is correctly alinged with conduits of a cartridge filling system, the latter conduits being formed, at their lower ends, to present coupling elements constructed to mate with those of the cartridge conduits. The objects of the invention are further achieved by the provision of a novel system for storage and encapsulation of radioactive particulate waste, comprising: a cartridge having a liquid impervious casing enclosing a waste storage region, a ferromagnetic waste storage matrix housed in the cartridge and occupying at least a major portion of the waste storage region, and an inlet conduit and at least one outlet conduit projecting from the cartridge and communicating with the waste storage region; means for establishing a magnetic field in the matrix; fluid handling means including a source of liquid containing the radioactive waste to be stored in the cartridge, a source of flushing water, a source of air, a source of encapsulating material, and a receptacle of receiving flushing water; cartridge filling means including a plurality of conduits releasably couplable to the conduits associated with the cartridge; and fluid flow control means including a plurality of remotely controllable valves connected between the fluid handling means and the cartridge filling means, the fluid flow control means having a first operating state for selectively supplying liquid containing the radioactive waste, flushing water, or air from their respective sources to the inlet conduit for loading the matrix with radioactive waste, while placing one outlet conduit in communication with the receptacle, and the fluid flow control means having a second operating state, for supplying encapsulating material from its source to the inlet conduit for filling the cartridge with encapsulating material, while permitting air in the cartridge to be expelled via one outlet conduit. Preferably, of course, the cartridge of the above system is of the type described earlier herein. The means for establishing a magnetic field is preferably constituted by an annular solenoid presenting an axial passage dimensioned to permit introduction of the cartridge. Systems employing a solenoid to apply an electromagnetic field to a ferromagnetic storage medium are already known in the art. According to one preferred embodiment of the system according to the invention, the cartridge has first and second outlet conduits, the cartridge filling means is movable between a first operating position associated with the first operating state of the fluid flow control means and second operating position associated with the second operating state of the fluid flow control means, the conduits of the cartridge filling means are grouped into a first set of conduits releasably couplable to the conduits associated with the cartridge when the cartridge filling means are in the first operating position, and a second set of conduits releasably couplable to the conduits associated with the cartridge when the cartridge filling means are in the second operating position, the one outlet conduit which is placed in communication with the receptacle is the first outlet conduit, and the one outlet conduit via which air is permitted to be expelled when the fluid flow control means is in the second operating state is the second outlet conduit. The cartridge filling means includes one conduit which is common to both sets, and the movement of the cartridge filling means between its first and second operating positions is effected by pivoting the filling means about the axis of the common conduit. Preferably, this common conduit is coupled to the second outlet conduit of the cartridge and constitutes the conduit via which air is vented from the waste storage region of the cartridge during filling with encapsulating material. The cartridge filling means is preferably constituted by a turret carrying the various conduits. The turret is supported by a column which is, in turn, supported in a loading head constituted by a cylindrical housing having a closed upper end and an open lower end. The housing is constructed to permit its lower end to form a sealed connection with the top of the cartridge. In addition, the column is movable vertically relative to the loading head to displace the turret between a raised, or retracted, position when the conduits associated with the cartridge filling means are separated from those of the cartridge, and a lowered, or coupled, position in which one set of conduits of the cartridge filling means will be coupled in a sealed manner to the cartridge conduits. The turret further carries a siphon tube which is positioned to be introduced into a reservoir formed at the top of the cartridge, around the projecting ends of the associated conduits, to permit aspiration of any liquid which may spill onto the top of the cartridge during the course of the filling operation. The loading head further carries a group of locking cams arranged to cooperate with a flange formed at the upper end of the cartridge in order to lock the cartridge against the lower end of the loading head during the various filling operations. The valves of the fluid flow control means can all be of a conventional type. These valves are preferably electrically controllable to permit remote-controlled operation of the system. The objects of the invention are further achieved by operating the system defined above as follows: placing the cartridge filling means in the first operating position and coupling the first set of conduits in a sealed manner to the conduits associated with the cartridge; operating the establishing means for establishing a magnetic field in the matrix; while the first set of conduits is coupled to the conduits associated with the cartridge, effecting loading of waste material by operating the fluid flow control means for blocking the second outlet conduit and sequentially supplying liquid containing the radioactive waste via the inlet conduit to the waste storage region until the matrix is loaded with waste while conducting liquid from the waste storage region via the first outlet conduit to the receptacle, supplying flushing water via the inlet conduit to the waste storage region and from the waste storage region via the first outlet conduit to the receptacle, and supplying air via the inlet conduit to the waste storage region and from the waste storage region via the first outlet conduit until substantially all liquid has been removed from the waste storage region; after the step of effecting loading of waste material, decoupling the first set of conduits from the conduits associated with the cartridge, placing the cartridge filling means in the second operating position, and coupling the second set of conduits in a sealed manner to the conduits associated with the cartridge; and while the second set of conduits is coupled to the conduits associated with the cartridge, effecting encapsulation of the waste material by operating the fluid flow control means for blocking the first outlet conduit, and supplying encapsulating material via the inlet conduit until the waste storage region is filled with encapsulating material, while removing air from the waste storage region via the second outlet conduit. The method and apparatus according to the present invention serves to produce a dewatered dispersion of reactor coolant system corrosion products or steam generator blowdown sludge in a water-impervious organic resin-steel wool matrix, encapsulated in a plastic cylinder which can be automatically loaded, immediately after the encapsulation operation, into a standard shipping container whose top is subsequently sealed in a subsequent conventional operation. Operation of this system according to the present invention provides a convenient, economical and efficient technique for preparing backflush slurry or sludge for disposal at minimal risk of significant radiation exposure to personnel. The invention is particularly applicable to the storage of particulate matter which has been transported through a reactor coolant system and its auxiliaries. Such particulate matter consists largely of corrosion products having small particle sizes ranging from a few microns to colloidal size, as well as a certain amount of ion exchanger resin fines and, occasionally, bits of debris from various sources. All of the material is radioactive since it has been subjected to the neutron flux in the reactor. The coolant system auxiliaries include cartridge filters installed to remove particulate matter having sizes in the range of several microns. Some of the smaller particles either agglomerate or adhere to larger particles and are removed therewith. The auxiliaries may also include a CVCS letdown demineralizer which filters larger particles and absorbs some colloidal material which then is removed when the resins are replaced. The internal surface of the reactor coolant system, and especially the core, are the overwhelming competitors of the waste particles since the rate of filtration through the auxiliaries is a small fraction of the reactor coolant flow (less than 0.05%). The waste particles depositing on these surfaces account for 85% of the occupational radiation exposure experienced by plant personnel. When the steam generators of such reactor systems exhibit primary-to-secondary leakage, the blowdown sludge from these generators must be handled as radioactive waste. This sludge can also be removed by collecting it in backflushable filters forming part of the reactor system, and then transporting it to cartridges for permanent storage in accordance with the present invention. The system according to the invention for premanently storing such particulate waste material in cartridges takes advantage of the magnetic properties of the reactor coolant particulates and the steam generator sludge to effect a quantitative dewatering of the waste material, together with efficient packaging for burial. In order to utilize the magnetic properties of the waste particles, introduction of the waste material into the cartridge, as well as subsequent introduction of encapsulating material, are carried out while the cartridge is within a magnetic field. According to techniques known in the art, this can be achieved by introducing the cartridge into an axial passage enclosed by an annular solenoid generating the requisite magnetic field level. To avoid damaging the cartridge or in case of failure to achieve proper engagement between the cartridge and the cartridge filling means, appropriate interlocks can be provided to prevent attempted engagement of the cartridge filling means with the cartridge when the two are not correctly positioned relative to one another. The system according to the present invention would be well suited for handling waste products derived from high temperature large flow rate filtration of reactor coolant. The system according to the invention could also be used for disposal of radioactive material produced during decontamination of steam generators. For example, it is known to decontaminate the primary side channel heads of such steam generators by means of a slurry of boron oxide in water which is applied in a manner to "grit blast" the channel head surfaces. Other grits which have been considered are magnetite and aluminum oxide. The purpose of such decontamination is to remove the highly radioactive primary corrosion deposit, i.e. nickel ferrite, so that the radiation source strength will be lower and result in a lower radiation does to operating personnel. A process of this type employing boron oxide could produce of the order of 20,000 gallons of 4-5% boric acid solution with a high concentration of nickel ferrite particles. The system according to the present invention could be employed to dispose of the waste (or crud), enabling the boric acid solution itself to be reused as a carrier for boron oxide for other steam generators, or to be disposed of. If magnetite were used as the abrasive, it, along with the nickel ferrite "crud" , could be collected and permanently stored according to the invention leaving essentially clean water for reuse or disposal. The system according to the present invention could, if desired, be mounted on a trailer and carried to the place where it would be used, such use being carried out either in connection with existing tankage, or temporarily constructed tankage. The system according to the invention could also be constructed with a plurality of operating stations disposed along a conveyor line, so that a plurality of cartridges could be simultaneously loaded with waste material . |
abstract | A rod assembly for a fuel bundle of a nuclear reactor may include an upper end piece, lower end piece and a plurality of rod segments attached between the upper and lower end pieces and to each other so as to form an axial length of the rod assembly. The rod assembly may include an adaptor subassembly provided at given connection points for connecting adjacent rod segments or a given rod segment with one of the upper and lower end pieces. The connection points along the axial length of the rod assembly may be located where the rod assembly contacts a spacer in the fuel bundle. One (or more) of the rod segments may include an irradiation target therein for producing a desired isotope when a fuel bundle containing one (or more) rod assemblies is irradiated in a core of the reactor. |
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abstract | Disclosed are embodiments of an ion beam sample preparation apparatus and methods for using the embodiments. The apparatus comprises an ion beam irradiating means in a vacuum chamber that may direct ions toward a sample, a shield blocking a portion of the ions directed toward the sample, and a shield retention stage with shield retention means that replaceably and removably holds the shield in a position. The shield has datum features which abut complementary datum features on the shield retention stage when the shield is held in the shield retention stage. The shield has features which enable the durable adhering of the sample to the shield for processing the sample with the ion beam. The complementary datum features on both shield and shield retention stage enable accurate and repeatable positioning of the sample in the apparatus for sample processing and reprocessing. A retention stage lifting means allows the creation of a loading chamber that is isolated from the main vacuum chamber where sample ion beam milling takes place. A heat sink means is configured to conduct heat away from the sample undergoing sample preparation in the ion beam. |
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claims | 1. An actuating apparatus that measures rotation of a jet pump beam bolt in a nuclear reactor, comprising:a socket member for engaging the jet pump beam bolt;a sleeve member configured to engage a handling pole, the sleeve member being rigidly connected to the socket member;an actuator hex rigidly connected between the socket member and the sleeve member in an axial direction, the actuator hex not rigidly connected to the socket member in a rotational direction and not rigidly connected to the sleeve member in the rotational direction, the actuator hex configured to engage a stationary tensioner so that as the socket member and the sleeve member rotate in the rotational direction, the actuator hex remains stationary;a displacement wheel rigidly attached to the actuator hex; anda displacement dial rigidly attached to the sleeve member or socket member, the displacement dial corresponding to the displacement wheel so as to permit rotational displacement measurement. 2. The actuating apparatus of claim 1, wherein the displacement wheel includes a degree wheel on the actuator hex for indicating the degree of rotation of the beam bolt. 3. The actuating apparatus of claim 1, wherein the displacement dial includes a degree dial corresponding to the displacement wheel. 4. The actuating apparatus of claim 2, wherein the degree wheel demarcates angular rotation in increments of 5 degrees. 5. The tension apparatus of claim 1, wherein the socket member is a drive-deep impact type. 6. The tension apparatus of claim 1, wherein the sleeve member includes a drain hole and a plurality of pin holes. 7. A system for tensioning jet pump beam bolts in a nuclear reactor, the nuclear reactor comprising at least one jet pump with each jet pump comprising a jet pump beam and a jet pump beam bolt, the system comprising:an actuating apparatus includinga socket member for engaging the jet pump beam bolt,a sleeve member configured to engage a handling pole, the sleeve member being rigidly connected to the socket member,an actuator hex rigidly connected between the socket member and the sleeve member in an axial direction, the actuator hex not rigidly connected to the socket member in a rotational direction and not rigidly connected to the sleeve member in the rotational direction, the actuator hex configured to engage a stationary tensioner so that as the socket member and the sleeve member rotate in the rotational direction, the actuator hex remains stationary,a displacement wheel rigidly attached to the actuator hex, anda displacement dial rigidly attached to the sleeve member or socket member, the displacement dial corresponding to the displacement wheel so as to permit displacement measurement; andthe stationary tensioner includinga base block having an opening for accommodating the jet pump beam bolt,a top plate having an opening for accommodating the actuator hex, anda hydraulic actuator for providing tension to the jet pump beam. 8. The system of claim 7, wherein the displacement wheel includes a degree wheel on the actuator hex for indicating the degree of rotation of the beam bolt. 9. The system of claim 7, wherein the displacement dial includes a degree dial corresponding to the displacement wheel. 10. The system of claim 8, wherein the degree wheel demarcates angular rotation in increments of 5 degrees. 11. The system of claim 7, wherein the opening in the top plate is a hex-like shape. 12. The system of claim 7, wherein the top plate further comprises a lifting member, the lifting member includes a handle for lowering the tensioning apparatus. 13. The system of claim 7, wherein the tensioning apparatus further comprises tension hooks for hooking onto the jet pump beam. 14. The system of claim 13, wherein the tension hooks are controlled by a pneumatic cylinder. 15. The actuating apparatus of claim 1, wherein the displacement dial is rigidly attached to the sleeve member or socket member in the rotational direction, and wherein the displacement dial corresponds to the displacement wheel so as to permit rotational displacement measurement of the jet pump beam bolt. 16. The system of claim 7, wherein the displacement dial is rigidly attached to the sleeve member or socket member in the rotational direction, and wherein the displacement dial corresponds to the displacement wheel so as to permit rotational displacement measurement of the jet pump beam bolt. |
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abstract | The absorber rod comprises cladding (12) of stainless steel closed by plugs (14, 16) and containing a column of absorber pellets (24) e.g. of boron carbide. It also has an end bar (26) of hafnium secured to the bottom plug by a purely mechanical connection. |
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