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description | The present invention relates to a submerged or underwater electricity production module. More particularly, the invention relates to an underwater electricity production module, which includes means in the form of an elongated cylindrical box in which means are integrated forming an electricity production unit including means forming a nuclear boiler, associated with electricity production means connected to an external electricity distribution station by electrical cables. Such modules are known in the state of the art. Reference may for example be made to documents U.S. Pat. No. 5,247,553, JP 50 018 891 and U.S. Pat. No. 4,302,291. These various documents describe underwater or submerged electricity production units in which electricity production means can be integrated associated with means forming a nuclear boiler, for example. It is known that such structures have a certain number of advantages, since nuclear-based energy is an effective and profitable answer to energy and ecological problems. Such structures also make it possible to resolve a certain number of problems, in particular related to safety and accounting for risks, whether natural, such as tsunamis, hurricanes or others, or human, for example such as plane crashes or malicious acts. It is also known that these various projects have not resulted in industrial exploitations for the moment, due to the fact that their technical feasibility and economic relevance have not been demonstrated. Work to improve this type of structure has been conducted by the Applicant for a number of years. This work has already resulted, for example, in the filing of a number of patent applications to which reference can be made, and in particular documents FR 2951008, FR 2951009, FR 2951010, FR 2951011, FR 2951012, FR 2958782, FR 2958783 and FR 2958784. Several of these documents in particular deal with the operating safety of these types of modules, and in particular their safety in case of major incident, as has recently occurred for land-based power plants. The aim of the invention is to propose still other improvements to this type of submerged module to further improve the operating safety thereof. To that end, the invention relates to an underwater electricity production module of the type including means in the form of an elongated cylindrical box in which means are integrated forming an electricity production unit including means forming a nuclear boiler, associated with electricity production means connected to an external electricity distribution station by electrical cables, characterized in that the nuclear boiler-forming means are placed in a dry chamber of the reactor compartment associated with the chamber forming a safety water storage reservoir of the reactor whereof at least the radial wall is in a heat exchange relationship with the marine environment, in that the dry chamber of the reactor compartment is associated with a compartment receiving electricity production means, and in that the latter includes means for introducing quenching water of the dry chamber receiving the reactor, placed in its lower portion and including a seawater inlet formed in the radial wall of the module at that compartment receiving the electricity production means, a conduit between that seawater inlet and the dry chamber of the reactor compartment, and means forming a quenching valve for that chamber. According to other aspects of the invention, the underwater module comprises one or more of the following features: the nuclear boiler means include a primary circuit comprising at least one reactor container, a pressurizer, a steam generator and a primary pump and a primary backup circuit in parallel on that primary circuit and including at least one primary passive heat exchanger placed in the safety water storage reservoir chamber of the reactor; the primary passive heat exchanger placed in the safety water storage reservoir chamber of the reactor is placed at a higher level than that of the reactor container; each branch of the primary backup circuit includes valve-forming means; the primary backup circuit is connected to the primary circuit upstream or downstream from the primary pump; the primary backup circuit is connected to the primary circuit upstream from the primary pump and in that it includes means for short-circuiting that primary pump; the nuclear boiler means include a secondary circuit associated with the electricity production means and a secondary backup circuit in parallel on that secondary circuit and including at least one secondary passive heat exchanger placed outside the underwater module in the marine environment; the secondary passive heat exchanger placed outside the underwater module in the marine environment is placed at a higher level than that of the steam generator; each branch of the secondary backup circuit includes valve-forming means; the secondary circuit includes isolating valve-forming means and in that the secondary backup circuit is connected between said isolating valve means; the secondary circuit extends partially in a compartment receiving the electricity production means and in that the secondary backup circuit passes through the radial wall of the compartment and is connected to the secondary passive heat exchanger placed outside said compartment; the dry chamber of the reactor compartment is connected to the safety water storage reservoir chamber of the reactor by depressurizing means including means forming a depressurizing valve placed in the upper portion of the dry chamber and connected to means forming a bubbler placed in the lower portion of the reservoir-forming chamber and in that excess flow check means are provided between the upper portion of said reservoir-forming chamber and the dry chamber; the nuclear boiler means include a reactor container, placed in a reactor pit whereof the lower portion is connected to the lower portion of the safety water storage reservoir chamber of the reactor through means forming a water intake conduit placed along the radial wall of the module and whereof the upper portion is connected to a corresponding portion of the storage reservoir chamber through means forming a water return conduit; valve means are placed in the means forming intake and return ducts; an enclosure made from a thermally insulating material is placed around the portion of the reactor container housed in the reactor pit, at a distance from the wall of that container, so as to define an interstice forming a thermal barrier between said enclosure and said container; during normal operation, the interstice between the enclosure and the container is filled with a gaseous material and in that the enclosure includes, in the lower portion thereof, at least one water inlet opening; during normal operation, the water placed in the reactor pit is borated water; the end of the water inlet duct connected to the water storage reservoir chamber is associated with a filtering screen; the nuclear boiler means include a pressurizer connected by the depressurizing means to the safety water storage reservoir chamber of the reactor; the depressurizing means include a depressurizing circuit provided with a depressurizing valve connected to means forming a bubbler placed in the lower portion of the safety water storage reservoir chamber of the reactor; means for deviating the jet of water are placed across from the means for introducing seawater into the dry chamber of the reactor compartment; vent-forming means are placed in the upper portion of the dry chamber of the reactor compartment between the latter and the compartment receiving the electricity production means; the inlet of the vent means is associated with filtering means; and it includes valve means for connecting the safety water storage reservoir chamber of the reactor to the reactor container. As previously indicated, the invention relates to a submerged or underwater electricity production module. Such modules are for example illustrated in said FIG. 1 and are for example designated by general references 1, 2 and 3 in that figure. These modules are for example associated and submerged off a coast designated by general reference 4, and they are for example placed on the bottom or kept at some distance from the bottom of the sea, on an electricity production site designated by general reference 5. These different modules are connected by electrical cables, designated by general reference 6, to an external electricity distribution station, also for example serving as a remote control/command center for the modules, that center for example being land-based and designated by general reference 7 in FIG. 1. This external electricity distribution station is then traditionally connected using electricity distribution lines designated by general reference 8, for example to an electricity distribution grid for example powering a town located nearby and designated by general reference 9 or any other electricity consumer in general. It will also be noted that land-based infrastructures, for example such a support designated by general reference 10, can be considered to house support means, for example such as support vessels, one of which is designated by general reference 11 in that figure, making it possible to intervene on the production site. The support means for example make it possible to place the modules, ensure that they are kept in operational condition, or recover them for major operations to be performed on land, such as replacing the nuclear fuel. In fact, and as illustrated in FIG. 2, each underwater electricity production module designated by general reference 1 in this FIG. 2 includes means in the form of an elongated cylindrical box, the ends of which are for example rounded. These means are designated by general reference 12 in this figure, and are placed on the bottom or kept at some distance from the bottom, for example 13, of the sea and to that end include leg assemblies designated by general reference 14 and anchor means designated by general reference 15, making it possible to position, place, and maintain that module on the bottom. Different embodiments of the leg assemblies and anchor means can be considered. FIG. 2 also describes one possible embodiment of the inside of such a module, which in fact includes a certain number of compartments placed next to each other, and separated by partitions. Thus, for example, such a module 12 may include, at each end thereof, means in the form of a ballast designated by general references 16 and 17, for example making it possible to control the submersion of the module. Furthermore and going from left to right in FIG. 2, this module may include a reactor compartment designated by general reference 18 in that figure, the reactor compartment in turn being divided into two associated chambers, i.e., a dry reactor compartment chamber strictly speaking, designated by general reference 19 and in which means forming a nuclear boiler are housed, and the chamber forming a safety water storage reservoir of that reactor, designated by general reference 20. These chambers of the reactor compartment 18 are for example placed next to each other and are separated by a so-called tight partition. Next to this reactor compartment, a compartment is provided for receiving electricity production means, that compartment being designated by general reference 21 and for example comprising a turbo-alternator group or assembly or other auxiliary systems, as will be described in more detail hereafter. After this compartment 21 for receiving electricity production means, the module 12 may include a compartment forming an electrical plant designated by general reference 22 for example for voltage conversion, etc., traditionally, and a compartment 23 including a control station for all of the elements of the module, for example. Of course, other embodiments of the inside of the module and other configurations and arrangements of the elements thereof may be considered. Thus, for example, a living compartment intended to house a crew member, for example for exploitation or intervention purposes, may also be considered. FIG. 3 shows, in more detail, the part of the module 12 where the reactor compartment 18 and the compartment 21 intended to receive the electricity production means are provided. As also previously indicated, the reactor compartment 18 is therefore intended to receive means forming a nuclear boiler and includes two chambers, i.e., the dry chamber for receiving the reactor strictly speaking, designated by general reference 19, and the safety water storage reservoir chamber thereof, designated by general reference 20. In fact, and traditionally, the nuclear boiler means, which are designated by general reference 30 in this FIG. 3, then include a primary circuit designated by general reference 31 comprising at least one reactor container 32, a pressurizer 33, a steam generator 34 and a primary pump 35. These nuclear boiler means 30, and more particularly the steam generator 34 thereof, also include the secondary circuit, which passes through the separating partition of the reactor and electricity production means receiving compartments 18 and 21, and associated with said electricity production means. The secondary circuit is designated by general reference 36 in this FIG. 3 and the electricity production means are designated by general reference 37 and are therefore positioned in the compartment 21. In fact, these electricity production means 37 for example include a turbine designated by general reference 38 in this figure, associated with an alternator designated by general reference 39, a condenser designated by general reference 40 and a secondary pump designated by general reference 41 in FIG. 3. This architecture of the electricity production means is simplified here for comprehension purposes. As everyone knows, it is in reality more complex to increase the output of the thermodynamic cycle. Also traditionally, the boiler means 30 are connected to different means making it possible to inject water therein at different pressures, for example in the case of a primary water loss accident. These means are for example designated by general reference 50 in FIG. 3 and comprises injection means, for example high-, medium- or low-pressure, for injecting water into the reactor depending on the nature of the accident and the selected backup strategy. Thus for example, the safety water storage reservoir chamber 20 can be connected to the reactor container 32 by means of a conduit designated by general reference 51 in this figure, associated with the valve means designated by general reference 52. Other traditional injection systems for injecting water into the reactor are also provided. Vent means 50a are then provided between the dry chamber 19 and the safety reservoir chamber 20. If needed, these vent means, which are normally closed, open to allow air to enter the reservoir chamber and therefore to allow the low-pressure injection of water from the reservoir into the container 32 by means of the direct injection line 51. If the pressure is too great in the primary circuit to perform this injection, the primary circuit can be depressurized rapidly using depressurizing valve means designated by general reference 31a, in addition to other depressurizing means that will be described in more detail hereafter. These valve and vent means are then controlled and commanded by control-command members, which can be automated or driven by human operators. In the underwater module according to the invention, the safety water storage reservoir 20 of the reactor is used for other safety functions thereof and at least its radial wall designated by general reference 53 is in a heat exchange relationship with the marine environment in which that module is submerged. This makes it possible to form a quasi-unlimited cold source that is naturally continuously available, independently of the circumstances and operating problems that may arise, to cool the module and in particular the nuclear boiler means. The problems having recently arisen in nuclear power plants have in fact become serious following the loss of this cold source. It is in fact known that one of the major problems related to the operation of nuclear reactors relates to the fact that such a reactor continues to generate extremely significant quantities of heat, even after the chain reaction is stopped and for a relatively long period of time. As an example, a small reactor of 160 electric MW (500 thermal MW) still creates a power of 3 thermal MW three days after it is stopped. This characteristic requires that these reactors be associated with specific cooling systems to discharge that residual power and that the continuous availability thereof be ensured. Without such a system, the core of the reactor has a very high likelihood of melting and causing radioactive materials to be dispersed into the environment. Recent events have shown the potential consequences of the simultaneous loss of a cold source, for example such as seawater intakes, and electricity making it possible to provide energy to those cooling systems. In fact, the great majority, if not all currently known nuclear reactors use backup systems using pumps to discharge the residual power from the core toward a cold source, for example through exchangers. These systems are of course made redundant, diversified, and are subject to careful inspection and maintenance to maximally reliabilize the cooling function of the core in case of stop or accident. In the same spirit, land-based nuclear power plants have various redundant electricity sources to power those backup systems, for example such as power supply means using redundant electric grids, generator groups, or backup batteries, etc. However, experience has shown that all of these systems may fail at one time or another, which ultimately amounts to the loss of the cold source and therefore a cooling failure of the reactor, with the consequences that have been seen on several reactors. This is not the case and cannot occur in the energy production module according to the invention. In fact, that module may include various so-called “passive” safety systems, i.e., not requiring electricity to operate, except, depending on the selected embodiment, for example for the power required for their control-command. The reactor may first include a primary passive cooling circuit in parallel on the primary circuit of the reactor. This primary passive backup circuit is designated by general reference 54 in FIG. 3 and includes at least one primary passive heat exchanger designated by general reference 55, placed in the safety water storage reservoir chamber of the reactor, that chamber being designated by general reference 20 in FIG. 3. In fact, this heat exchanger 55 may for example be placed in the safety water storage reservoir chamber 20 of the reactor, at a higher level than that of the container 32 of the reactor, and one or each branch of that primary backup circuit 54 may include valve means. Such valve means are for example designated by general reference 56 in FIG. 3 and the primary passive backup circuit 54 may be connected to the primary circuit upstream or downstream of the primary pump previously described and designated by general reference 35. In the case where the primary passive backup circuit is connected to the primary circuit upstream from the primary pump 35, it also that includes means for short-circuiting the primary pump. On the other side, the backup circuit is connected between the container and the steam generator. One can then see that this primary passive backup circuit makes it possible to discharge the residual power of a submerged nuclear reactor for a very long period of time by using a natural cooling loop. In fact, opening this primary passive backup circuit makes it possible to form a bypass loop of the primary circuit, so as to discharge the heat produced in the core of the reactor toward the cold water reserve through an exchanger, that cold water reserve being formed by the safety water storage reservoir 20 of the reactor. Furthermore, the radial wall 53 of that chamber forming a water reservoir 20 is, as previously indicated, in a heat exchange relationship with the marine environment, and therefore makes it possible to form a long-term, or even quasi-unlimited cold source, by heat dissipation in the marine environment. Thus, the discharge of the residual power from the reactor is done through the bypass loop of the primary circuit of the reactor, that loop comprising: a cold water reservoir situated in the dedicated chamber 20 of the reactor compartment, formed by the safety water storage reservoir of the reactor, two pipe elements tapped at the outlet of the container of the reactor and the inlet of the primary pump, for example, a heat exchanger 55 submerged in the safety water storage reservoir constituting a primary passive exchanger, the shell 53 of the compartment ensuring the heat exchanges between the safety water storage reservoir and the sea, and associated control/command valves. During normal operation of the module, a valve can then close that passive cooling loop and no fluid circulates therein. The water reserve in the reservoir-forming chamber 20 is at a low temperature, i.e., for example at the temperature of the seawater, and at a low pressure, while the primary fluid, i.e., that circulating in the primary circuit of the reactor, is at a high pressure and high temperature. The thermal power of the reactor is discharged toward the steam generator(s) of the primary circuit using the primary pump(s). When the reactor is stopped in a normal or accidental situation, the reactor is halted and the passive backup cooling thereof is implemented. The valve(s) of the passive cooling loop for example open automatically or on command, and, for example, the inertia of the primary pump starts a fluid movement in that loop, i.e., in the primary passive backup circuit. The hot water leaving the core of the reactor then rises in the cooling loop up to the exchanger 55, where it transmits its heat to the cold water reserve contained in the chamber 20 in a heat exchange relationship with the sea. The water, becoming heavier, then re-descends in the loop to rejoin the cold branch of the circuit and the core of the reactor, where it is heated again. The water in that backup circuit is liquid throughout the entire cycle. The cycle maintains itself indefinitely as long as the temperature difference between the core and the safety water storage reserve is large, i.e., for several days, or even several weeks. In fact, the submersion of the module, and in particular of the reactor compartment thereof, in the sea gives the safety water storage reservoir a significant cooling capacity through the shell in a heat exchange relationship with the marine environment, to dissipate the power transmitted by the passive exchanger. It is thus possible to see that such a safety system, applicable to a submerged reactor, has a major asset relative to land-based reactor systems, in particular in terms of passive safety operation, inasmuch as the primary passive backup circuit operates based on a natural circulation between the hot source (the container of the reactor) and the quasi-unlimited cold source (the primary passive exchanger placed in the safety water storage reservoir of the reactor, in a heat exchange relationship with the sea). Such a backup system is then not dependent on any power supply for a pump, the availability of water intake, for example seawater, etc., to ensure cooling of the reactor. Likewise, a secondary passive cooling circuit can also be provided in parallel on the secondary circuit of the reactor. This secondary passive backup circuit is for example designated by general reference 60 in FIG. 3. The latter is then connected in parallel on the secondary circuit 36 of the reactor, for example in the compartment 21 designed to receive the turbo-alternator group 37, and then also includes at least one secondary passive heat exchanger designated by general reference 61, placed outside the underwater module in the marine environment and connected to the rest thereof by pipe elements passing through the radial wall of the compartment 21. This secondary passive heat exchanger 61 is then also placed a higher level than that of the steam generator 34 so as to form a backup cooling circuit with natural circulation. This also makes it possible to discharge heat from the secondary circuit of the reactor, using a quasi-inexhaustible cold source, i.e., also the marine environment. In a module as considered, the heat generated by the nuclear reaction in the core of that reactor is, in a normal exploitation situation, transmitted to a coolant of a primary circuit and discharged in primary heat exchangers, generally called steam generators, like that designated by general reference 34 in FIG. 3. In these exchangers, a second fluid circulates and begins to boil. The steam thus produced supplies a turbine driving an alternator to generate electricity. This is called the secondary circuit, like that designated by general reference 36 in FIG. 3, associated with the electricity generating means designated by reference 37. These steam generators therefore act as a cold source for the primary circuit of the reactor and the heat extraction is driven by secondary pumps of that circuit. In an accident situation, for example for a traditional land-based reactor, nuclear fission stops, but the core continues to generate significant heat due to the radioactivity. The heat generators can still perform their role as cold source and discharge power from the core on the condition that the secondary pumps, and in general the secondary circuit, continue to function correctly. That is why it is crucial that in that type of reactor, the secondary circuit continues to receive electricity, and in particular the secondary pumps, such as the pump designated by reference 41 in FIG. 3. However, and as previously indicated, the loss of electricity is an eventuality that cannot be completely ruled out. The pumps may then not operate and the cooling of the reactor is then no longer performed. The pumps may also break down. Here again, in the module according to the invention, the marine environment can be used to form a cold source and resolve these problems. Thus, in the module according to the invention, the secondary passive heat exchanger 61 is used and placed outside the module to form, with the marine environment, a quasi-inexhaustible natural cold source for the secondary circuit. Also in this case, the secondary passive heat exchanger 61 is placed outside the underwater module 12 in the marine environment at a higher level than that of the steam generator 34 with which it is associated, so as to allow natural circulation between those elements. One or each branch of the secondary passive backup circuit designated by general reference 60 in FIG. 3 may also include valve means designated by general reference 62 in that figure. In fact, the secondary circuit 36 strictly speaking includes isolating valves such as the valves designated by general references 63 and 64 in FIG. 3, the secondary passive backup circuit then being connected to the secondary circuit between those isolating valves. As also shown in FIG. 3, the secondary circuit in fact passes through the transverse wall 65 separating the reactor compartment 18, and in particular its dry chamber 19, from the compartment 21 receiving the turbo-alternator assembly. In that case, the secondary passive backup circuit includes pipe elements passing through the radial wall of the module at that compartment receiving the turbo-alternator assembly 21, to connect the secondary passive heat exchanger 61, which makes it possible to avoid any crossing of the shell at the reactor compartment. One can see that in the module according to the invention, the secondary circuit is also equipped with a passive backup cooling loop bypassed on said main secondary circuit. In the event electricity is lost and power is therefore cut to the pumps of the secondaries of the steam generators, the backup system may be used to extract the heat from the primary circuit of those steam generators and therefore the reactor through natural circulation through a diphasic passive secondary heat exchanger toward the sea, which then represents a quasi-inexhaustible cold source. The passive secondary heat exchanger is then placed outside the shell of the module and at a higher level than that of the steam generator to allow that natural circulation, which makes it possible to avoid using pumps that could fail. Such a system is in fact also completely passive and does not require any electricity. Such a system then includes: a heat exchanger 61 transmitting the heat from the secondary circuit of the boiler means toward the sea, placed outside the compartment designed to receive the turbo-alternator assembly 21, two pipe elements for the bypass of the secondary circuit, and the tapping of which may then be done upstream after the isolating valves 63 of the circuit and downstream after the supply pumps 41 or the steam generator(s), a valve 62, which is normally closed, situated upstream from the passive exchanger on the bypass line, a valve 64, which is normally open, situated upstream from the turbo-alternator group on the secondary circuit and downstream from the tapping of the bypass line, tight shell crossings, and a control-command system for those valves. During normal operation of the reactor, no fluid crosses through the secondary passive backup exchanger. The turbo-alternator group is supplied with steam by the secondary circuit of the steam generator and generates electricity. The secondary of the steam generator is supplied with water by the secondary pump 41. In an accident situation, typically in case of loss of electricity resulting in a failure of the secondary pump(s), the valve 62 that is normally closed opens and the valve 64 that is normally open closes. This action occurs in several seconds, for example automatically or upon demand by an operator. The turbo-alternator group 37 is then no longer supplied with steam and electricity production stops. It is then the passive backup secondary exchanger 61 that is supplied with steam. That steam, for example in contact with the tubes of that exchanger cooled by the cold seawater, condenses, discharging its heat toward the environment. The liquid water then returns by gravity into the steam generator 34 without requiring a secondary pump. This water heats in the steam generator and is again vaporized before leaving toward the backup circuit. The cycle maintains itself naturally until the heat transmitted by the primary circuit is no longer sufficient to create steam in the steam generator, i.e., for example after several days of backup operation as previously described. As previously indicated, the shell crossing for the secondary passive exchanger is situated at the turbo-alternator compartment 21, so as to reinforce the sealing of a third barrier and the confinement of the radioactive materials in the event the first two barriers, i.e., the sheaths and the primary circuit, are no longer tight. One can thus see that such a system has many advantages relative to land-based system, since it is simple and very effective. Other safety means are provided in a submerged module according to the invention. Thus, for example, one possible scenario of a major accident for a pressurized water nuclear reactor is the rupture of a pipe of the primary circuit in the dry chamber 19 of the reactor compartment 18. This pipe rupture then releases high-temperature water, which, undergoing an abrupt pressure drop, vaporizes instantly in the dry chamber of the reactor compartment. The confinement enclosure surrounding the reactor is then quickly invaded by high-temperature steam. The value of the pressure and temperature peak occurring during this accident dimensions the resistance of the enclosure and the equipment contained therein. For a land-based reactor, the pressure peak reaches several bars and dimensions the thickness of the concrete and metal enclosure that needs to be provided. For a submerged reactor like that considered in the module according to the invention, this peak reaches higher values due to the smaller volume of the reactor compartment, i.e., in particular the dry chamber 19, relative to a land-based power plant. Any pressure reduction system during an accident may be interesting then to limit the impact in terms of the stresses applied and that the equipment installed in that chamber in particular must undergo. In the module according to the invention, the dry chamber 19 of the reactor compartment 18 is connected to the safety water storage reservoir chamber 20 of the reactor, by depressurizing means designated by general reference 70 in FIG. 3. In fact, these means include means 71 forming a depressurizing valve placed in the upper portion of the dry chamber 19 and which are connected to means forming a bubbler designated by general reference 72, placed in the lower portion of the reservoir chamber 20. Excess flow check means designated by general reference 73 are provided between the upper portion of that reservoir chamber 20 and the dry chamber 19. Thus, in the event of a pipe rupture, for example of the primary circuit, the steam from the dry chamber 19 of the reactor compartment 18 is conducted, by a pipe and valve assembly, toward the safety water reservoir 20, which then serves as an overpressure elimination reservoir, where that steam is injected and condenses in contact with the cold water. In the case of the accident previously considered, the pressure inside the dry chamber 19 is thus immediately reduced and the danger of a break of that enclosure is eliminated. In general, the radial wall of the reactor compartment 18 is in a heat exchange relationship and is continuously cooled by the seawater, which makes it possible to ensure a discharge of heat toward the marine environment and therefore to cool the water contained in that chamber 20. In particular, the contact between the cold marine environment and the radial wall of the dry chamber 19 receiving the reactor also makes it possible to ensure condensation of the steam and generally cooling thereof, for example in case of a break of a primary pipe as previously indicated. The cooling of the wall of this chamber in fact causes the condensation of at least part of the steam contained in that chamber 19 in such an accident case, and also a natural and prolonged manner. It is in fact not necessary to project water on the outside of this compartment, as is the case in certain land-based power plants, because the module according to the invention is submerged and the wall of the reactor compartment is therefore continuously in contact with the cold water. The pressure is therefore reduced over the short term by the depressurizing means designated by general reference 70 and over the long term by the radial wall of the dry chamber 19 receiving the reactor, in a completely passive manner. Furthermore, the pressurizer designated by general reference 33 in FIG. 3 may also be equipped with depressurizing means connected to the reservoir chamber 20. Thus for example, in FIG. 3, the pressurizer 33 is connected by depressurizing means designated by general reference 80 to the reservoir chamber 20. In fact, these depressurizing means include a depressurizing circuit provided with a depressurizing valve designated by general reference 81 for example, and connected to means forming a bubbler designated by general reference 82 and also placed in the lower portion of the safety water reservoir chamber 20 of the reactor. This also makes it possible to discharge, in that reservoir chamber 20, any overpressure of the pressurizer and the primary circuit in general. Other safety systems may also be considered, like those illustrated in FIG. 3, and the operation of which is more clearly shown in FIGS. 4 and 5. FIGS. 4 and 5 in fact show partial views of a module according to the invention. This module still includes the means in the form of an elongated cylindrical box 12, the reactor compartment 18 with the dry chamber 19, and the safety water storage reservoir chamber 20 of the reactor. The nuclear boiler means 30 with the reactor container 32 are also shown. In fact, and as more clearly illustrated in these FIGS. 4 and 5, this reactor container 32 is placed in a reactor pit designated by general reference 90, for example placed at the bottom of the dry chamber 19. The lower portion of this reactor pit 90 is connected to the lower portion of the safety water storage reservoir chamber 20 of the reactor, through means forming a water intake conduit designated by general reference 91, placed along the radial wall of the module, that wall still being designated by general reference 53. The upper portion of the reactor pit 90 is connected by means of a water return conduit, designated by general reference 92, to the corresponding portion of the storage reservoir chamber 20. As illustrated, valve means are for example placed in these means forming a water intake and return of said reactor pit 90 to the reservoir chamber 20. These valve means are respectively designated by references 93 and 94 for the water intake and return conduits. Of course, other embodiments can be considered. It will also be noted, as illustrated, that the end of the water intake conduit 91, connected to the storage water reservoir chamber 20, is associated with a filtering screen, which is designated by general reference 95. As also illustrated in FIGS. 4 and 5, that enclosure made from a thermally insulating material can be traditionally placed around the reactor container portion 32 housed in that reactor pit 90. Thus, for example, in FIGS. 4 and 5, this enclosure is designated by general reference 96 and for example assumes the form of a bowl or cup, and is placed away from the wall of the container, so as to define an interstice forming a thermal barrier between said enclosure 96 and said reactor container 32. In fact, during normal operation, this interstice between the insulating material enclosure 96 and the reactor container 32 can be filled with a gaseous material, for example such as air or another material, as illustrated in FIG. 4, so as to form an additional thermal barrier making it possible to insulate the container so as to avoid heat losses. The enclosure 96 also includes, in the lower portion thereof, at least one water inlet opening designated by general reference 97 connected with the water intake conduit 91 and then allowing the water to penetrate the interstice around the reactor container. It will also be noted that during normal operation, the water placed in the reactor pit 90 around the bottom thereof may be borated water. The water contained in the reservoir chamber 20 may also for example be borated water. One can see that in case of a serious accident and, for example, melting of the core of the reactor, the corium formed is deposited at the bottom of the container, as illustrated in FIG. 5. This fusion wash is then capable of piercing the shell if the latter is not cooled. It should be noted that receiving ash from the corium may be provided under the container if that were to occur. To avoid this phenomenon, in the system according to the invention, the valves 93 and 94 are opened to cause the natural circulation of water in the reactor pit 90 around said reactor container 32, between the container and the reservoir chamber 20. In fact, when the valves 93 and 94 are opened, the interstice between the container 32 and the insulating material enclosure 96, which is normally filled with air, is then invaded by cold water coming from the reservoir chamber 20. In contact with the container 32 at a high temperature and in particular the bottom thereof, because this container is heated by the melted corium, the water around the container is brought to boiling and rises in the interstitial space between the insulating material enclosure 96 and the container. This space being connected to the water reservoir chamber 20, the steam and the hot water rise and escape from the reactor pit as illustrated in FIG. 5, to penetrate the rest of the reservoir chamber 20 where the steam condenses and the water cools. At the same time, the cold water, which is more dense, from the bottom of the reservoir chamber 20 rushes from the lower portion of that reservoir chamber 20 into the reactor pit 90 through the water intake conduit 91, which extends along the radial wall 53 of the module so as to be in heat exchange contact with the marine environment and thus slightly more cooled. There is thus a permanent natural water circulation regime established between the reservoir chamber and the reactor pit so as to cool it and for example prevent the corium that is formed from piercing that container. The water that circulates in that circuit is therefore doubly cooled on the one hand during its passage in the reservoir chamber 20, since the radial wall thereof is in a heat exchange relationship with the marine environment, and on the other hand during its passage in the water intake conduit into the reactor pit, since the latter is also formed along that radial wall. In fact, the radial wall of the reservoir chamber 20 and the dry chamber 19 of the reactor compartment 18 being in heat exchange contact with the marine environment, the cooling water of the reactor container is cooled continuously and naturally by a quasi-inexhaustible source. This is also an improvement making it possible to control the temperature, in particular of the reactor container and the corium in case of accident to prevent any new deterioration of the situation. Lastly and as a last resort, it is also possible to provide quenching of the dry chamber 19 receiving the reactor in the module according to the invention, using seawater. A decision may in fact be made, for one reason or another, to completely quench the dry chamber 19, and therefore the reactor, using seawater, which has particularly interesting properties for this type of situation. To that end, as illustrated in FIG. 3, it is provided that the compartment receiving the electricity production means, designated by general reference 21, includes means for introducing quenching water into the dry chamber 19 receiving the reactor. These quenching means are designated by general reference 100 in FIG. 3 and are for example placed in the lower portion of the compartment 21 receiving the electricity production means. These quenching means then include at least one seawater inlet designated by general reference 101 in FIG. 3, formed in the radial wall of the module for example at the bottom of that compartment 21 receiving the electricity production means, a water conduit between the seawater inlets 101 and the dry chamber 19 of the reactor compartment 18 passing through the partition separating the reactor compartment and the compartment receiving the electricity production means and means forming a quenching valve for that dry chamber 19, designated by general reference 102. It will also be noted that means for deviating the water jet leaving said quenching means, designated by general reference 103 in FIG. 3, are for example placed across from said means for quenching the dry chamber of the reactor compartment, to deviate the jet for example toward the bottom of that dry chamber and prevent any additional deterioration of the elements contained in that chamber. Vent means 104 are also provided in the upper portion of the dry chamber 19 of the reactor compartment 18, between the latter and the compartment 21 receiving the electricity production means, the inlet of those vent means 104 being associated with means designated by general reference 105, for example for filtering particles such as radioactive particles. One can then see that all of these arrangements make it possible to improve the safety and security of the operation of this type of structure. In particular, the submersion of this module and its proximity to the marine environment make it possible to take advantage of the fact that that environment may constitute a quasi-inexhaustible and continuously available cold source and which may be used to resolve, by natural circulation or pressure difference, a certain number of problems related to any accident. In addition, being submerged at a depth makes this module in sensitive to surface phenomena such as, for example, tsunamis or hurricanes. It also protects it from malicious acts. |
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047864613 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Reference will now be made in detail to the present preferred embodiments of the invention, examples of which are illustrated in the accompanying drawings. Turning first to FIG. 1, there is depicted a nuclear reactor pressure-vessel assembly 100. The assembly comprises a reactor pressure vessel 102 which, in the illustrated embodiment, is a generally cylindrically shaped vessel closed at one end by a generally hemispherically shaped end portion 104. The vessel 102 has a coolant inlet 106 and a coolant outlet 108 formed therein to circulate coolant through the vessel. As viewed in FIG. 1, the top end of the vessel 102 is closed by a reactor pressure-vessel head 110 which sits on a flange portion 112 of the vessel 102. A lower internals assembly 114 is supported within the vessel 102 and hung from the flange 112. The lower internals assembly comprises a perforated lower core support plate 116 and a perforated upper core support plate 118. The lower internals are used to support an array of fuel rod assemblies 120, two of which are illustrated in FIG. 1 by way of example. Positioned axially above the lower internals assembly 114 in the reactor pressure vessel 102 is the upper internals assembly 122. The upper internals assembly is generally disposed within the core barrel 124 of the lower internals assembly 114 and also hung from the flange 112. The upper internals assembly 122 comprises a hat-shaped upper internals top plate 126 and a perforated upper internals barrel 128 disposed within the core barrel 124 and forming an annulus 127 therewith. As illustrated in FIG. 1, the upper internals assembly 122 contains a plurality of control rod guide structure assemblies 130, only two of which are illustrated for clarity. The guide structure assemblies 130 pass through the upper internals top plate 126 forming an annulus therewith through which coolant in the upper head region 132 can pass into the upper internals. Coolant flows into the reactor pressure vessel through the coolant inlet 106. Most of the coolant flows through the inlet annulus 134 and towards the bottom region 136 of the vessel. The coolant then flows up through the perforated lower core support plate 116 and is heated to working temperatures while passing through the lower internals. The coolant then flows through the upper internals, the perforated upper internals barrel and out through the coolant outlet 108. As will be appreciated by the artisan, the rate of coolant flow in a large reactor is tremendous and the forces exerted by the coolant against the upper and lower internals structure as it flows therethrough tends to displace those structures causing vibrations and the like. A small portion of the coolant flows to the upper head region 132 through the annulus 127 and a coolant flow passage 158 (FIG. 4), described in detail below, which is provided in the Belleville spring assembly 138 of the present invention. The coolant flowing into the head region 132 serves the important safety function of maintaining the control rod drive mechanism components which pass through the upper head region at coolant inlet temperatures. Turning now to FIG. 2 there is illustrated a sectional view of the pressure vessel of FIG. 1 omitting the reactor internal details for clarity. As can be appreciated from FIG. 2, a plurality of stud holes 137 are axially spaced about the pressure vessel flange 112 for accommodating stud bolts or the like used to fasten the pressure vessel head 110 to the pressure vessel 102. A series of alignment keys slots 142 are formed in the core barrel 124. In the embodiment illustrated the alignment keys slots are rectangular in shape and each is operable to accommodate a key member 140 (FIG. 1) in order to at least grossly align the core barrel 124 and the pressure vessel 102. As will be appreciated from FIG. 2, a plurality of Belleville spring assemblies 138 are angularly spaced about the flange 144 of the core barrel 124. Each of the assemblies 138, which are described in detail below, are relatively small and have a large deflection capability. Consequently, they require less precision in design and manufacture than a single large Belleville spring or a few large spring assemblies. Moreover, the small Belleville spring assemblies 138 are easily replaced when necessary and can be easily handled, inspected and decontaminated. The small hold down spring assemblies 138 represent a significant cost savings, both in terms of reactor construction and maintenance, over prior art spring assemblies. In the embodiment of FIG. 2, a plurality of shim members 146 are provided to limit the amount of travel of the upper internals band with regard to the core barrel. As clearly seen in FIG. 2, the Belleville washer assemblies 138 are positioned on a flange 144 of the core barrel 124 and the shims 146 are dimensioned to prevent the spring assemblies from over deflection or collapse in case of a seismic event or the like by limiting the maximum allowable deflection of the Belleville spring assemblies 138. Turning now to FIG. 3, there is depicted in greater detail a section of the reactor vessel assembly of FIG. 1 in the vicinity of a Bellevile spring assembly 138 indicating reactor coolant flow directions. All upper internals components are omitted for clarity. The core barrel 124 has a flange 144 that sits on a ledge 150 formed along the inside circumference of the reactor vessel 102. A coolant flow passage 152 is formed through the core barrel flange 153 at the site of the Belleville spring 138. The upper internals barrel 128 includes flange 154 which is dimensioned to sit axially above the core barrel flange 144. The flange 154 has a plurality of coolant passages 156 angularly spaced to align with the passages 152 when the core barrel and upper internals barrel are assembled in the pressure vessel. As best seen in FIG. 3, each Belleville spring assembly 138 is disposed between the upper internals barrel flange 154 and core barrel flange 144 and has a connecting coolant flow passage 158 aligned with the passages 152 and 156, thus providing a continuous passage for inlet coolant in the annulus 134 to flow into the upper head region 132 via an upper head annulus 131 (FIG. 1) and then through a small annulus formed in the drive rod hole 160 when the drive rod mechanism (not illustrated) is in place. The upper internals barrel 128 is resiliently supported above and in coaxial alignment with the core barrel 114 by the annularly spaced plurality of Belleville spring assemblies 138 and alignment keys 140. Turning now to FIG. 4, there is illustrated a first preferred embodiment of a Belleville spring assembly according to the present invention. Each spring assembly 138 comprises means for carrying a stack of Belleville springs. Such means may comprise a central retainer 162, preferably having an upper flange 164 which seats against the upper internals barrel flange 154. Grooves 166 may be cut into the sealing surface 168 of the flange 164 to improve the seal with the flange 154. The lower portion 170 of the retainer 162 is formed with a tapped length of reduced diameter for accommodating locking nut 172 which is used during the assembly of the hold down spring assembly 138 to preload the springs. A locking pin 174 or other similar fastening device is used to arrest the locking nut 172 on the central retainer 162. Belleville springs 176 are stacked vertically on the retainer 162. As illustrated in FIG. 4, the Belleville springs may be stacked as single layers of opposing spring disks of alternating angular orientation or may be arranged in alternating and opposing groups of two or more spring disks of the same angular orientation. In either manner, a spring assembly is built up which, after assembly, bears against generally flat inner flange surface 178 of the retainer 162 and against the locking nut 172 with a set preload. If the retainer 162 is fabricated without a flange 164, an external snap ring or the like may be used to retain the spring disks. In the embodiment illustrated, the assembly comprises a stack of nine spring disks which, when compressed, exert a force of on the order of 20,000 pounds. If, for example, fifty spring assemblies are mounted on the core barrel flange 144, they will cumulatively exert a force of 1,000,000 pounds against the upper internals which is sufficient to fix the internals in place during normal and accident conditions. The stack of Belleville springs sits in a recess 179 formed in the flange 144. Positioned within a bore 180 of the retainer 162 is a means defining the connecting coolant passage 158 which comprises a bellows flange 182, a spring bellows 184 and a plunger 186. The bellows flange 182 seats against a shoulder 188 formed in the bore 180. A snap ring 190 or similar fastener holds the bellows flange in place in the bore 180. The bellows flange and plunger are provided with shoulders 192 and 194 respectively between which the spring bellows 184 is adapted to be retained. The compressed deflection of the spring assembly is typically in the range of about 1/2 to 1 inch depending upon the specific design chosen. This deflection, which is several times that of a large single Belleville spring, provides protection against unloading due to stress relaxation. With the present invention, a small deformation due to relaxation produces a small change in spring force whereas with the large Belleville spring the same deformation produces a large reduction in loading. The difficulties in replacing a large Belleville spring were previously described. The small spring assemblies of the present invention weight on the order of 50 pounds and are therefore easily handled. Decontamination and inspection of the small spring assembly is far easier and results in less man-rem exposure than is the case with large hold down springs. Preferably, the core barrel cooling passage 152 is formed with a cone shaped seating surface 196 and the plunger 186 is formed with a spherical end which seats against the surface 196 in a generally fluid tight manner under a bias from the spring bellows 194. The plunger 186 has a bore 198 which aligns with a tube 200 extending from the bellows flange 182. The tube 200 is dimensioned to fit within an expanded diameter section 202 of the plunger bore 198 and to be sufficiently spaced from the plunger bore to allow for spring assembly deflection. It will be appreciated that the coolant flow path from the passage 152 to the passage 156 is through the seat 196, the plunger 186 and the bellows flange tube 200. The spring bellows biases the plunger against the seating surface 196 and prevents any coolant flow past the tube 200 from dispersing into the upper internals. Thus, basic design consists of employing multiple stacks of small Belleville springs arranged in a circle on the flange of the core barrel flange 144. The small Belleville springs have the configuration of a conical thick wall washer and are typically 7 to 10 inches in outside diameter with a center hole of 31/2 to 4 inches diameter. Typically 5 to 10 small spring washers are stacked vertically and held together by the cylindrical retainer 162. The bore 180 in the center of the retainer provides space for the connecting flow passage generally 158 with its sealing and expansion accommodating features. Turning now to FIG. 5 there is illustrated a second embodiment of a Belleville spring assembly in accordance with the present invention. The only significant difference between the embodiments of FIGS. 4 and 5 is that an upper plunger 204 is used in lieu of the bellows flange assembly 182 of FIG. 4. The passage 156 is formed with a conical seating surface 206 similar to the surface 196 discussed above against which the generally spherical end 208 of the plunger 204 is biased by the spring bellows 184. The upper plunger 204 is formed with a retaining plate 210 which, after assembly, is restrained to move axially between the snap ring 190 and the shoulder 188 of the retainer 162. Thus, a flow passage is provided with improved sealing and expansion accommodation features. The connecting flow passage, generally 158, does not require leak tightness but it is desirable to minimize leakage between the coolant flow passage and the upper internals interior volume since such leakage will not contribute to head cooling and will reduce the reactor outlet temperature by dilution, and will add to the plant pumping lower load. The flow passage design of the spring assembly of either FIGS. 4 or 5 will achieve this goal. In the embodiment of FIG. 4, a single ball and cone seal 186, 196 is utilized at the bottom of the assembly. Leakage is limited at the top of the assembly by tight clearances between the retainer flange 168 and the upper internals barrel flange 154. The element is secured within the retainer by a snap ring. The embodiment of FIG. 5 is similar except that ball and cone seats are furnished at both ends of the flow element to further reduce bypass leakage. As alluded to above, a typical spring assembly may be compressed 1/2 to 1 inch when it is installed for service. As a consequence the flow sealing arrangement must accommodate that movement. A single ply standard spring bellows 184 is preferably incorporated in the assembly to perform this function. The assemblies of the design of either FIGS. 4 or 5 provide a simple and inexpensive means for establishing and controlling a coolant flow to the pressure vessel head region 132. If it is desired to reduce the flow, members having reduced or valved bores can be inserted by removing the snap ring 190 and inserting an appropriately dimensioned element in lieu of the bellows flange 182 of FIG. 4 or the upper plunger 204 of FIG. 5. In a third embodiment of the invention, illustrated in FIG. 6, a hold-down spring assembly 138 is disclosed intended for use in a pressure vessel assembly in which the travel limiting shims 146 of FIG. 2 have been omitted. In the embodiment of FIG. 6, the core barrel flange 144 has an increased axial thickness relative to the flanges of FIGS. 4 or 5 and the Belleville spring assemblies 138 are disposed in a generally cylindrically counter bore 209 formed in the flange 144 of the core barrel. This counter bore 209 has a shoulder 205 for supporting the Belleville spring stack. In this embodiment the flange not only supports the spring assemblies 138 but also functions to limit the maximum deflection of the assembly. The spring assemblies of FIG. 6 are designed so that when the upper internals and core are loaded into the pressure vessel, the resulting deflection of the spring assemblies 138 produces a gap 206 between the flanges 154 and 144. If the springs are overloaded due to a seismic event or the like or due to stress relaxation, the maximum additional deflection is limited to a distance equal to the gap 206 after which the flanges 144 and 154 abut one another. The gap 206 is typically on the order of about 60 to 90 mils. Other elements of FIG. 6 are similar to the assemblies of FIGS. 4 and 5 and are not further described in detail. As should now be appreciated, the spring assemblies of the present invention are relatively small and have a large deflection capability. In addition, the Belleville spring assemblies disclosed require less precise design and less manufacturing precision than a single or a few large spring assemblies. Consequently, the present invention can provide the required spring load and deflection capability at lower cost, afford easier replacement, handling, inspection and decontamination. In addition, multiple small spring assemblies provide a degree of redundancy which tolerates some assembly failures. Moreover, the present spring assembly design provides a generally fluid tight flow path for upper head cooling using a flow passage which is easily replaceable and adjustable to change the reactor head cooling flow rate. Importantly, the present small spring assemblies can be load tested before installation which is almost impossible with very large spring assemblies. In addition, the small diameter of the spring assemblies reduces the required reactor flange diameter by 4.5 inches or more with a consequent savings of some 26,000 pounds or more in reactor vessel weight. The foregoing description of the preferred embodiments of the invention have been presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise form disclosed, and obviously many modifications and variations are possible in light of the above teaching. For example, the size and shape of the spring stacks, the various retaining means, and type of coolant passage connections used may be modified within the spirit and scope of the invention. It is intended that the scope of the invention be defined by the claims appended hereto. |
summary | ||
051184613 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a flow rate measuring apparatus for measuring a flow rate of a fluid which is flowing in a tank by a plurality of pumps and, in particular, to an apparatus for accurately measuring a flow rate of a coolant, by a plurality of circulation pumps, in a pressure vessel in an nuclear reactor. 2. Description of the Related Art In a boiling-water reactor a coolant in a pressure vessel is circulated, by a plurality of circulation pumps, in the pressure vessel past a reactor core and it is necessary to accurately measure a flow rate, as well as a flow distribution, of the coolant circulated in the pressure vessel past the core and monitor the state of the reactor at all times. A general structure of a conventional boiling-water reactor and means for measuring a flow rate of the coolant will be explained below with reference to FIG. 1. The reactor contains a pressure vessel 1 with a core 2 held therein. The core 2 is held within a shroud 6 and a steam separator 3 is arranged over the shroud 6 to separate water from a steam generated at the core and supply it as dried steam to a turbine, etc. A coolant separated from the steam flows down a passage, defined between the outer periphery of the shroud 6 and the inner wall of the pressure vessel 1, onto a location under the core 2 by a plurality of circulation pumps 10 and goes up from under the core 2 into the core where it flows out of its upper zone after being boiled. The coolant circulates in such a passage as set forth above. The circulation pumps 10 are each driven, by the corresponding motor 11 outside the pressure vessel, through the corresponding shaft 9. In this type of reactor, it is necessary to precisely measure a flow rate of the coolant into the core and monitor the state of the reactor. A conventional means for measuring a flow rate of a coolant is so arranged as will be set forth below. Openings 25A, 25B of pipes for a plurality of sets of core plate differential pressure gauges 25 are opened at the inlet 14 of the core 2 and located at a core support plate or at an entrance nozzle of a fuel assembly. Pressure signals corresponding to pressure at these openings are sent to a differential pressure/flow converter 26 so that a flow rate of the coolant is measured. A plurality of pump differential pressure gauges 23 are also provided to correct and back up the core inlet differential pressure gauges 25. Openings 23A, 23B of pipes for the pump section differential pressure gauge 23 are provided at the suction and discharge sides, respectively, of the respective circulation pump 10. A differential pressure gauge 23 is inputted to a pump section calculator 24. The rotational speed of the motor 11 for driving the circulation pump 10 is measured by the corresponding speed transducer 22 and a speed signal output from the speed transducer 22 is inputted to the pump section calculator 24. With a differential pressure between the suction and discharge ports and the rotational speed of the pumps as parameters, relations between these parameters and the pump discharge are obtained in advance, using a test stand, for the respective circulation pump 10 and also have been programmed into the pump section calculator 24. The flow rate of the coolant through the pump 10 is calculated by the pump section calculator 24 and the respective circulation pump's flow rate output signal is gained from the pump section calculator 24 and inputted to a calculator 27 where a total flows rate of all the circulation pumps is obtained. A flow rate output signal of the calculator 27 and that of the differential pressure/flow rate rate calculator 26 are fed to an operation monitor device 29 of the reactor through a correction switch 28. By the switching operation of the correction switch 28, it is possible to make a readjustment of the core plate differential pressure gauge 25 and, in the case of an functional failure of this core differential pressure line, provide a backup function. A line of the aforementioned pump deck differential pressure gauge 23 reveals lowered accuracy in the case of a temporary stoppage, or a partial operation, of the circulation pumps 10. The reason for this will be explained below with respect to FIGS. 2A and 2B. FIGS. 2A and 2B are modified cross-sectional views showing a pressure vessel 1 at a height level where circulation pumps 10 are located. Reference numeral 12 in FIGS. 2A and 2B shows a cylindrical support leg 12 for a shroud with leg openings 13 located in front of the respective circulation pump 10 and also located at a middle area of respective adjacent circulation pumps. FIG. 2A shows a case in which all the circulation pumps 10 are operated with equal rotating speed with a coolant flowing in the directions indicated by open arrows in FIG. 2A. In this case, some of the discharge from the respective circulation pump 10 flows directly from the leg opening 13 which is in front of the respective circulation pump into a lower plenum and a remaining portion of the discharge from the respective circulation pump 10 horizontally flows in a circumferential direction. The latter coolant flow and a coolant flow coming from the next adjacent circulation pump meet at the middle area of both the pumps and flow into the lower plenum through the leg opening 13 which is located between the circulation pumps. FIG. 2B shows a case in which some circulation pump (for example, a circulation pump 10B) is out of service and the other circulation pumps 10A are operated with equal rotating speed, a coolant flowing in the directions indicated by solid arrows in FIG. 2B. In this case, the discharge flowing in the circumferential direction from the neighboring pumps 10A located at both side of the idle pump 10B flow toward the discharge side of the idle pump 10B. Some of the coolant flowing toward the discharge side flows into the lower plenum through the leg opening 13 in front of the idle pump 10B and others flow backward via the idle pump 10B. Even if, in this case, the flow rate of the discharge from the individual operating pumps are summed up it is not possible to accurately calculate a flow rate of coolant through the core. Although a simpler case has been explained above in conjunction with FIG. 2B for ease in understanding, some of circulation pumps may, in practical case, be stopped or a plurality of circulation pumps may be operated with different rotating speed. It has been difficult, in such a complicated case, to measure a flow rate of coolant precisely. The aforementioned problem arises upon the exact measurement of a flow rate of coolant in the pressure vessel in the reactor as well as in other apparatus and chemical plants. In a heat exchanger, a boiler, a agitating apparatus in a chemical plant, a dialyzer, a gas reaction apparatus, a solid/fluid separation apparatus etc, a complex fluid flow/circulation passage is formed inside of vessel and a fluid flow through the passage. Even in these apparatuses, it is difficult to exactly measure the state of a fluid flow in the vessel in an off-normal operational condition and there is a growing demand for an apparatus of accurately measuring a flow rate of a complex fluid flow in a vessel, or a container, as in the aforementioned nuclear reactor. SUMMARY OF THE INVENTION It is accordingly the object of the present invention to provide a flow measuring apparatus which, in an apparatus for flowing a fluid in a vessel or container, by a fluid circulation unit, along a predetermined passage in the vessel or container, accurately measures a flow rate of such a fluid even if the fluid flows in a flow pattern deviated from a normal one. According to the present invention, a rotational speed detectional unit detects the rotational speed of a motor for driving a fluid circulation unit and delivers a corresponding rotational speed signal to a pipe network calculation unit which is stored with an initially programmed pipe network model equivalent to a real passage of the fluid in the container. The pipe network constants are set corresponding to the fluid circulation unit on the pipe network model to a value corresponding to the rotational speed of the motor for driving the fluid circulation unit, and analytically calculates a flow rate at any point on the pipe network model, from this result it is possible to measure a flow rate of the fluid at corresponding point in the container. Since the present apparatus derives a flow rate on the pipe network with the flow passage in the container replaced as the pipe network model, it is possible to simply and accurately calculate an actual flow rate at any portion of the passage in the container. Even if the passage of the fluid in the container is complicated or a plurality of fluid circulation means are located in the passage, it is possible to simply and accurately calculate the flow rate at any portion of in the passage in the container. According to a preferred embodiment of the present invention, the present invention is advantageously applied to an apparatus which measures a flow rate of a coolant circulated in a pressure vessel for a boiling-water reactor in particular. The pressure vessel contains a core, a shroud, shroud support legs and so on, by which a passage is formed to allow the coolant to circulate by circulation pumps. That passage is not of such a type as to be surrounded completely with a wall and hence of a complex type in flow pattern so that the circulation of the coolant is conducted by a plurality of circulation pumps. These pumps cause a back-flow upon being partially stopped, resulting in a complicated pattern of a coolant flow in the pressure vessel. It is thus difficult to measure a actual flow of the coolant at any portion in the pressure vessel. In the measurement of a flow rate in the pressure vessel of the reactor, according to the present invention, a pipe network model is prepared based on the actual passage of the coolant in the pressure vessel with the rotating speed of the corresponding circulation pump replaced with corresponding variables of the pipe network model and a coolant flow is calculated based on the pipe network model whereby it is possible to simply and accurately measure a flow rate of the coolant in the pressure vessel. According to another preferred embodiment of the present invention, a result of measurement is fed back to a circulation unit, for example, to a control unit of the circulation pump in the reactor and, by so doing, it is also possible to automatically maintain a flow rate of a coolant in the reactor at a predetermined level. According to this kind of correction, therefore, one or some of circulation pumps are operated at a low rotating speed or stopped, a drop in the flow rate of the coolant is measured, corresponding of the result, the rotating speed of other circulation pumps are increased automatically in accordance with a result and maintain the flow rate of the coolant through the core at a predetermined normal level. It is thus possible to enhance the reliability of the operation and the control of the reactor plant. |
043538630 | abstract | There is described a method for localizing a leaking rod from a nuclear fuel assembly, in which for each rod in an assembly, the radio-activity in at least two discrete rod rows in which said rod lies is measured and a leaking rod is localized by sensing a lowering of the radio-activity in the tested rows where said rod lies relative to the radio-activity in an identical row of non-leaking rods.. There is further described an equipment for the working of this method. |
053135059 | abstract | ROD HANDLING APPARATUS comprises a continuously operated modified walking beam conveyor apparatus activated mechanically, electrically and electronically, utilizing computer and servomechanism technology to control the movement of a series of nuclear fuel rods with provision for their varying length and diameter so that their ends may be heat treated simultaneously. Rods are loaded into a conveyor and rest on conveyor V-blocks attached to conveyor chains which are then moved forward, powered and controlled by a conveyor servomotor, to align the rods with a heat treater units. Then the conveyor stops and a separate set of lifting V-blocks controlled by an eccentric arm powered by a lift plate servomotor, raises the rods to align them with receptacles in the heat treater units. The heat treater units move transversely so that the ends of the rods are inserted a set distance into the receptacles and the heat treating cycle is started. After completion of this cycle, the heat treater units are retracted and the lift plate V-blocks lower the rods into the conveyor. The conveyor again moves forward, unloading the treated rods and the cycle is repeated. This apparatus may be used for processing the ends of nuclear rods and other type of rods, tubes, bars, pipes, etc., for other operations including welding, coding, polishing, deburring, attaching fittings, etc., and, with the addition of processing units on the opposite side of the conveyor, both ends of the rods can be operated on simultaneously. |
abstract | A dry FCVS for a nuclear reactor containment is provided. The dry FCVS includes a housing and a round and/or elongated aerosol filter inside the housing for removing contaminant aerosols from gas passing through the housing during venting of the containment. The housing includes at least one inlet portion configured for directing gas into the aerosol filter during the venting of the containment and an outlet portion for gas filtered by the aerosol filter during the venting of the containment. The dry filtered containment venting system is arranged and configured such that when a flow of gas through the outlet portion is closed off at least one of convective, radiant and conductive heat transfer removes decay heat of aerosols captured in the aerosol filter. |
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abstract | A membrane mask for use in an electron beam lighography or X-ray lighography has a membrane film formed on a silicon wafer, and a mask body pattern formed on the membrane film. The membrane film has a heavy-metal-implanted area underlying a portion of the mask body pattern other than the opening of the mask body pattern. The implanted area achieves a higher contrast ratio in the pattern obtained from the membrane mask. |
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043604960 | abstract | Cooling system for auxiliary systems of a nuclear installation for heat removal from heat exchangers, the heat exchangers being connected on the primary side thereof to lines which may contain radioactive liquids or gases, being disposed in a secured area of the nuclear installation, and having connections on the secondary side thereof for cooling liquid lines, including an outgoing line for the cooling liquid connected to the connections on the secondary side of the heat exchangers, a dry cooling tower having cooling elements connected to the outgoing line, a return line for the cooling liquid connected to the cooling elements, a refrigeration loop having a supplemental heat exchanger with the primary side thereof connected in the return line, a bypass line connected from the outgoing to the return line parallel to the cooling elements and supplemental heat exchanger, and a control valve connected in the bypass line. |
056339020 | claims | 1. A method for dismantling a storage rack used to store nuclear fuel assemblies, said storage rack including a base joined to an array of tubular or round storage cells having fuel entry ends which are opposite to said base for a fuel assembly, said method including the steps of: pressing a metal cutting saw blade against an interior wall of a selected one of said storage cells at a site spaced from said base, moving said metal cutting saw blade in a direction outwardly from the interior of the tube to cause said saw blade to initially puncture and extend partly through said selected storage cell, and thereafter shifting said metal cutting saw blade about a cutting path from the puncture site within a storage cell generally transverse to the length of the storage cell to cut the remaining portion of the selected storage cell wall from said base. introducing said metal cutting saw blade horizontally along the selected one of said tubular storage cells; anchoring said metal cutting saw against horizontal movement in the selected one of said tubular storage cells and thereafter performing said step of pressing a metal cutting saw blade against an interior wall. selecting a metal cutting saw having a cutter blade driven about an axis of rotation displaceable between an inoperative location and an operative location; passing said metal cutting saw through the entry end of a selected tubular storage cell forming part of said storage rack to a site at a preselected space from said base; energizing said metal cutting saw; displacing said cutter blade from said inoperative location to an operative location causing said saw blade to cut into a wall segment of the selected storage cell; and orbiting the rotational axis of said cutter blade about a path within the confines of the selected tubular storage cell to sever the remaining portion of the wall section of the selected cell from said base. a support housing including means to rotatably support a drive shaft, said drive shaft having driven and drive shaft portions extending from opposite ends of said support housing; means for anchoring said support housing within the tubular configuration of a selected one of said tubular storage cells; a carrier supported by the driven shaft portion of said drive shaft; a metal cutting saw including a cutter blade rotatably driven at one end of a motor housing; means for supporting said motor housing on said carrier for displacement of the motor housing transverse to a rotational axis of said drive shaft for causing said saw blade to puncture a wall portion of such a tubular storage cell; and means for rotating the drive shaft for shifting said metal cutting saw blade from the puncture site about a storage cell cutting path generally transverse to the length of the storage cell to cut the remaining portion of the selected storage cell wall from said base. 2. The method according to claim 1 wherein said step of shifting said step of said cutting saw blade includes shifting the saw blade both laterally and transversely in the plane of movement by the saw blade from said puncture site. 3. The method according to claim 1 including the further step of supporting said storage rack on horizontal supports with the extended length of the tubular storage cells extending in a horizontal direction; 4. The method according to claim 3 wherein said storage rack is placed on said horizontal supports within a contained area and filtering air circularly in the contained area in the HEPA filtration system to extract airborne particulates contaminated with low level radiation. 5. The method according to claim 1 wherein said step of shifting said cutting saw blade includes displacing the rotational axis of the metal cutting saw blade radially from the central axis of the selected storage cell and thereafter swinging the rotational axis of the saw blade about an arc whose center coincides with the pivot axis of a pivot shaft. 6. The method according to claim 1 including the further step of positioning said metal cutting saw blade at a site remotely to said puncture site at a selected storage cell near the fuel entry end thereof and again performing said step of pressing the metal cutting saw blade for severing a tubular storage wall section of the selected tubular storage cell free of support in said storage rack. 7. The method according claim 6 including the further step of reducing the volume of space by said tubular storage wall section. 8. The method according to claim 7 wherein the said step of reducing the volume includes crushing said tubular storage wall section between opposed dies by moving one die toward each other. 9. The method according to claim 7 wherein said step of reducing the volume includes sub-dividing said tubular storage wall sections to form a plurality of waste wall sections having a size and weight capable of manual manipulation by workman. 10. A method for dismantling a storage rack used to store nuclear fuel assemblies, said storage rack including a base joined to an array of tubular storage cells having fuel entry ends which are opposite to said base for a fuel assembly, said method including the steps of: 11. Apparatus for dismantling racks used to store nuclear fuel assemblies, said storage rack including a base joined to an array of tubular storage cells having fuel entry ends which are opposite to said base for a fuel assembly, said apparatus including a combination of: 12. The apparatus according to claim 11 the means for anchoring include piston and cylinder assemblies carried by said support housing to anchor the said support housing within a said selected one through said storage cells. 13. The apparatus according to claim 11 wherein said means for supporting includes a linear bearing and an actuator for displacing said metal cutting saw relative to said drive shaft. 14. The apparatus according to claim 13 wherein said means for supporting includes an L-shaped bracket supported by said drive shaft and supporting said linear bearing for defining a displacement distance between the rotational axis of said drive shaft and rotational axis of said metal cutting saw. 15. The apparatus according to claim 14 wherein said supporting housing has four sidewalls and wherein said means for anchoring said support housing comprises piston and cylinder assemblies arranged on adjacent ones of two sidewalls of said support housing for anchoring said housing between opposing sidewall of tubular storage cells having a square cross-section configuration. |
description | This application claims priority from Japanese Patent Application No. 2014-135066, filed on Jun. 13, 2014, the entire subject matter of which is incorporated herein by reference. 1. Technical Field The present disclosure relates to an X-ray fluorescence analyzer. 2. Description of the Related Art X-ray fluorescence analysis is a technique of irradiating a sample with an X ray emitted from an X-ray source, and detecting fluorescent X-rays emitted from the sample by an X-ray detector, and performing qualitative analysis of the sample or quantitative analysis of a concentration, a film thickness, or the like based on the intensities of the fluorescent X-rays. The constituent elements of the sample generate fluorescent X-rays having energies characteristic of the individual elements. Therefore, the spectra of the measured X rays are searched for peaks of the fluorescent X-rays characteristic of the individual elements, whereby it is possible to find out which elements are contained. This analysis is called qualitative analysis. Meanwhile, quantitative analysis uses a fact that the intensity of the fluorescent X-ray of each constituent element which is obtained is determined by the relation between the state of the X ray with which a sample is irradiated and the amount of the corresponding element existing in the irradiation area. Specifically, first, a sample is irradiated with X rays having known X-ray states, that is, known X-ray intensities for each energy, and the intensities of fluorescent X-rays of individual elements generated as the results of the irradiation is measured, and then the amounts of the elements capable of making X rays with those intensities be generated is calculated. In this calculation process, various methods are used. In every method, for the accuracy of the analysis, it is important that the amount and energy distribution of X-ray for irradiation matches with the premise of calculation. In concentration analysis, it is possible to cancel the influence to a certain extent in a case where the entire intensity increases while the energy distribution is kept. However, in film thickness calculation, it is difficult to distinguish between increase or decrease in intensity and increase or decrease in film thickness, and thus the influence of variation in X ray for irradiation is more serious. Since it is practically almost impossible to prepare an X-ray irradiation system which is completely stable and has a precisely defined energy distribution, a technique of measuring a sample having a known composition and a known structure and calibrating a device using the intensity of an X ray obtained from the measurement is generally used. Therefore, device calibration is a very important technique for accurate quantitative analysis and being required to be performed routinely and accurately. Due to this request, there has been disclosed a configuration in which a sample for calibration is mounted on a shutter member and automatically performs calibration. An example of such configuration is disclosed in Japanese patent publication No. JP-A-S59(1984)-067449. In X-ray fluorescence analysis, the size of an area to be measured is often to be one of major concerns. In general, the measurement area is defined by restricting an X-ray irradiation area. According to the sizes of samples which are measurement targets, various devices having irradiation areas in the order of several millimeter to several tens micrometer are provided. Among those devices, devices having particularly small irradiation areas use advanced technologies of a collimator which is a structure capable of blocking X rays and having tiny holes, a capillary X-ray optical element for focusing X rays on a tiny area by using a total reflection phenomenon of the inner surface of a hollow glass fiber, and the like. It is known that a misalignment in the positional relation between a collimator or capillary X-ray optical element and an X-ray generator influences the intensities and energy distributions of X rays to be emitted. That is, it can be said that the positional relation of the constituent elements of each of those devices is an element which should be calibrated to have accurate quantitative analysis. The configuration disclosed in the Japanese patent publication No. JP-A-S59(1984)-067449 allows to move a calibration sample mounted on a shutter member into an X-ray irradiation area, and calibrate intensity variation attributable to variation of an X-ray source for irradiation with an X ray, but the route of the X ray and the incident direction of the X ray to a detector are different from those of normal measurement. For this reason, there may be a problem that, in the strict sense, a measurement condition for calibration is different from a normal measurement condition, and it is not possible to calibrate variations attributable to every element. Further, in devices using X-ray focusing elements such as poly-capillaries recently being in widespread use, a small position gap between a focusing element and an X-ray generating portion of an X-ray source is attributable to intensity variation, even if these mechanisms are disposed above the focusing element, device calibration inevitably may become incomplete. Furthermore, in general, in order to reduce an X-ray irradiation area, it is preferable to reduce a distance between a focusing element and a sample. Therefore, it may be difficult to dispose a calibration below the focusing element. For these reasons, devices using X-ray focusing elements may have a problem that they do not have means for automatically calibrating the devices. The present disclosure has been made in view of the above-described circumstances, and one of objects of the present disclosure is to provide an improved X-ray fluorescence analyzer. According to an exemplary embodiment of the present disclosure, there is provided an X-ray fluorescence analyzer including: an X-ray source; an irradiation area restricting member that restricts an area of a measurement sample to be irradiated with an X ray emitted as a primary X-ray from the X-ray source; a detector that detects a secondary X-ray generated from the measurement sample; a calibration sample for calibrating a device including the X-ray source, the irradiation area restricting member, and the detector; a sample stage on which the measurement sample is mounted, the sample stage arranging the measurement sample at a position in which a surface of the measurement sample faces the irradiation area restricting member with a predetermined distance from the irradiation area restricting member on an irradiation axis of the primary X-ray; and a calibration sample moving mechanism that holding the calibration sample and moving the calibration sample between a retraction position which is an arbitrary position deviated from a route of the primary X-ray and a spatial position that is the same with an irradiation position of the primary X-ray on the surface of the measurement sample spaced apart from the irradiation area restricting member by the predetermined distance, the calibration sample moving mechanism being configured to be independent from the sample stage. According to another exemplary embodiment of the present disclosure, there is provided an X-ray fluorescence analyzer including: a measurement device having: an X-ray source that emits an X-ray; an irradiation area restricting member that restricts an area of a measurement sample to be irradiated with the X-ray as a primary X-ray; and a detector that detects a secondary X-ray generated from the measurement sample by being irradiated with the primary X-ray. The X-ray fluorescence analyzer further includes: a sample stage that holds and moves the measurement sample between a measurement position at which the measurement sample is irradiated with the primary X-ray to detect the secondary X-ray by the detector and a first retracted position at which the measurement sample is retracted from the measurement position; and a calibration sample moving mechanism that holds a calibration sample for calibrating the measurement device and moves the calibration sample between the measurement position and a second retracted position at which the calibration sample is retracted from the measurement position. Hereinafter, an embodiment of an X-ray fluorescence analyzer according to the present disclosure will be described with reference to FIGS. 1 and 2. As shown in FIG. 1, an X-ray fluorescence analyzer 1 according to the present embodiment includes an X-ray source 11, an irradiation area restricting member 12, a detector 13, a calibration sample moving mechanism 16, and a sample stage 19. As the X-ray source 11, a versatile X-ray tube universal for X-ray fluorescence analyzers being generally sold may be used. As the irradiation area restricting member 12, a poly-capillary may be used. In a case where an irradiation diameter is fixed, poly-capillaries can perform irradiation with X rays having higher intensities as compared to collimators which restrict X-ray irradiation areas by simple apertures. Therefore, use of poly-capillaries in cases where high-accuracy measurements are required is increasing. In situations where high-accuracy measurements are required, there are high requirements with respect to device calibration. Therefore, those situations are suitable as examples to which the present disclosure is applied. The detector 13 is configured by a detection element, a pre-amplifier, a digital pulse height analyzer, and a spectrum memory. In response to incidence of an X ray, the detection element generates electric charge proportional to the energy of the X ray. The generated electric charge is output as a voltage signal through the pre-amplifier, and the digital pulse height analyzer converts the voltage signal output from the pre-amplifier into a sequence of digital values arranged in chronological order, and calculates the incidence timing and energy of the X ray. If X ray incidence is detected, a channel count value of the spectrum memory corresponding to the calculated energy is added. The spectrum memory is cleared when measurement starts, and when measurement finishes, addition into the spectrum memory is stopped. In the present embodiment, since the intensity of the output of the poly-capillary is high, as a detection element which is unlike to be saturated before a counting rate becomes high, a semiconductor drift type detector is used. As a calibration sample 14, various samples can be used as long as their compositions and structures are known. In the present embodiment, a zirconium plate is used. Based on the intensity of a fluorescent X-ray which is generated from zirconium, variation in the intensity is calibrated, and based on the position of a peak of the fluorescent X-ray of zirconium, the relation between the channel of the spectrum memory and energy is calibrated. When calibration is required, the calibration sample moving mechanism 16 avoids interference between a measurement sample 8 and the calibration sample 14 while moving the calibration sample 14 such that the calibration sample 14 is disposed at the position of a primary X-ray irradiation position 15 of the measurement sample 18. Therefore, the calibration sample moving mechanism 16 is attached to the calibration sample moving mechanism 16. Subsequently, an operation related to an actual configuration will be described. The calibration sample moving mechanism 16 is configured using a linearly movable mechanism, a rotation mechanism (not shown), and the like, so as to be able to move the calibration sample 14 between differentiation learning function two different positions, that is, a retraction position and a measurement position. The two positions are a primary X-ray irradiation position 15 which is on a primary X-ray irradiation axis 21 which is configured by the X-ray source 11 and the irradiation area restricting member 12, and a retraction position 20 which is not irradiated with an X ray. Normally, the calibration sample 14 is disposed at the retraction position 20. Also, the calibration sample moving mechanism 16 is provided so as to be structurally independent from the sample stage 19. Therefore, the operable range of the calibration sample moving mechanism is not influenced by the shape of a sample 18 being on the sample stage 19. The sample stage 19 has a structure to be used with a sample 18 as a measurement target, which is mounted on the sample stage 19. The sample stage 19 is driven along three axes perpendicular to one another, that is, X, Y, and Z axes. One axis of the three axes is disposed substantially parallel to the primary X-ray irradiation axis 21. Adjustment of the position of a sample 18 first needs movement in a plane perpendicular to the primary X-ray irradiation axis 21 such that a desired position of the sample can be irradiated with a primary X-ray. Further, movement in the direction of the primary X-ray irradiation axis 21 influences the irradiation area size and irradiation intensity of the primary X-ray and the sensitivity of detection of a secondary X-ray generated. For this reason, the primary X-ray irradiation position 15 is always adjusted such that the surface (measurement target portion) of the measurement sample 18 keeps constant distances (heights) from the irradiation area restricting member 12 and the detector 13. As shown in FIGS. 1 and 2, since the positions of the irradiation area restricting member 12 and the detector 13 are fixed, the primary X-ray irradiation position 15 can be determined only by specifying the position of the height h1 of the lower surface of the irradiation area restricting member 12 from the upper surface of the measurement sample 18. Therefore, the primary X-ray irradiation position 15 is a position specified in a three dimensional space. In a case of moving the calibration sample 14 to the primary X-ray irradiation position 15 as described above in order to perform calibration, it is confirmed that the calibration sample 14 will not collide with the measurement sample 18. If necessary, the sample stage 19 is operated to retract the measurement sample 18. As shown in FIG. 2, the retraction position is arbitrarily specified according to the distance (height) h2 of the upper surface of the measurement sample 18 from the lower surface of the irradiation area restricting member 12 which does not collide with the calibration sample 14 in a case where the measurement sample 18 is moved by the calibration sample moving mechanism 16. Next, the calibration sample moving mechanism 16 is operated, whereby the calibration sample 14 is moved to the primary X-ray irradiation position 15 which is the position of the distance (height) h1 from the lower surface of the irradiation area restricting member 12. By doing this, the calibration sample 14 can be irradiated with the primary X-ray in the substantially same condition as that during measurement of the measurement sample 18. Therefore, if the detector 13 is instructed to start, the spectrum of the calibration sample 14 according to the substantially same condition as that during measurement of the measurement sample 18 is integrated in the spectrum memory. If a predetermined measurement time comes, the measurement is finished, and the spectrum of the calibration sample 14 is obtained. From this spectrum, the intensity of the fluorescent X-ray of zirconium which is a calibration element is read. Then, based on the magnitude of the intensity, the sensitivity of the spectrometer is calibrated. After the measurement for calibration, the calibration sample moving mechanism 16 is operated again, whereby the calibration sample 14 is moved to the retraction position 20 again, and then the calibration operation finishes. In addition to the embodiment described until now, a controller 17 capable of controlling the X-ray source 11, the detector 13, and the calibration sample moving mechanism 16 may be provided, whereby it is possible to automatically perform an operation necessary for calibration. For example, the controller can be programmed so as to perform calibration at predetermined time intervals, whereby it is possible to always the fluorescent X-rays in a good state. In a case of automatically performing calibration, it is required to avoid collision with the measurement sample 18 placed on the sample stage 19. Specifically, the controller 17 is programmed to operate the calibration sample moving mechanism 16, whereby the calibration sample 14 is moved from the calibration sample retraction position 20 to the primary X-ray irradiation position 15, while checking whether the measurement sample 18 is at the primary X-ray irradiation position 15, and automatically retract the sample stage 19 to a safe position (the distance (height) h2 of the upper surface of the measurement sample 18 from the lower surface of the irradiation area restricting member 12) in a case of determining that the calibration sample 14 would collide with the measurement sample 18. Thereafter, a calibration operation is automatically performed, and then the sample stage 19 is returned to the original position. In the embodiment described here, the sample stage 19 movable along three axes perpendicular to one another is used. However, methods of vertically moving the whole of a structure including the sample stage for moving a sample along two axes in a horizontal plane, the X-ray source 11, the irradiation area restricting member 12, the detector 13, and the calibration sample moving mechanism 16, thereby adjusting the distance from the sample 18, can be considered. These methods are also included in the disclosure range of the present disclosure. Also, in the embodiment, in order to describe a specific operation, with respect to the individual constituent elements, specific examples have been given and described. However, forms to be used as the X-ray source 11, the irradiation area restricting member 12, the detector 13, and the sample stage 19 does not influence defining of the present disclosure. Also, even with respect to which kind of calibration information to be calculated from the material of the calibration sample 14 and the spectrum obtained from the calibration sample, various modifications are possible mounting of the spectrometer. As described with reference to the exemplary embodiment, according to a first mode of the present disclosure, there is provided an X-ray fluorescence analyzer including: an X-ray source; an irradiation area restricting member that restricts an area of a measurement sample to be irradiated with an X ray emitted as a primary X-ray from the X-ray source; a detector that detects a secondary X-ray generated from the measurement sample; a calibration sample for calibrating a device including the X-ray source, the irradiation area restricting member, and the detector; a sample stage on which the measurement sample is mounted, the sample stage arranging the measurement sample at a position in which a surface of the measurement sample faces the irradiation area restricting member with a predetermined distance from the irradiation area restricting member on an irradiation axis of the primary X-ray; and a calibration sample moving mechanism that holding the calibration sample and moving the calibration sample between a retraction position which is an arbitrary position deviated from a route of the primary X-ray and a spatial position that is the same with an irradiation position of the primary X-ray on the surface of the measurement sample spaced apart from the irradiation area restricting member by the predetermined distance, the calibration sample moving mechanism being configured to be independent from the sample stage. According to this configuration, it is possible to irradiate a calibration sample with an X ray having passed through all components through which an X ray needs to pass until a normal measurement sample is irradiated with the X ray, and the way how the influence of variation of the spectrometer appears in the result of measurement of the calibration sample and the way how the influence of variation of the spectrometer appears in the result of measurement of a normal measurement sample become very similar. Therefore, accurate and precise spectrometer calibration becomes possible. According to a second mode of the present disclosure, the X-ray fluorescence analyzer according to the first mode may further include: a controller that controls the X-ray source, the detector, and the calibration sample moving mechanism; and an automatic device calibrator that operates the calibration sample moving mechanism to move the calibration sample to the spatial position that is same with the primary X-ray irradiation position of the measurement sample, and operates the detector to detect a secondary X-ray which is generated by irradiating the calibration sample with the X ray emitted from the irradiation area restricting member, and automatically calibrates the device based on a result of detection f the secondary X-ray. According to this configuration, it is possible to automatically perform a procedure from calibration sample disposition to calibration measurement, and the spectrometer can keep a correctly calibrated state, without making a user do special work. According to a third mode of the present disclosure, in the X-ray fluorescence analyzer according to the second mode, the sample stage may be provided with a sample stage moving mechanism that moves in at least one direction of a three-dimensional directions of X, Y, and Z directions, and the controller may control the sample stage moving mechanism to retract the measurement sample to a position where the calibration sample does not interfere with the measurement sample before operating the calibration sample moving mechanism. According to this configuration, due to automation of calibration, it is possible to move the calibration sample and the measurement sample without interference between them, and it is possible to realize stable measurement over a long time. According to the X-ray fluorescence analyzer having the above described configuration, before normal measurement or at arbitrary intervals in the course of measurement, it is possible to automatically irradiate the calibration sample with an X ray emitted from a focusing element such as a collimator or a poly-capillary under the almost same condition as a normal sample measurement condition, thereby calibrating the spectrometer. Therefore, it is possible to perform long-term unattended measurement while performing accurate calibration, and significant improvement of work efficiency becomes possible without degrading measurement accuracy. |
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052951650 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Disclosed hereinbelow is a self-locking plug for plugging a hole defined by a surrounding structure, which structure may be a nuclear power reactor pressure vessel core barrel flange. The core barrel flange has an access hole transversely therethrough for providing access to a neutron radiation detector or activation foil disposed in an annular space defined between the core barrel and a surrounding pressure vessel shell. The plug closes the hole when access to the detector is not required in order to prevent the coolant from flowing through the hole and bypassing a reactor core disposed in the pressure vessel shell. However, the plug is removed from the hole when access to the radiation detector is desired. In this regard, when access to the detector is desired, an upper support structure, which is mounted atop the core barrel, is removed. When the upper support structure is removed, transient hydraulic forces that may occur in the pressure vessel will tend to dislodge the plug from the hole prior to it being grasped and controllably removed. Such inadvertent dislodgement of the plug from the hole prior to being grasped and controllably removed is undesirable because such a plug may become a "loose part" that may migrate in the coolant to damage fuel assemblies and other internal reactor components. Therefore, a problem in the art is to provide a plug that is resistant to being inadvertently dislodged even in the presence of such transient hydraulic forces and that is readily controllably removed when access to the detector is desired. According to the invention, this problem is solved by providing a self-locking plug that is resistant to begin inadvertently dislodged and that is also readily controllably removed to access the radiation detector. However, before describing the subject matter of the present invention, it is instructive first to briefly describe the structure and operation of a typical nuclear power reactor and its associated core barrel flange. Therefore, referring to FIG. 1, there is shown a typical nuclear power reactor, generally referred to as 10, for producing heat by controlled fission of fissionable material (not shown). Reactor 10 includes a vertically-oriented and generally cylindrical reactor pressure vessel shell 20 open at its top end and having a plurality of inlet nozzles 30 and outlet nozzles 40 attached to the upper portion thereof (only one of each nozzle is shown). A generally hemispherical reactor vessel closure head 40 is mounted atop vessel shell 20 and is sealingly attached to the open top end of vessel shell 20, such as by a plurality of hold-down bolts 60, so that closure head 50 sealingly caps vessel shell 20. Capping vessel shell 20 in this manner allows for suitable pressurization of the coolant (not shown) circulating through vessel shell 20 as reactor 10 operates. The coolant may be borated demineralized water maintained at a relatively high pressure of approximately 2500 psia and circulated at an average velocity of approximately 30 feet per second. Still referring to FIG. 1, disposed in reactor 10 is a nuclear reactor core, generally referred to as 70, comprising a plurality of nuclear fuel assemblies 80 containing the fissionable material. Extending through the top of closure head 50 are a plurality of control rod drive mechanisms (CRDMs) 90 having control rod clusters (not shown) connected thereto. The control rod clusters extend into their respective fuel assemblies 80 for controlling the fission process in reactor core 70. As reactor 10 operates, the coolant enters vessel shell 20 by means of inlet nozzle 30 and circulates therethrough generally in the direction of arrows 95. As the coolant circulates through vessel shell 20, it also circulates over fuel assemblies 80 for assisting in the fission process and for removing the heat produced by fission of the fissionable material contained in fuel assemblies 80. After circulating through vessel shell 20, the coolant leaves vessel shell 20 by means of outlet nozzle 40. The heat carried by the coolant exiting outlet nozzle 40 is ultimately transferred to a turbine-generator for producing electricity in a manner well known in the art of electricity production from nuclear power. Referring to FIGS. 1, 2, and 3, a generally cylindrical core barrel 100 is concentrically disposed in vessel shell 20 for supporting fuel assemblies 80 thereon. Vessel shell 20 and core barrel 100 define an annular space 110 therebetween. Disposed in space 110 and adjacent vessel shell 20 is at least one retrievable radiation sensitive detector 120, which may be a neutron radiation detector or activation foil, connected to core barrel 100 such as by bracket 130. Radiation detector 120 is sensitive to the neutron flux penetrating vessel shell 20, which neutron flux is produced by operation of reactor core 70. Core barrel 100 has an integrally attached core barrel flange 140 extending therearound for connecting core barrel 100 to vessel shell 20. Flange 140 has an undersurface 145 and a top surface 147. Formed through flange 140 is at least one access port or hole 150 generally aligned with detector 120 for providing access to detector 120. Hole 150 may be a counter-bore or step bore having a first diameter and also having a second diameter less than the first diameter. The first diameter and second diameter define an annular depending shoulder or land 155 at the interface thereof. Allowing access to the radiation detector 120 allows it to be retrieved for analysis in order to determine the integrated neutron flux seen by vessel shell 20 during reactor operation. Still referring to FIGS. 1, 2 and 3, spaced-apart but atop core barrel 100 and connected to vessel shell 20 is a generally cylindrical upper support structure 160 for supporting upper internal reactor components (e.g., CRDMs), which components are generally referred to as 170. Upper support structure 160 has a bottom surface 165 thereon. Thus, core barrel 100 and support structure 160 define a space 180 therebetween because core barrel 100 is spaced-apart from upper support structure 160. Disposed in space 180 and thus interposed between upper support structure 160 and core barrel 100 is a hold-down spring 190 having a generally S-shaped transverse cross section for exerting a downwardly biasing force on core barrel 100 and an equal but opposite upwardly biasing force on upper support structure 160. Thus, it will be appreciated from the description immediately hereinabove that hold-down spring 190 maintains core barrel 100 and upper support structure 160 in their predetermined spaced-apart relationship. Referring to FIGS. 2, 3, 4 and 5, a self-locking plug, generally referred to as 200, is there shown for plugging hole 150 which is defined by its surrounding structure (i.e., core barrel flange 140). Plug 200 comprises a generally cylindrical plug body 210 having a distal end portion 220 and a proximal end portion 230. As used herein, the terminology "proximal end portion" means that end portion located nearer reactor core 70 and the terminology "distal end portion" means that end portion located further away from reactor core 70. Plug body 210 may be formed of a corrosion resistant material, such as "TYPE 304" stainless steel, for resisting the corrosive effects of the borated demineralized water coolant. In this regard, "TYPE 304" stainless steal comprises by weight percent approximately 0.08% carbon, 20% chromium, 11% nickel, and 78.92% iron. Distal end portion 220 of plug body 210 defines a circumferentially extending first groove 240 therein for reasons provided hereinbelow. The transverse contour of first groove 240 defines an upper stop 250 and a lower stop 260 for reasons disclosed more fully hereinbelow. Distal end portion 220 also defines a circumferentially extending second groove 270 therein for reasons provided hereinbelow. The innermost circumference of second groove 270 is larger than the innermost circumference of first groove 240 for defining an annular depending plug body shoulder 280 therebetween. Second groove 270 also defines a shelf 290 in plug body 210 for reasons provided hereinbelow. Moreover, proximal end portion 230 has an outermost diameter less than the outermost diameter of distal end portion 220. The outermost diameter of distal end portion 220 is sized so that is capable of being matingly disposed in the smaller diameter of hole 150 and so that the bottom of distal end portion 220 is capable of being seated on land 155. In addition, proximal end portion 230 includes an integrally attached cone-shaped nose 300 for allowing plug body 210 to be easily slidably inserted or guided into hole 150. Still referring to FIGS. 2, 3, 4 and 5, distal end portion 220 defines a longitudinal bore 310 therein which terminates in a slot 320 formed transversely through proximal end portion 230. Thus, slot 320 is in communication with bore 310. Proximal end portion 220 also defines a channel 330 transversely therethrough for reasons disclosed hereinbelow. Moreover, connected to plug body 210 is locking means, such as a locking mechanism, generally referred to as 340, for locking plug body 210 to core barrel flange 140. In the preferred embodiment, locking mechanism 340 includes a plurality of parallel, spaced-apart and generally cylindrical pivot pins 350 outwardly projecting from plug body 210 in slot 320. A plurality of catches 360 are disposed in slot 320, each catch 360 being pivotally connected to a respective one of the pins 350. In the preferred embodiment of the invention, there are two catches 360. Each catch 360 includes an integral outwardly turned dog-leg shaped hook portion 370 at one end thereof for engaging undersurface 145 of flange 140. The other end of each catch 360 includes a contoured portion defining a cam surface 380 thereon for reasons provided hereinbelow. Cam surface 380, which may be a generally V-shaped notch, has a relatively steeply inclined (e.g., ten degrees with respect to vertical) portion 385 for reasons disclosed hereinbelow. Interposed between catches 360 is first biasing means, such as a first coiled spring 390 for outwardly biasing at least a portion of each catch 360, so that hook portions 370 are outwardly biased to align with undersurface 145 in order to engage undersurface 145. Spring 390 may be "INCONEL 600" comprising by weight percent approximately 76.0% nickel, 0.08% carbon, 0.5% magnesium, 8.0% iron, 0.008% sulfur, 0.25% copper, and 15.5% chromium for resisting the corrosive effects of the borated demineralized coolant. Referring again to FIGS. 2, 3, 4 and 5, slidably extending through bore 310 and into slot 320 is camming means, such as an elongate cam 400 having a passageway 410 transversely therethrough for reasons disclosed hereinbelow. Cam 400 has a distal end portion 420 which may have external threads for reasons disclosed presently. Cam 400 also has a proximal end portion 430. Proximal end portion 430 defines an external surface of predetermined contour for slidably matingly engaging cam surface 380 belonging to catch 360. Interposed between catches 360 and coaxially aligned with cam 400 may be a center stop 435, center stop 435 being integrally formed with plug body 210. The purpose of center stop 435 is to limit or stop the downward travel of cam 400. Due to the predetermined position of center stop 435, cam 400 downwardly travels only to the extent necessary to fully engage cam surface 380 at which time proximal end portion 430 of cam 400 will abut center stop 435. Cam 400 includes second biasing means, such as a second coiled spring 440, surrounding elongate cam 400. Second coiled spring 440 may be "INCONEL 600" for resisting corrosion. Second coiled spring 400 is interposed between distal end portion 220 of plug body 210 and proximal end portion 430 of cam 400 to exert a downwardly biasing force for biasing proximal end portion 430 into engagement with cam surface 380. In the preferred embodiment, the compressive force required to fully compress second coiled spring 400 may be approximately 30 to 50 pounds for reasons provided hereinbelow. Connected to distal end portion 420 of cam 400, such as by an internally threaded nut 450 threadably engaging the external threads of distal end portion 420, is a generally cylindrical and hollow piston 460 for sliding or driving cam 400 into engagement with catches 360 in the manner provided hereinbelow. In this regard, piston 460 is capable of slidably translating cam 400 in bore 310 as cam 400 is driven into engagement with catches 360. Piston 460 may be "INCONEL 600" for resisting corrosion. More specifically, piston 460 defines an inner wall 470 therein and an integrally attached floor 475 at an end thereof, floor 475 having a plurality of orifices 480 transversely therethrough for reasons disclosed hereinbelow. Integrally attached to inner wall 470 and extending perpendicularly inwardly with respect thereto is a generally annular ring member 485 having a notch 490 for reasons disclosed more fully hereinbelow. Referring yet again to FIGS. 2, 3, 4 and 5, connected to plug body 210 is ramming means, such as a ram 500, for ramming plug body 210 into hole 150. Ram 500 may be "INCONEL 600" for resisting corrosion. In the preferred embodiment, ram 500 is a generally cylindrical slide 510 having an inner wall 520 for slidably surrounding piston 460. Slide 510 also has a distal end portion 515 and a proximal end portion 517. Slide 510 may include a flange 530 integrally formed therewith and extending circumferentially therearound. It will be appreciated that flange 530 abuts top surface 147, which belongs to core barrel flange 140, when hole 150 is not counter-bored. Thus, one purpose for flange 530 is to prevent plug 200 from passing through hole 150 when hole 150 is a straight bore rather than a counter-bore. Another purpose of flange 530 is to allow plug 200 to be readily suspended in a bore (not shown) formed in a storage rack (not shown) for storing plug 200 when not in use. Slide 510 also includes a plurality of perpendicularly inwardly extending pegs 540 integrally attached to proximal end portion 517 for slidably engaging first groove 240. Thus, it will be appreciated from the description hereinabove that slide 510 is capable of downwardly slidably advancing and upwardly slidably retreating on plug body 210 because slide 510 slidably surrounds piston 470. It will also be appreciated from the description hereinabove that upper stop 250 is capable of stopping the upward retreat of pegs 540 for stopping the upward travel of slide 510 and that lower stop 260 is capable of stopping the downward travel of pegs 540 for stopping the downward travel of slide 510 as pegs 540 slidably engage first groove 240. Interposed between proximal end portion 280 and shelf 290 and surrounding second groove 270 is third biasing means, such as a third coiled spring 550, for biasing plug body 210 snugly into engagement with land 155. Third coiled spring 550 may be "INCONEL 600" for resisting corrosion. Turning now to FIGS. 6 and 7, plug 200 is shown in its unlocked position for insertion into hole 150 and for removal from hole 150. In this regard, a plug installation and removal tool 560 is provided for inserting and locking plug 200 in hole 150 and for unlocking and removing plug 200 from hole 150. Removal tool 560 may include an elongate shaft 570 having a plurality of perpendicularly outwardly extending and integrally attached wings 580 at one end thereof, wings 580 being sized to pass through notch 490. Integrally attached to the other end of shaft 570 is a handle 590 for rotating shaft 570 about its longitudinal axis. Wings 462 are capable of passing through notch 490, rotated by means of shaft 570. Wings 462 are then caused to abut the bottom surface of ring member 462 and then upwardly translated (by means of handle 560 and shaft 570) to upwardly translate piston 460 and cam 400 so that plug 200 is unlocked from flange 140 in the manner disclosed hereinbelow. Moreover, shaft 570 is capable of bearing against distal end portion 420 of cam 400 and its associated nut 450 to downwardly translate piston 460 and cam 400 for locking plug 200 to flange 140 in the manner disclosed hereinbelow. In this regard, shaft 570 is downwardly translated to bear against distal end portion 420 and nut 450 by downwardly pushing on handle 560. OPERATION Before plug 200 is installed in hole 150, hold-down bolts 60 and CRDMs 90 are removed in a customary manner well known in the art. Closure head 50 is then dismounted from vessel shell 20. Next, upper support structure 160 is dismounted from atop core barrel 100. Plug 200 is then remotely coaxially aligned with hole 150 and lowered therein. Plug 200 is coaxially aligned with hole 150 and lowered therein preferably by means of a suitable robotic device (not shown) rather than manually due to the relatively high radiation field emanating from reactor core 70. When being inserted into hole 150, plug 200 is in its unlocked configuration as shown in FIGS. 6 and 7. Tool 560 is then coaxially aligned with piston 460 and downwardly translated, by means of handle 590, such that wings 580 pass unimpeded through notch 490. Tool 560 is further downwardly translated until the end of shaft 570 engages nut 450 and distal end portion 420 of cam 400. As shaft 570 engages nut 450 and distal end portion 420 to exert a downwardly-directed force thereon, piston 460 downwardly translates to downwardly slidably translate cam 400 in bore 310 until contoured proximal end portion 430 of cam 400 engages cam surface 380 belonging to each catch 360. The downward travel of cam 400 is limited or stopped by center stop 435. At the time when center stop 435 stops the downward travel of cam 400, proximal end portion 430 of cam 400 will have fully engaged cam surface 380. The force exerted by proximal end portion 430 against cam surface 380 causes at least a portion of each catch 360 to outwardly pivot about pivot pin 350 in a direction generally toward flange 140. As each catch 360 outwardly pivots, hook portion 370, which belongs to catch 360, is aligned with or disposed opposite undersurface 145 of flange 140. Moreover, once the contoured proximal end portion 430 of cam 400 suitably engages cam surface 380 in the manner described hereinabove, proximal end portion 430 will tend to resist disengagement from cam surface 380 due to the V-notch shape or contour of cam surface 380. In addition, it will be appreciated from the description hereinabove that the relatively shallow angle (approximately ten degrees) at the interface between proximal end portion 430 of cam 400 and inclined portion 385 of cam surface 380 assures that when catch 360 engages cam 400, the V-notch shape of cam surface 380 assists in sliding proximal end portion 430 downwardly as catch 360 pivots outwardly through slot 320. In addition, this feature of the invention assures that proximal end portion 430 suitably seats on inclined portion 385 of cam surface 380 to resist disengagement. Moreover, second coiled spring 440, which surrounds cam 400 and which is interposed between plug body 210 and cam 400, exerts a downwardly axially biasing force on cam 400 for providing additional assurance that proximal end portion 430 firmly seats on inclined portion 385 of surface 380 so that catches 360 remain in an outwardly locked position. In the preferred embodiment, this downwardly axially biasing force exerted by spring 440 on cam 400 may be approximately 200 pounds in order to firmly seat proximal end portion 430 on inclined portion 385 of surface 380. Moreover, first coiled spring 390, which is interposed between catches 360, exerts an outwardly radially biasing force against catches 360 for providing assurance that at least the hooked portion 370 of each catch 360 remains in an outwardly locked position until removal of plug 200 is desired. In this manner, plug 200 is locked to flange 140. Thus, it will be understood from the description hereinabove that plug 200 is self-locking. That is, once locked to flange 140, plug 200 remains in its locked position until removal is desired, as described in detail hereinbelow. After plug 200 is locked to flange 140, in the manner described hereinabove, tool 560 is removed from piston 460 in substantially the reverse order of its insertion into piston 460. Next, upper support structure 160 is replaced atop core barrel 100 and hold-down spring 190 such that bottom surface 165 belonging to upper support structure 160 abuts hole-down spring 190 and distal end portion 515 of slide 510. As bottom surface 165 abuts distal end portion 517 of slide 510, slide 510 translates downwardly by a predetermined amount such that proximal end portion 515 of slide 510 compresses third coiled spring 550, which is disposed in second groove 270. In the preferred embodiment, the force required to fully compress third coiled spring 550 is approximately 30 to 50 pounds. As slide 510 downwardly translates, pegs 540 will downwardly translate to a like extent and slide in first groove 240 because pegs 540 are attached to slide 510. The downward translation of pegs 540, and thus the downward translation of slide 510, are stopped by lower stop 260. The force downwardly acting on slide 510 is transferred through third coiled spring 550 to shelf 290 because third coiled sprig 500 rests on shelf 290. As this force bears on shelf 290, it is transferred to plug body 210 for ramming plug body 210 snugly into hole 150 and into intimate engagement with land 155. Ramming plug body 210 into hole 150 in this manner provides assurance that plug 200 firmly seats on land 155. Moreover, as upper support structure 160 is mounted atop core barrel 100, it will exert a downwardly directed force of approximately 100,000 to 200,000 pounds against distal end portion 515 of slide 510 for maintaining plug 200 in hole 150 even as it is acted upon by upwardly directed hydraulic forces. Closure head 50, bolts 60 and CRDMs 90 are then replaced atop vessel shell 20 in substantially the reverse order of their removal from atop vessel shell 20. As stated hereinabove, it is periodically necessary to retrieve radiation detector 120 from annular space 110 to measure the integrated neutron fluence seen by vessel shell 20. However, in order to access detector 120, hole 150 must be unplugged by removing plug 200. To remove plug 200 the CRDMs 90, bolts 60, closure head 50 and upper support structure 160 are removed in the manner previously described. Removal of upper support structure 160 allows slide 510 to upwardly slidably translate due to the upwardly biasing force exerted on proximal end portion 517 by third coiled spring 550. As slide 510 upwardly slidably translates, pegs 540 will also upwardly slidably translate in first groove 240 because pegs 540 are attached to slide 510 and slidably engage first groove 240. However, the upward travel of pegs 540 and thus the upward travel of slide 510 are stopped by upper stop 250. Next, tool 560 is coaxially aligned with piston 460 and downwardly translated, by means of handle 590, such that wings 580 pass unimpeded through notch 490. After wings 580 pass through notch 490, handle 590 is rotated approximately 90 degrees and upwardly translated such that wings 580 abut or engage the underside of ring member 462. Handle 590 is then further upwardly translated such that wings 580 exert a predetermined upwardly directed force (approximately 200 pounds force) against the underside of ring 462 for upwardly translating piston 460. As piston upwardly translates, proximal end portion 430 of cam 400 will upwardly translate such that proximal end portion 430 disengages cam surface 380. As proximal end portion 430 disengages cam surface 380 it will cause catches 360 to inwardly radially retract into slot 320 as they inwardly pivot about pivot pins 350. Moreover, it will be appreciated from the description hereinabove that as cam 400 is translated upwardly, coiled springs 390/440 are compressed; thus, the upward force exerted on cam 400 by tool 560 should be sufficient to overcome the downwardly biasing forces of coiled springs 390/440 in order to move catches 360 to their unlocked position (as best seen in FIG. 7). Tool 560 is then further upwardly translated to remove plug 200 from hole 150 for allowing access to detector 120. It will be appreciated that the "as-built" plug 200 may not be perfectly leak-tight due to tolerances and deviations from nominal design dimensions occurring during its manufacture. Thus, during reactor coolant pump transients or surges, a relatively small amount of the coolant may seep into plug 200 because plug 200 may not be leak-tight. Hence, the coolant may enter plug 200 due to the relatively high pressure and velocity of the coolant in vessel shell 20 during such transients. It is therefore important to allow the coolant to escape through plug 200 in order to avoid build-up of internal pressures in plug 200. Build-up of pressure inside plug 200 (i.e., pressurization of plug 200) is undesirable because such pressure build-up may ultimately damage the internal components of plug 200 and compromise the ability of plug 200 to be locked to flange 140. Therefore, a fluid path is provided through plug 200 to allow the coolant to escape through plug 200 and into the space above top surface 147 of flange 140. In this regard, the coolant is allowed to pass through plug 200 along a flow path extending through slot 320, through passageway 410, through bore 310, through channel 330, into first groove 240, between slide 510 and distal end portion 220 of plug body 210, through orifices 480 and then to exit plug 200 through annular ring member 485. Moreover, it should be appreciated from the description hereinabove that the various elements of plug 200 are preferably fabricated from corrosion resistant material, such as "INCONEL 600", "TYPE 304" stainless steel, or the like, so that the elements of plug 200 are not fouled by scaling caused by corrosion. This is important because corrosion of the elements comprising plug 200 may compromise the locking ability of plug 200 by retarding the sliding, pivoting, and translational movements of the elements. It will be evident from the teachings herein that an advantage of the present invention is that plug 200 resists being inadvertently dislodged from hole 150 due to upset hydraulic forces because plug 200 is capable of being locked to the structure defining hole 150. It will also be evident from the teachings herein that another advantage of the present invention is that the reactor coolant does not substantially flow through the hole to bypass core 70 when plug 200 plugs hole 150. Although the invention is fully illustrated and described herein, it is not intended that the invention as illustrated and described be limited to the details shown, because various modifications may be obtained with respect to the invention without departing from the spirit of the invention or the scope of equivalents thereof. For example, elastomeric seals may be beneficially located in plug body 210 to preclude coolant from entering plug body 210. This would eliminate the need for providing a fluid flow path through plug body 210. Therefore, what is provided is a self-locking plug for plugging a hole defined by a surrounding structure, which structure may be a nuclear power reactor pressure vessel core barrel flange. |
abstract | The present invention provides a charged particle beam apparatus used to measure micro-dimensions (CD value) of a semiconductor apparatus or the like which captures images for measurement. For the present invention, a sample for calibration, on which a plurality of polyhedral structural objects with known angles on surfaces produced by the crystal anisotropic etching technology are arranged in a viewing field, is used. A beam landing angle at each position within a viewing field is calculated based on geometric deformation on an image of each polyhedral structural object. Beam control parameters for equalizing the beam landing angle at each position within the viewing field are pre-registered. The registered beam control parameters are applied according to the position of the pattern to be measured within the viewing field when performing dimensional measurement. Accordingly, the present invention provides methods for reducing the variation in the CD value caused by the variation in the electron beam landing angle with respect to the sample with an equal beam landing angle and methods for reducing the instrumental error caused by the difference in the electron beam landing angle between apparatuses. |
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047120150 | claims | 1. A roof shield for a nuclear reactor comprising a normally fixed radially outer portion having a radially inner wall, a radially inner portion rotatable about a vertical axis, and a connection between the inner and outer portions which permits relative angular movement between the portions without loss of containment, the connection being on the radially inner side of an upper portion of the radially inner wall of the outer portion, the upper portion of the inner wall being connected to the outer portion only at the lower end of the upper portion and the upper portion being so arranged that on upward movement of the inner portion the upper portion of the inner wall receives substantially no angular movement, the lower end of the upper portion being at a position below the connection and below the level of the upper surface of the outer portion. 2. A roof shield as claimed in claim 1, in which the outer portion has top and bottom walls, said radially inner wall connected to the bottom wall such that on said upward movement the top wall moves away from said upper portion which bends in the region of said position. 3. A roof shield as claimed in claim 1, in which the connection comprises keys. 4. A roof shield as claimed in claim 1 wherein the connection is below the upper surfaces of the inner and outer portions. 5. A roof shield for a nuclear reactor comprising a normally fixed radially outer portion having a radially inner wall, a radially inner portion rotatable about a vertical axis, a connection between the inner and outer portions, the connection being on the radially inner side of an upper portion of the radially inner wall, the upper portion of the inner wall being operable as a cantilever from a position below the connection and below the level of an upper surface of the outer portion, such that on upward movement of the inner portion the radially inner wall bends in the region of said position and the connection is maintained without loss of containment. 6. A roof shield as claimed in claim 5, in which the connection comprises keys at a level which is below the upper surface of the outer portion and above said position. |
054250714 | description | Referring to FIG. 1, cylindrical fuel pellets 1 which are made of a mixed oxide fuel comprising uranium dioxide and plutonium dioxide are formed into rows within a containment area 2 which is bounded by radiation shielding walls 3. The fuel pellets are subsequently loaded into a fuel pin cladding tube 4 located outside the containment area to form a nuclear fuel pin. Within the containment area 2 the pellets 1 are arranged in end-to-end relationship in a trough provided in a supporting tray 5. The tray 5 has a number of parallel troughs in which several rows of fuel pellets are formed. From the tray 5 a row of pellets 1 is pushed onto a weighing channel 6 by fingers 7 associated with a transfer mechanism 8. A feed stop 9 restrains the leading pellet until a stack containing the required number of pellets is formed. When the stack is of the required length, as determined by weighing equipment associated with the weighing channel, it is forwarded to the vibratory table 10. This may be achieved by vibrating the weighing channel 6. Outside the containment area 2 a cladding tube 4, fitted with a bottom end plug 11, is positioned in alignment with te pellet stack. Pressed on the end of the cladding tube is a disposable sleeve 12, preferably made of a plastics material, which is slidably located in a resilient sphincter seal 13. The sphincter seal 13 is attached to a mounting plate 14 which is resiliently connected to the containment shielding wall 3 by a diaphragm 15. The sleeve 12 extends through a central opening 16 in the mounting plate with an end surface 17 located against a retractable stop plate 18. The sphincter seal is provided with several rubber rings 19 each of which is divided into sectors. As the sleeve 12 is passed through the centre of the seal the rings 19 are deformed but maintain positive sealing contact by pressing against the sleeve. Clamps 20 are operated to lightly hold the cladding tube 4 in position and an end stop 21 is moved over the bottom end plug 11 to prevent longitudinal movement of the tube. In operation, the vibratory table 10 and the cladding tube 4 are subjected to vibration by connection to a bi-modal vibrator (not shown). This causes the stack of pellets to migrate along a guide tube 22 integrally formed with the vibratory table 10. The end of the tube 22 is in the form of a semi-circular passage 23 which guides the pelets through the sleeve 12 and introduces them into the cladding tube 4. Advantageously, the guide tube 22 is made of a material which is compatible with the fuel pellets, suitable examples being stainless steel or a zirconium alloy. The pellets migrate along the cladding tube 4 until the leading pellet encounters the bottom end plug 11. The arrangement described for inserting the fuel pellets into the cladding tube is by way of example only and other systems incorporating, for example, soft handling surfaces or pelelt inserting rams, can also be used. After retractiing and parking the vibratory table 10 and guide tube 22, a disposable plug carrier 24 (see FIG. 2c) is used to insert a top end plug 25 and plenum spring 26 into the end of the cladding tube 4. Preferably the plug carrier 24 is made of a plastics material. Reaction plates 27 are operated to bring them behind the sleeve 12, the clamps 20 are released and the end stop 21 is removed. The cladding tube 4 is withdrawn while the sleeve 12 and plug carrier 24 are restrained by the reaction plates so that they remain trapped in the sphincter seal 13. The withdrawn cladding tube 4 passes through a girth welder 28 which makes a circumferential weld to join the top end plug 25 to the cladding tube. Further withdrawal of the cladding tube brings the top end plug 25 into a helium filling and welding device 29 which injects helium into the tube through a fine hole 30 (see FIG. 2c) in the plug and then fills the hole with weld material. The pellet loading sequence will now be described with particular reference to FIGS. 2a, 2b, 2c, 2d and 2e. FIG. 2a shows a cladding tube 4 fitted with a bottom end plug 11 at one end and with a disposable sleeve 12 mounted on the other end. At one end the sleeve has a reduced diameter seating 31 which seats as a press fit on the end of the cladding tube 4. An internal end face 32 of the seating 31 coincides with an end surface 33 of the cladding tube. Thus the external end surface of the cladding tube 4 is covered by the seating 31 and is thereby protected from radioactive contamination by the environment in the containment area 2. As the cladding tube 4 moves in the direction of arrow A the sleeve 12 engages a sleeve 12a, enclosing a plug carrier 24a, which have been retained in the sphincter seal 13 following the preceding pellet loading procedure. With the stop plate 18 in a retracted position, the new sleeve 12 pushes the sleeve and plug carrier assembly out of the sphincter seal 13 into the containment area 2 (FIG. 2b). Since the new sleeve 12 enters the sphincter seal 13 before the previously used sleeve and plug carrier assembly is removed from the seal, leakage of radioactive substances from the containment area is prevented. The stop plate 18 is then lowered to provide a location for end surface 17 of te sleeve 12. FIG. 2c shows the semi-circular end passage 23 of guide tube 22 extending into the sleeve 12. Upon vibration of the vibratory table 10 ad cladding tube 4, as previously described, the pellets 1 migrate along the semi-circular passage 23 and then down the cladding tube until the leading pellet encounters the bottom end plug 11. When the stack of pellets has been loaded into the cladding tube 4 the table 10 and guide tube 22 are retracted. The top end plug 25 together with a plenum spring 26 are then pressed into the end of the cladding tube 4 using the disposable plug carrier 24. The plug carrier is of cylindrical shape and ahs an outer diameter such that it is a sliding fit within the sleeve 12. A blind hole 35, drilled along the longitudinal axis of the plug carrier 24, has an increased diameter recess 36 at its forward end. The depth of the recess is sufficient to completely receive a head portion 37 of the top plug 25. It is particularly important to ensure that the end surface 38 and peripheral surface 39 of the head portion 37 are completely covered so as to protect them from possible contamination. At the leading end of the top plug 25 is a shoulder 40 which provides a seating for the plenum spring 26. After retracting and parking the vibratory table 10 and guide tube 22 a remotely controlled manipulator, not shown, provided in the containment area, may be used to push the plug carrier 24 through the sleeve 12 and to press fit the top plug 25 into the end of the cladding tube 4. During installation of the top plug the end stop 21 prevents axial movement of the cladding tube. The plenum spring 26 serves to restrain the fuel pellets during subsequent handling and transportation of the fuel pin so that they do not become damaged. As seen in FIG. 2d, the reaction plates 27 are moved inwardly to locate behind the sleeve 12 while clamps 20 are released and the end stop 21 is retracted clear of the bottom plug 11. As the fuel pin is withdrawn, the sleeve 12 and plug carrier 24 remain located in the sphincter seal 13 (FIG. 2e). The fuel pin then passes through the girth welder 28 for end plug welding and the helium filling and welding device 29, as previously described, to complete the fuel pin assembly. The sequence of operations is then repeated for the next fuel pin. It will be seen that during the pellet loading operation none of the external surfaces of the cladding tube or top end plug is exposed to the radioactive environment existing in the containment area. This eliminates the need to subject the fuel pin to a costly decontamination process. |
summary | ||
044118627 | description | Referring now to the drawings, FIG. 1 shows a portion of a grating for fuel assemblies of a light water nuclear reactor near a corner of this grating. The whole grating comprises seventeen cells along each of the sides of the grating. The grating is limited on its periphery by walls such as 1 and 2 constituted by a plate of long length bearing openings for constituting stiffeners. The outer walls of the grating also bear rigid stops such as 4 and 5 at the level of each of the cells on one wall, as far as the cells on the sides are concerned and on two walls, as far as the corner cells 6 are concerned. The grating as shown in FIG. 1 comprises cells 7 for the fuel elements such as 8 and cells 10 for the guide tubes. The greater part of the cells 7 in which pass the fuel cells 8 have two walls equipped with double springs 12 disposed successively on the perimeter of the cell, the two walls opposite these walls equipped with double springs 12 comprising double rigid stops 13 and 14, the rigid stops 13a and 13b and the rigid stops 14a and 14b respectively located on one and the other wall, being directly inwardly of the cell and inwardly of the adjacent cell disposed on the other side of the wall in question. The plates of the cells 10 reserved for the passage of the guide tubes do not comprise any spring on their wall, but solely rigid stops 15 directed towards the adjacent cells and connection means towards the inside of the cell where the guide tube passes for the rigid fixing of this guide tube inside the cell. In a 17.times.17 assembly, i.e. an assembly comprising 17 cells on each of the sides of the grating, 24 guide tubes are disposed which obviously necessitate 24 cells such as cells 10. This presence of cells 10 of which the walls do not comprise a spring introduces a disturbance in the lattice of the elements holding the fuel pencils, so that it is necessary to dispose, at certain spots, single springs 16 comprising one active face only in one of the cells, the adjacent cell comprising on the wall where the single spring is located, a rigid stop 17 made on the wall in question independently of the spring 16. In the same way, single springs 16 are necessary to take into account the disturbances in the distribution of the transverse holding elements caused by the fact that the number of cells on the sides of the grating is an odd number and the outer walls of this grating bear only rigid stops such as 4 and 5. Taking into account the inevitable dissymmetries of the whole of the grating due to the presence of the guide tubes, it is therefore seen that the cells of the grating are made by associating four different wall types, namely: the walls equipped with springs on their two faces; the walls equipped with rigid stops on their two faces; the mixed walls equipped on one side with springs and on the other side with rigid stops; the walls of the cells for guide tubes, equipped with rigid stops on the face not in contact with the guide tube. The grating shown in FIG. 1 is constituted by plates 19 and 20 intersecting at right angles, such a plate being shown in FIG. 2. In FIG. 2, the plate has been shown after cutting out and before bending, said bending itself preceding the assembly of the grating. In FIG. 3, the same plate has been shown in plan view after bending and pressing preceding the operation for assembly of the grating. The plate shown in FIGS. 2 and 3 is constituted by a strip cut out in a particular form and presenting a series of slits 25 spaced apart by a distance equal to the sides of the cells of the grating. The grating is assembled by lap-jointing the plates, of the same width as the plate shown in FIG. 2, introduced reciprocally in the slits 25 so as to define the cells of the grating. These intersecting plates are connected together after assembly by a weld 26-2 made at each lower and upper corner of the defined cell. When the plates are being intersected, edges of the slits 25 are guided and positioned through embossed portions 26 obtained by cutting out and pressing the plate shown in FIG. 2. The embossed portions 26 ensure in a first function the guiding of the plates when they are assembled. In order to facilitate this guiding, the embossed portions 26 have a bevelled section obtained directly when the plate is cut out. FIGS. 2a and 2b show an embodiment of this guiding obtained by a prior cut-out of the plate in the form of a key-hole (26-1) which, after the embossed portions have been formed, has a bevelled section in the direction of introduction of the plates. The second function of the embossed portions is to ensure a distribution of the transverse forces transmitted to the rigid stops (14a and 14b). FIG. 2c shows how the torque induced by the forces F of two pencils located in two adjacent cells is taken up by the reactions of abutment f.sub.1 and f.sub.2 in the embossed portions 26 and by the reaction of abutment f.sub.3 at the level of the weld spots 26-2. The preferred arrangement which consists in implanting the embossed portions in the vicinity of and on each side of the support elements 14a and 14b enables the forces of type f.sub.1, f.sub.2 and f.sub.3 to be distributed. The plate 24 comprises in particular openings 27 and 28 for the introduction of springs constituted in the form of pins which will be described in greater detail hereinafter. The plate 24 also comprises openings 30 between which the metal is pressed and bent along edges 31 to form pressed rigid stops 14 such as those which are also visible in FIG. 1. These rigid stops 14 disposed on either side of the wall are provided in the upper part and in the lower part of the plate 24 to form rigid stops for the fuel pencil located in the corresponding cell and coming into contact with this pencil at points spaced apart by a relatively long length with respect to the height of the plate. The plate 24 also comprises openings 32 for forming rigid stops such as 14, 15, by pressing. The plate shown in FIGS. 2 and 3 corresponds to the plate 19 of the grating shown in FIG. 1; it has been shown solely to illustrate the different types of openings and rigid stops made in the plates constituting the grating. FIG. 4 shows a first embodiment of a double spring, added to a plate 24 such as the plate shown in FIG. 2 and surrounding this plate 24 at openings such as 27. This double spring 40 is constituted in the form of a pin having two arms 40a and 40b constituted from a leaf bent on itself and welded in the vicinity of the two ends of the pin in zones 41 and 42. The top part of the pin constitutes a loop 43. The arms 40a and 40b of the pin are symmetrical with respect to the plate 24 and comprise, in their active parts, two flattened support surfaces 44a and 45a, 44b and 45b projecting inwardly of the adjacent cells to come into contact with the pencils 8. Between these two support surfaces 44a and 45a or 44b and 45b, the rigidity of the metal leaf constituting the spring is reinforced by forming the edges 46a and 46b of this leaf which, in this central part of the arms of the pin, has a C-section. The support surfaces 44 and 45 are spaced apart by a distance in the vertical direction substantially equal to the distance between the rigid stops 14 made on the plates 24. A double spring such as shown in FIG. 4 enables balanced forces to be applied, due to the rigid central part of the pin forming the spring, at two points of a fuel pencil 8 disposed in one of the cells limited by the plate 24. When transverse forces are applied by one of the arms of the spring on a fuel pencil, these forces are taken up directly in the other arm of the spring pin since the latter is not fixed on the plate 24 but simply surrounds this plate without tightening the pin on the plate. This second arm itself applying transverse forces on the fuel pencil of the adjacent cell which it serves, these forces are taken up directly in the first arm, thus defining a symmetry of the forces exerted on the two fuel pencils. In this way, the transverse forces exerted on the pencils are therefore virtually not applied on the plates of the grating, a balancing of these forces being possible at the level of each pin which thus works by essentially axial deformation with respect to the wall, the pin being able to slide along said wall. FIG. 5 shows a double spring having a form different from the spring pin shown in FIG. 4, but performing the same functions. This spring pin 50 also comprises two arms 50a and 50b of which the upper and lower rounded parts 55 and 57 respectively project and come into contact with the pencils 8 to exert a transverse force on these pencils 8. The top and bottom rounded parts of each of the arms of the pin are made rigid by shaping the edges of the metal leaf constituting the two arms of the pin to obtain a C-section (56) at the level of these rounded parts, as shown in FIG. 6. On the other hand, the central part 54 of the arms 50 is not rigidified and constitutes a zone of articulation between the two rigid parts of each of the two arms of the pin. It is the deformation of this central part of the two arms of the pin which comes into light contact with the plate 24 of the arms of the pin and enables the balance of the forces exerted by the pencils 8 to be taken up, with the zones 52 and 51. As in the spring pin shown in FIG. 4, the upper zone 51 and lower zone 52 are welded one to the other to connect the two arms of the pin around the plate 24, whilst the top part 53 is in the form of a loop. As in the case envisaged hereinabove, the forces exerted on the pencils 8 in the cells on either side of the plate 24 are only weakly supported by this plate 24 and are mostly transmitted from one arm of the pin to the other and consequently are balanced, bringing about a substantially axial deformation. FIGS. 7 and 8 show single springs comprising one active part 60 only, the pin closing on the other face of the plate 24 by a straight arm 61 which does not come into contact with the fuel pencil disposed in the corresponding cell of the grating. The central zone 62 of the active arm of the pin is rigidified by forming the metal leaf and support surfaces 64 and 65 are disposed on either side of this rigidified part of the arm of the pin. The spring shown in FIG. 8 comprises, on the contrary, a flexible central zone 70 and two upper and lower rounded zones 71 and 72 coming into contact with the pencil and rigidified by shaping the leaf into a C. In the case of the springs shown in FIGS. 7 and 8, the deformation of the active arm is adapted by elongation of the opposite arm 61. The elongation of the active part of the spring under the effect of the transverse forces and that of the opposite arm 61 are compensated in the case of the springs shown in FIGS. 7 and 8 by two deformation zones 68. In the case of the springs shown in FIGS. 7 and 8, openings 28a made in the plate 24 enable the bent parts 68 of the arm 61 to be introduced into these openings. As in the case of the double springs, the end zones 69 of the spring pins are connected by welding and the upper part of the pin is constituted by a loop 73. In the case of the single springs and the double springs, the spring pin is mounted with clearance on the plate 24 at an opening 27 allowing the spring pin to be held in position on the plate after the ends have been welded, whilst providing a considerable initial axial clearance. This initial clearance makes it very simple to position the springs and allows a broad tolerance on the distance between the zones of weld of the spring at its end and the openings 27 of the plate. The arrangement of the openings in the plates and the shape of the pins allow easy assembly and possibly dismantling of these pins when the plates forming the grating are assembled. This constitutes an advantage for the possible pin replacing operations during manufacture. The springs are designed to work elastically with elongation of the pin; the ends of the springs are therefore mounted to slide freely with respect to the plate which supports them. However, in those cases where considerable transverse forces are exerted on the preceding springs, the transfer of these forces on the plates of the cell would be limited to the central zone only of these springs, due to the arrangements of the invention. FIGS. 9 and 10 respectively show a double spring and a single spring which are mounted on a plate 24 in the same manner as the springs which have been described with reference to FIGS. 4 to 8, but which comprise only one support zone 75 for each of their active parts projecting in a cell containing a fuel pencil. The rigidity of this support zone is ensured by folding the arms 76 of the leaf to obtain a C-section, as may be seen in FIG. 11. This amplified form of spring does not make it possible to obtain a semi-rigid fit which is as efficient as when two support zones of each active part of the springs are disposed at the level of two rigid stops borne by the faces of the cell opposite the faces bearing the springs. However, this type of spring may be sufficient if a very strong moment of semi-rigid fit is not sought for the fuel pencil in the cell of the grating. In the case of a grating of a 17.times.17 assembly, i.e. an assembly disposed in a square lattice comprising 17 rows of 17 cells of which 24 are occupied by guide tubes, one by a tube comprising automatic control instruments, and the others by fuel pencils, 244 double springs and 40 single springs of the type shown in FIGS. 4 and 6 are used in association with rigid stops disposed in two's on the faces of the cells to effect transverse holding of the fuel pencils. 284 spring pins have therefore been used for a total of 528 springs, which clearly shows the simplified assembly with respect to a device preferably using single springs. With reference to FIGS. 12 to 15, other possible embodiments of the supports made in walls of cells of gratings for fuel elements will now be described. In all these embodiments, the essential and original feature lies in the fact that all the springs and rigid stops are added and made separately from one another of a nickel alloy before being assembled by welding on the walls of the cells which bear openings provided to this end. This system allows the limitation of the axial elongation of the springs when they slide along the plates of the cell, and the taking up of the forces on the parts forming spring whether the wall comprises a double spring or a single spring cooperating with a rigid stop on the other face. In the embodiments which will now be described, the walls of cells comprise solely openings which are made by machine without stamping and consequently allow a more resistant and less ductile material to be used. The support pieces are added to the plate constituting the wall and, as they are generally made of material which considerably absorbs the neutrons, the quantity of material is strictly limited to that necessary for making the different springs and rigid stops, as will appear from the following description. FIG. 12 shows, at 80, the wall of a cell equipped on either side with two double springs 81a, 81b and 82a and 82b. FIG. 12a is a section along A--A of FIG. 12b, which is a plan view of FIG. 12a. The plate 80 comprises three rectangular openings 83, 84 and 85 located respectively near the upper and lower edges and at the centre of this plate. In each of these openings 83, 84 and 85 is located a contact shoulder, namely 86, 87 and 88 respectively, via which the two preceding springs are in contact, welded to one another, and tighten on the plate 80. To this end, a certain number of weld spots such as 89 are provided on each shoulder. The central sliding shoulder 87 comprises a longitudinal clearance 90 in its housing 84, which makes it possible thus to effect a sliding fix, without longitudinal contact for the central shoulder 87. This clearance is very important, as it allows the balancing of the reactions of the different springs on the fuel pencils (not shown) by longitudinal displacement of all the pieces 81 and 82 added to the plate 80. At the ends of the springs 81 and 82, the stop is established by compression in the plane of the plate 80 on the section of the housings of shoulder 86 and 88 in the plate itself. Finally, FIG. 12c which is a section through FIG. 12a along B--B at the top end shows how the edges 91 and 92 of the shoulder 88 make it possible, by tightening on the plate 80 in the edges 91 and 92, to avoid the shoulder 88 escaping through the cut-out orifice 85 in the plate 80. The same applies to the shoulder 86 located in the lower part. As may be verified in the different FIGS. 12, the only forces which the plate 80 may undergo in operation are contained in its plane to the exclusion of any bending moment which might cause inadmissible deformation thereof. In fact, the forces exerted on either side on the springs 81 and 82 are balanced and are transferred to the shoulders 86, 87 and 88 where they are compensated two by two and face to face. The different FIGS. 13 show in section, elevation and plan view, the embodiment of a wall comprising two springs on one face and three rigid stops on the opposite face. The plate 80 comprises in this example five openings referenced 95, 96, 97, 98 and 99 respectively. The added pieces comprise, on one side, the springs 100 and 101 and, on the other side, the rigid stops 102, 103 and 104. The mode of implanting the preceding added pieces on the plate 80 is of the same nature as that of the embodiment of FIG. 12 concerning the ends and the centre. The central join constituted by a rigid stop 103 and by the zone of connection 105 between the two springs 100 and 101 is mounted to slide with two longitudinal clearances 90 in the opening 97 of the plate 80 so as to form a central sliding shoulder. As in the preceding embodiment of FIG. 12, FIG. 13c which is a section of FIG. 13a along B--B, shows the edges 91 and 92 of the end shoulder making it possible to avoid the passage of said shoulder through the corresponding orifice 95 or 99. In this embodiment, the end of the rigid end stops 102 and 104 opposite the orifices 95 and 99 is fitted in the orifices 96 and 98 by crimping in the edges of the plate 80. It is also very advantageous to make the three rigid stops 102, 103 and 104 by giving them an ovoid form, which largely facilitates the flow of the water and avoids the highly undesirable phenomena of cavitation during operation. Rigid stops formed in this way may easily be obtained by cold stamping. As in the example of FIG. 12, it is readily seen that the balancing of the reactions occurring on either side of the wall 80 is effected face to face at the welds without introducing any bending torque on the plate 80, which is affected only at its ends by forces contained in its plane. The embodiment of a plate of a cell 80 equipped on either side with two rigid stops of ovoid form, will now be described with reference to FIGS. 14a, 14b and 14c. In this embodiment, the plate 80 comprises four rectangular openings 110, 111 at one of its ends and 112, 113 at the other end, in which are welded the shoulders 114, 115, 116 and 117 of the two plates in which the ovoid rigid stops 118, 119, 120 and 121 are formed on either side. The four shoulders 114, 115, 116 and 117 are held in position on the wall 80 by the edges 91 and 92 comparable with those of FIGS. 12c and 13c. This is shown very clearly in FIG. 14c which is a section along B--B of FIG. 14a at the level of one of the shoulders 114, 115, 116 or 117. Clearances 122 and 123 on the one hand and 124 and 125 on the other hand are provided in the longitudinal direction between the shoulders 114, 115 on the one hand and 116, 117 on the other hand and the corresponding openings 110, 111, 112 and 113, so as to allow the balancing of the forces exerted transversely on the corresponding rigid stops. In the embodiment of FIGS. 14, it is seen that the pieces constituting the different stops 118, 119, 120 and 121 and their weld shoulder at the ends are strictly limited and do not cover the plate 80 entirely, in order to limit as much as possible the introduction into the grating of a material which considerably absorbs the neutrons. Furthermore, the same applies to the balancing of the pressures on either side and the absence of any bending torque and any embossed portion in the plate as far as the advantages of this embodiment are concerned. Referring now to FIG. 15, the manner will finally be described in which the plate 80 of a cell 130 is equipped, in which a guide tube 131 slides. Such guide tubes are inserted with easy fit in a sleeve 132 comprising a window 133 at the level of its contacts with each of the walls such as 80. A rigid stop 134 is fixed on the face of the wall 80 opposite the guide tube 131 with the aid of a double crimping 135 through slots 136 and 137 made in the wall 80. In the case of this preferred application of the invention, each of the fuel pencils is therefore held inside each of the cells by two springs each comprising two support faces and by two sets of rigid stops disposed substantially opposite the support surfaces of the springs. The pencil is therefore in contact with the holding elements at at least eight different zones for each of the cells. This procures a semi-rigid fit avoiding considerable angular deviations of the pencils with respect to the vertical direction. It is therefore seen that the main advantages of the invention are to allow an efficient semi-rigid fit of the fuel pencils inside the cells of the spacer grating without exerting on these pencils high pressures which might destroy the cladding of the pencil; to reduce the number of added elements made of material absorbing the neutrons, whilst enabling all the pencils to be held identically despite the presence of the guide tubes. In an assembly having an odd number of cells on each of the sides of the grating, the invention allows a simple assembly of the elements for transversely holding the fuel pencils on the grating, these holding elements not passing onto the plates of the grating all the transverse forces exerted by the fuel pencils. However, the invention is not limited to the embodiments which have just been described, but covers all the variants thereof. Plates may thus be made in different shapes from those which have been described. Pieces made of material identical to the material of the plates may also be added, welded at the desired spots. Rivets may also be used, fixed on the plates and whose heads projecting inside the cells would constitute rigid stops. Similarly, for the double springs or the single springs, shapes may be used which differ from those which have been described. These springs may in particular have any number of support points adapted to come into contact with the fuel pencils along a generatrix of these pencils which are in contact via the diametrically opposite generatrix with rigid stops in equal number or in different number. In all the embodiments of the present invention, it is judicious to choose the materials constituting the grating so that the absorption of the neutrons is reduced to the strict minimum. To this end, for example, the plates are most often made of an alloy based on zirconium such as zircaloy which is transparent to neutrons, and the added holding elements are made of a nickel based alloy such as Inconel or nimonic with high mechanical strength. Finally, the spacer grating according to the invention is applicable whatever the number of fuel pencils in the lattice and whatever the type of reactor for which the fuel assembly is intended. |
abstract | A collimator, and in particular a method for making a collimator for use with a small high resolution single-photon emission computed tomographic (SPECT) imaging tool for small animal research. The collimator is sized, both functionally and structurally, particularly smaller than known collimators and appropriately scaled to achieve a highly sensitive collimator which facilitates desired reconstruction resolutions for small animals, as well as compliments other functional imaging modalities such as positron emission tomography (PET), functional magnetic resonance imaging (fMRI), electroencephalography (EEG), and event-related potential (ERP), magneto-encephalography (MEG), and near-infrared optical imaging. |
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abstract | A rotary accelerator (10) accelerates electrons and discharges them through each of a plurality of discharge ports, electrons discharge from each port having a different energy. The electron beams are channeled to scan horns disposed to irradiate products (14) traveling on conveyers (12). More specifically, some of the scan horns are positioned in pairs with an upper scan horn (18) on one side of the product, and a lower scan horn (20) on an opposite side of the product. A beam splitter splits the electron beam alternately between the two scan horns. Alternately, two scan horns (18, 20) are both disposed on the same side of the product. As yet another alternative, a scan horn (60) is disposed horizontally to irradiate the product from the side. Optionally, one or more of the scan horns includes a x-ray target (26) for converting the electrons into x-rays. |
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description | The application claims priority from Application No. DE 201 20 609.9, filed on Dec. 20, 2001. The invention pertains to a diagnostic device for a fluidic device, in particular for a valve array, a fluidic actuator or a maintenance unit. Furthermore, the invention pertains to a fluidic device equipped with said diagnostic device. A fluidic device, for example, a pneumatic valve array, is subject to wear during use, which will adversely impact its operational dependability to an increasing extent over time, and ultimately lead to malfunction or even to complete failure of the fluidic device. If a malfunction or failure has occurred, the fluidic device will then have to be repaired or replaced by an intact fluidic device. However, it is often better from a cost point of view to prevent the malfunction or complete failure by providing advance maintenance of the fluidic device in a timely manner. However, it is difficult to ascertain an optimum maintenance timepoint, since this moment in time can depend on many factors, in particular on the particular strain on the fluidic device caused by its operation. It is therefore the purpose of the present invention to design a diagnostic device for a fluidic device and also a fluidic device itself which will signal the necessary maintenance in a timely manner before occurrence of a malfunction or a total failure of the fluidic device. This problem is solved by a diagnostic device for a fluidic device, in particular for a valve array, a fluidic actuator or a maintenance unit, which features a diagnostic means to ascertain at least one wear parameter causing wear on the fluidic device, and to report at least one wear status determined on the basis of the at least one wear parameter. The problem is furthermore solved by a fluidic device equipped with this kind of diagnostic device. The diagnostic device constantly determines the current wear status of the fluidic device, e.g., of a maintenance unit or a pneumatic or hydraulic valve array, based on the wear parameter. In this regard, the number of piston strokes of a valve can be determined, for example, and after a limit value is reached, the wear status may be signaled so that a valve head, gaskets on the valve, or similar items can be examined and replaced as needed. The diagnostic device monitors the functional integrity of the fluidic device and reports the wear status preferably in a preventive manner, i.e., before there is a malfunction or even total failure of the fluidic device. Additional advantages of the invention are indicated from the dependent claims and from the description. As has already been indicated above, the diagnostic features report the wear status, preferably in a preventive manner. The fluidic device is then at least partly operational and/or has limited operating capability, so that at least a kind of emergency operation will still be possible. However, it is preferable that the fluidic device still be fully operational even after the occurrence of the worn state. Preferably, the fluidic device can still continue to be operated for a predefined time before the maintenance necessary to alleviate the wear status has to be performed. It is preferably to locate the [diagnostic device] at or on the fluidic device. For the wear parameters, according to this invention various quantities can be evaluated by the diagnostic device, of which only a few will be mentioned as examples. Preferably the diagnostic device will evaluate at least one load state of the fluidic device as a wear parameter. For example, it can count the working cycles of the fluidic device, determine the particular fluid consumption of the fluidic device, and/or it can evaluate the particular speed of motion of an actuator element, for example, a piston, of the fluidic device. It is also obvious that any particular combinations of wear parameters can be used, whereby the particular wear parameters can be differently weighted by the diagnostic device. It is preferable that the diagnostic device be equipped to control and/or monitor the fluidic device. Also, the converse is possible, that the diagnostic device be a constituent of a control unit provided to control and/or monitor the fluidic device, in particular one that can be locally attached to or located at the fluidic device. In any case, in both the aforementioned configurations, the control, monitoring and diagnosis should be combined into a single unit. Preferably the diagnostic features are designed to ascertain and/or report at least one interference parameter that signals a fault in the fluidic device. The diagnostic device then reports a fault in the fluidic device, for example, when its actuator element is no longer capable of movement and/or when overheating has occurred. Preferably the diagnostic device contains output features for optical and/or acoustical output of the at least one wear status. The output features can also be located at a distance from the diagnostic device, but still be associated with it. The diagnostic device communicates with the output features by means of a line connection, for example, or by wireless means, e.g., by radio. Preferably the diagnostic features are designed to indicate a need for at least one replacement part suitable for maintaining the functional integrity of the fluidic device. The diagnostic device then orders, more or less, the required spare part. For example, the diagnostic device can indicate in the message pertaining to the wear status, the spare parts' numbers of one or more required replacement parts which are needed to correct the wear status. A message of this kind is sent by the diagnostic device, preferably to a spare parts procurement apparatus, for example, a spare parts depot or such. The message regarding the wear status of the fluidic device can be sent by the diagnostic device to various locations. Preferably it sends the message to a display unit located away from the fluidic device, for example, to an alerting device for maintenance personnel. The display device can be, for example, a pager, a mobile telephone or similar device. But the diagnostic device can also send the message to a higher-order control unit for control of the fluidic device, for example, to a central computer or similar unit, and/or to a neighboring fluidic device cooperating with the fluidic device being monitored by it. It is also similarly for a fault message which the diagnostic device can send to one of the aforementioned destinations. Preferably the diagnostic device is equipped to execute an emergency program so that additional wear during continued operation of the fluidic device can be kept to a minimum or avoided entirely. For example, the operating speed of a pneumatic working cylinder can be reduced by means of the diagnostic device. Of course, with regard to the diagnostic device, in principle the particular values for the at least one wear status and/or the at least one wear parameter can be predetermined as defaults, for example, as permanently programmed values. However, it is preferable that these values be parameterized, i.e., that they be variable for example by a user input and/or by an instruction transmitted by a higher-order controller. Preferably the diagnostic device forms a constituent of the fluidic device. It can be permanently connected to the fluidic device, that is, it can be integrated into its housing or preferably also it can be designed as a replaceable module, for instance, as a circuitboard module that plugs into the fluidic device. Preferably the diagnostic device is designed as an integrated monochip microprocessor array whose component constituents form a single electronic assembly. Preferably the diagnostic device contains program code that can be executed by a processor unit. Of course, it is self-evident that the diagnostic device can be designed entirely as a hardware component or entirely as a software component, or can have both hardware and software constituents. A software diagnostic device according to this invention can be stored preferably on a storage medium, for example, on a diskette, a hard disk drive, a compact disc or similar device. One design example of the invention will be explained in greater detail below. We have: In the present case the fluidic device 10 pertains to a pneumatic device with several fluidic actuators upstream of the working cylinders 12. The device 10 is used, for instance, to drive a handling machine or similar unit. The device 10 is supplied with a compressed medium 15, in this case, compressed air, by a compressed air supply device (not illustrated). The compressed medium 15 is injected through a supply line 16 into the maintenance unit 11 which processes the compressed medium, e.g., by cleaning it and/or oiling it. The filters or additives, e.g., oil or such, needed for this are not shown in the illustration for the sake of simplification. In any case, the maintenance unit 11 supplies the working cylinder 12 with the treated compressed medium—in the present case, cleaned and oiled compressed air—15 through a supply line 17. A piston 18 forming the actuator element is seated in a piston chamber 19 and moves back and forth in the working cylinder 12 which each forms a fluidic device. The piston chamber 19 is located inside a housing 20. The piston chamber 19 can be supplied with compressed air and vented through lines 21, 22, and the piston 18 will move back and forth during these processes; the compressed air comes from a valve array 23 which contains pneumatic control valves actuated by an electromagnet, for instance. In the present device 10 one diagnostic device 14 is associated with each working cylinder 12; in the design example, this device also performs the function of a local controller. However, it would also be possible to associate with each working cylinder 12 a control unit separate from the diagnostic device 14 for local control, and this separate control unit could be an integral constituent of the valve array 23. The diagnostic device 13, which in the present case serves to monitor and to diagnose the maintenance unit 11, is associated with the maintenance unit 11. The diagnostic devices 13, 14 are connected along a bus 24 to a central control computer 25. The control computer 25 controls and monitors the functions of the maintenance unit 11 and of the working cylinder 12. The diagnostic devices 13, 14 each ascertain at least one wear parameter which causes wear on the fluidic device 10. As soon as a wear status of the particular device 10 is determined on the basis of the wear parameter, this status will be reported by the particular diagnostic device 13, 14. The mode of operation of the diagnostic devices 13, 14 will be explained in greater detail below based on the diagnostic device 14 illustrated in FIG. 2. The diagnostic device 14 in the present case is designed as a module, for example, as a monochip microprocessor array, which is locally associated with the working cylinder 12 or with the valve array 23. However, in principle it would also be possible to locate the diagnostic device 14 at a distance from the working cylinder 12 or from the valve array 23. The diagnostic device 14 contains a microprocessor 26, memory features 27, for example, RAM and/or ROM modules (RAM=random access memory, ROM=read only memory), input and output devices 28 as well as interface devices 29 to 31. The microprocessor 26 executes program code from an operating system 32, a control module 33 and a diagnostic module 34 for control and monitoring, or for the diagnosis of the valve array 23 and of the working cylinder 12. At system start of the diagnostic device 14, the modules 32 to 34 are loaded from the memory 27 into the microprocessor 26 and then the coded instructions are executed by it. The input/output devices 28 are composed in the present case of a keyboard and/or a mouse 35, optical output device 36, e.g., a monitor, a liquid crystal display and/or a light-emitting diode device, and acoustical output device 37, e.g., a loudspeaker. The components of the diagnostic device 14, for example, the microprocessor 26, the memory devices 27 and the input and output devices 28 are connected to each other along appropriate connections (not illustrated). A flow sensor 41 ascertains the quantity of compressed medium 15 that flows into the valve array 23 and thus into the working cylinder 12. The sensor 41 sends a measured value 38a representative for the particular rate of flow to the diagnostic module 34 along the interface device 29. The sensor 41 is connected in front of the valve array 23. A position sensor 42 determines the particular position of the piston 18 and depending on this position, it sends associated measured position values 39a along the interfaces of the device 29 to the diagnostic module 34. In addition to the position sensor 42, a temperature sensor 43 is also associated with the working cylinder 12. The sensor 43 determines its temperature and sends associated measured temperature values 40a along the interface device 29 to the diagnostic module 34. The diagnostic module 34 takes the measured values 38a, 39a, 40a and forms the wear parameters 38b, 39b and 40b, respectively. Based on the wear parameters 38b to 40b, the diagnostic module 34 determines at least one wear status of the valve array 23 and/or of the working cylinder 12. A wear status is defined, for instance, by a limit value 38c, which is assigned to the wear parameter 38b. Additional wear states are defined, e.g., by limit values 39c and 40c, which are associated with the wear parameters 39b and 40b. For example, the diagnostic module 34 adds the measured values for fluid flow 38a to the wear parameter 38b until the limit value 38c is reached. Wear on the valve array 23 or on the working cylinder 12 is caused by the compressed medium 15 which is needed to operate the working cylinder 12. Once an upper limit defined by the limit value 38c is reached, specified wear on the valve array 23 and/or on the working cylinder 12 has been reached. The diagnostic module 34 signals this wear status, for example, by means of the output devices 36, 37, by the output of a corresponding optical or acoustical message. For example, a warning tone may be output. Furthermore, a clear text spare parts number or other spare parts description can appear on the output device 37 to indicate the replacement part needed to correct the wear state ascertained based on the flow-through wear parameter 38b. For instance, the order number of gaskets or similar items to be replaced may be displayed. Also, the combined position value 39b forms a wear parameter in the sense of this invention. For example, the diagnostic module 34 counts the working cycles of the working cylinder 12, that is, each back and forth stroke of the piston 18 in the combined [position] value 39b. After a predetermined number of working cycles as defined by the limit value 39c, a default wear state is reached which the diagnostic module 34 sends out to the output devices 36, 37. In this instance, a message such as “20,000 working cycles completed. Check the cylinder!” will be output as text or as a voice message. Based on the measured position values 39b and to form the wear parameter 39b, the diagnostic module 34 can determine the particular speed of movement of the piston 18, its movement behavior or other wear parameters causing wear as derived from the measured position values 39a. For example, the movement behavior of the piston 18 has an effect on the wear of the working cylinder 12. When the piston 18 is moving at a high velocity and/or if it impacts against the particular end stop at a relatively high speed, then this will cause greater wear than if the piston is moving rather slowly and/or if it gently is moved up to the particular end stop. In any case, the diagnostic module 34 can evaluate the measured position value 39b in numerous ways in order to determine one or more wear states of the working cylinder 12. To form the combined value 40b, which likewise represents a wear parameter, the diagnostic module 34 evaluates the measured temperature value 40a. Here, too, the diagnostic module 34 can monitor and/or integrate the measured temperature value 40a for a specified period of time. Furthermore, it is also possible for the diagnostic module 34 to use the measured temperature value 40a only to form the wear parameter 40b when a specified limit temperature value is exceeded. The diagnostic module 34 determines the wear state when the wear parameter 40b exceeds the limit value 40c. The diagnostic module 34, for example, reports this wear status to the central control computer 25 by means of the interface device 31. In the described examples, the diagnostic module 34 determines the wear status based on a single measured value. But it is also possible for several different measured values to be taken into consideration by the diagnostic module 34 in the determination of a wear state. For example, the diagnostic module 34 takes the wear parameters 38b and 39b to form a combined wear parameter 44 in which the parameters 38b, 39b are included at a different weighting. For instance, the parameter 38b is weighted only half as much as the parameter 39b. If the combined wear parameter 44 exceeds a default limit value 45 stored in the memory 27, then the diagnostic module 34 will recognize a wear state which it will then output to the output devices 36, 37 and/or along the interface device 31 to the control computer 25 and/or along the interface device 30 to an alerting device 46. For example, in the determination of a leak in the valve array 23 and/or of the working cylinder 12 which represents a wear state, several different types of measured values can be taken into account by the diagnostic module 34. In the case of a stopped or barely moving piston 18, for example, no compressed medium 15 or almost no compressed medium 15 can flow into the valve array 23. If this is nevertheless the case, then a leak has occurred in the valve array 23 and/or in the working cylinder 12, which may be caused, for instance, by the wear on a gasket, by a porous hose or such. The diagnostic module 34 recognizes this kind of wear state by the fact that the measured position value 39a changes little or not at all, but the measured flow value 38a exceeds a default value. A message with information on the spare part needed to correct the wear state can be sent by the diagnostic module 14, for example, to a spare parts procurement device 60, e.g., a spare parts depot, or to another logistics system. In this case the diagnostic module 14 will send an SMS message, for example. The procurement device 60 will make the needed spare part available for the fluidic device 10 by having the particular spare part shipped by post (for example) to the site of the device 10. The alerting device 46 pertains to a mobile radio telephone or to a pager, for example. The diagnostic device 16 will send an SMS message (SMS=short message service), for example, to the alerting device 46, in which the particular wear state and/or the spare part needed to correct the wear state is indicated. The interface device 31 forms a radio interface. But it would also be possible for the alerting device 46 and/or the spare parts procurement system 60 to be connected by a wire, for example, by a bus connection, to the diagnostic module 14. With regard to the diagnostic module 14 the limit values 38c to 40c, 45 can be parameterized. To do this, for example, corresponding values can be input by the keyboard 35. But it would also be possible that the diagnostic module 14 could be parameterized at a distance, for example, via the Internet, and the diagnostic module 14 would then be made available for parametering of the interfaces 38a-40a, 45 by a user interface operated along an internet browser. Furthermore, it is possible that even more values can be parameterized by the diagnostic module 14. For instance, the weighting factors that are used for weighting of the parameters 38b 39b in the formation of the parameter 44, could be themselves parameterized. Furthermore, the parameter 39b could be parameterized, for example, by specifying whether the number of working cycles of the working cylinder 12 and/or its motion behavior are to be taken into account in the determination of the particular wear state. The diagnostic module 34 determines a fault in the working cylinder 12 and/or in the valve array 23 based on a fault parameter 47. For example, the fault parameter 47 may pertain to an overheating of the working cylinder 12 that the diagnostic module 34 determines based on the measured temperature value 40a. Also, the diagnostic module 34 can evaluate the measured values 38a, 39a in the determination of the fault parameter 47. For example, an unbraked impact of the piston 18 against an end stop can be determined with the aid of measured value 39a. Furthermore, a fault may also consist in that the working cylinder 12 and/or the valve array 23 has a leak. This can be determined by the diagnostic module 34 since the measured value 39b will not change, i.e., the piston 18 moves to a defined position, but the measured value 38a exceeds a default limit value, i.e., compressed medium 15 is flowing into the valve array 23, but the piston 18 is not then caused to move. In any case, the diagnostic module 34 detects a fault by reporting it, for example, to the alerting device 46 and/or to the control computer 25 and/or to a neighboring fluidic device, for instance, to the maintenance unit 11 or to a neighboring working cylinder 12. If one or more of the above-mentioned wear states occurs and/or one of the fault states occurs, then the diagnostic module 34 can execute an emergency program. For example, the diagnostic module 34 will send a corresponding instruction to the control module 33. The control module 33 will then reduce the working speed and/or the working cycle of the piston 18, which will result in a reduced wear on the working cylinder 12. If a fault occurs, it is also possible that the control module 33 will shut down the working cylinder 12 and/or the valve array 23. In the diagnostic unit 14, the diagnostic module 34 and the control module 33 are designed as separate units. A combined control and diagnostic module that performs the functions of both modules 33, 34 would also be possible. The control module 33 sends switching commands 48 to the valve array 23 and receives from it the associated acknowledgment messages 49 when a switching command 48 has been executed. It is obvious that the control module 33 can send a message regarding its execution of a switching command to the diagnostic module 34, so that the diagnostic module 34 can determine one or more wear states of the valve array 23 and/or of the working cylinder 12 on the basis of such messages. The diagnostic device 13 of the maintenance unit 11 can evaluate different wear parameters in order to determine the wear status. For instance, it can determine a quantity of additive that is injected into the compressed medium 15, and then if a default quantity has been reached, it will report the wear status, for example, to the control computer 25 and/or to the alerting device 46. Furthermore, as a wear parameter, the diagnostic device 13 can ascertain a temperature that causes elevated wear. In addition, for example, as a wear parameter it can also evaluate the quantity of the compressed medium 15 flowing through the maintenance unit 11, a pressure loading or such to determine the wear status. Combinations of the aforementioned parameters, and other wear parameters as well, would be possible for a determination of the wear status of the maintenance unit 11. |
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056404352 | summary | TECHNICAL FIELD The present invention relates to a fuel assembly, and more particularly to a fuel assembly which can be used in a boiling-water reactor to save the consumption of nuclear fuel substances. BACKGROUND ART In a conventional boiling-water reactor as is disclosed in Japanese Patent Laid-Open No. 121389/1979, the reactor core is loaded with a fuel assembly which has a pipe (hereinafter referred to as water rod) in which the cooling water only flows to decelerate the neutrons. Under the operation conditions of the conventional boiling-water reactor, the water rod exhibits an increased reactivity with the increase in the number of hydrogen atoms for uranium atoms, enabling the nuclear fuel substances loaded in the reactor core to be effectively utilized. In order to more effectively use the nuclear fuel substances, furthermore, it is recommended to change the number of hydrogen atoms in the reactor core as the nuclear fuel substances burn. Japanese Patent Laid-Open Nos. 125390/1982 and 125391/1989 teach one of the methods. That is, according to these patent publications, provision is made of slow neutron-absorbing water purge rods and intermediate neutron-absorbing water purge rods constituted by a stainless steel which has a larger reactivity value than that of the above water purge rods, and the amount of the cooling water in the reactor core is adjusted by controlling the amount for inserting the water purge rods in the reactor core. The water purge rods serve as means for changing the number of hydrogen atoms in the reactor core. The amount of the cooling water in the reactor core decreases with the increase in the amount for inserting the water purge rods in the reactor core, and the amount of the cooling water increases in the reactor core with the decrease in the amount of insertion. According to the above-mentioned method, water purge rods of different kinds must be newly provided and must be operated by drive means, requiring complex structure and cumbersome operation. Japanese Patent Laid-Open No. 38589/1986 discloses a fuel assembly which employs static means in order to solve the above-mentioned problems. According to this patent publication, the number of hydrogen atoms is changed by providing fuel rods having a low uranium 235 concentration in the water rod of fuel assembly, and by utilizing the change in the amount of voids in the water rod before and after uranium 235 of the fuel rods extinguishes. There is a method of adjusting the amount of the cooling water that flows in the reactor core without the need of newly providing operation means such as water purge rods. That is, the cooling water is permitted to flow in small amounts in the reactor core during the start of the fuel cycle, and is then permitted to flow in increased amounts as the fuel cycle proceeds halfway. Advantages will now be described in the case when the number of hydrogen atoms is changed in the reactor core accompanying the burn of the nuclear fuel substances. In the case of a typical fuel assembly used for boiling-water reactors, a higher burning degree can be obtained when the operation is carried out at a high void fraction (void fraction, 50%) during the period of a burning degree of 0 to 30 GWD/T and when the operation is carried out at a decreased void fraction (void fraction, 30%) during the period of a burning degree of 30 to 40 GWD/T than when the operation is carried out at a constant void fraction (e.g., at a void fraction of 30%). This is because, the neutrons have a high average speed and are easily absorbed by uranium 238 when the void fraction is high and the ratio of the number of hydrogen atoms to the number of uranium atoms is small, i.e., when the number of hydrogen atoms is small. The nuclear fuel substances used in the boiling-water reactor contains uranium 235 and uranium 238, uranium 235 occupying several percent of the whole nuclear fuel substances and uranium 238 occupying most of the nuclear fuel substances. Among them, uranium 235 absorbs the neutrons and develops chiefly the nuclear fission, but uranium 238 develops nuclear fission very little. Therefore, the burn-up decreases if uranium 235 burns and decreases. Uranium 238, however, is converted into plutonium 239 when it absorbs neutrons of a large energy produced by the nuclear fission. Like uranium 235, however, plutonium 239 absorbs decelerated thermal neutrons to develop nuclear fission. The higher the void friction, the larger the energy of the neutrons and uranium 238 is converted into plutonium 239 at an increased ratio, while suppressing the nuclear fission of uranium 235 and plutonium 239. Therefore, the higher the void fraction, the slower the rate of reduction of the total amount of uranium 235 and plutonium 239. A high void fraction, however, causes the absolute value of reactivity to decrease. If the void fraction is maintained high, therefore, a minimum level is reached quickly at which the reactivity maintains the criticality compared with when the void fraction is low. Therefore, if the void fraction is lowered at that moment, the neutrons exhibit increased deceleration effect, whereby nuclear fission of uranium 235 and plutonium 239 increases, so that good reactivity is obtained compared with when the fuel substances are burned at a high void fraction that is maintained constant. This makes it possible to burn the core material contained in the nuclear fuel substances for an extended period of time before a minimum reactivity necessary for the criticality is reached. In the foregoing was mentioned the principle which is called spectrum shift operation for effectively utilizing the nuclear fuel substances by changing the void fraction accompanying the burn of the core material. Neither the method which provides static means in a simply constructed water rod nor the method which changes the number of hydrogen atoms in the reactor core by changing the amount of the cooling water (called reactor core flow rate) which flows through the reactor core, makes it possible to widely change the void fraction in the reactor core; i.e., these methods can only give small effect in the practical nuclear reactors. That is, the lower limit of the flow rate in the reactor core is determined by the thermal limit, and the upper limit is determined by the capacity of the circulation pump and the flow-induced vibration. Under the condition where the boiling-water reactor is producing a rated thermal output, therefore, it is allowed to change the void fraction only within a narrow range with the rated 100% flow rate in the reactor core as a center. For example, if the flow rate in the reactor core is allowed to change over a range of from 80 to 120%, then the void fraction can be changed by about 9%. Even with the structure in which a heat generating member (nuclear fuel substance) of which the calorific power decreases accompanying the burn, is placed in the water rod as disclosed in Japanese Patent Laid-Open No. 38589/1986, the void fraction in the water rod changes by about 30% at the greatest. The water in the water rod does not contribute to the cooling, and it is not allowed to much increase the sectional area of the water rod in the fuel assembly. If it is presumed that the sectional area of the water rod occupies 30% of the cooling water path in the fuel assembly, the effective void fraction change of 30% becomes 9% (30%.times.0.3) if it is regarded as the whole fuel assembly. Further, since a fuel rod having a low enrichment is used as a heat generating member, the structure becomes complex and its production involves cumbersome operation. To achieve a wide range of void fraction change, the flow rate in the water rod should be changed extremely greatly or the calorific power of the nuclear fuel substance in the water rod should be changed greatly. In fact, however, the flow rate or the calorific power cannot be greatly changed without employing the moving portions. Provision of the moving portions, however, poses problems from the standpoint of reliability and makes the mechanism complex. SUMMARY OF INVENTION The object of the present invention is to provide a fuel assembly which is simply constructed and which is capable of greatly changing the internal average void fraction. The aforementioned object is achieved by the provision of a resistance member at the lower end portion of the fuel assembly; a coolant ascending path in which the water rods have coolant inlet ports that are open in a region lower than the resistance member; and a coolant descending path which is communicated with the coolant ascending path and which has a coolant delivery port that is open in a region higher than the resistance member, in order to guide the coolant downwardly which is opposite to the direction in which the coolant flows in the coolant ascending path. As the flow rate of the coolant that passes through the reactor core decreases, the coolant ascending path of the water rod is filled with water vapor and as the flow rate of the coolant increases, the amount of water vapor decreases conspicuously in the coolant ascending path. Therefore, the reactivity can be increased toward the last period of fuel cycle. |
summary | ||
description | The present application claims the benefit of priority to U.S. Provisional Application No. 62/061,089 filed Oct. 7, 2014, the entirety of which is incorporated herein by reference. The present invention generally relates to storage of nuclear fuel assemblies, and more particularly to an improved spent fuel pool for wet storage of such fuel assemblies. A spent fuel pool (sometimes, two or more) is an integral part of every nuclear power plant. At certain sites, standalone wet storage facilities have also been built to provide additional storage capacity for the excess fuel discharged by the reactors. An autonomous wet storage facility that serves one or more reactor units is sometimes referred to by the acronym AFR meaning “Away-from-Reactor.” While most countries have added to their in-plant used fuel storage capacity by building dry storage facilities, the French nuclear program has been the most notable user of AFR storage. As its name implies, the spent fuel pool (SFP) stores the fuel irradiated in the plant's reactor in a deep pool of water. The pool is typically 40 feet deep with upright Fuel Racks positioned on its bottom slab. Under normal storage conditions, there is at least 25 feet of water cover on top of the fuel to ensure that the dose at the pool deck level is acceptably low for the plant workers. Fuel pools at most (but not all) nuclear plants are at grade level, which is desirable from the standpoint of structural capacity of the reinforced concrete structure that forms the deep pond of water. To ensure that the pool's water does not seep out through the voids and discontinuities in the pool slab or walls, fuel pools in nuclear plants built since the 1970s have always been lined with a thin single-layer stainless steel liner (typically in the range of 3/16 inch to 5/16 inch thick). The liner is made up of sheets of stainless steel (typically ASTM 240-304 or 304L) seam welded along their contiguous edges to form an impervious barrier between the pool's water and the undergirding concrete. In most cases, the welded liner seams are monitored for their integrity by locating a leak chase channel underneath them (see, e.g. FIG. 1). The leak chase channels' detection ability, however, is limited to welded regions only; the base metal area of the liner beyond the seams remains un-surveilled. The liners have generally served reliably at most nuclear plants, but isolated cases of water seepage of pool water have been reported. Because the pool's water bears radioactive contaminants (most of it carried by the crud deposited on the fuel during its “burn” in the reactor), leaching out of the pool water to the plant's substrate, and possibly to the underground water, is evidently inimical to public health and safety. To reduce the probability of pool water reaching the ground water, the local environment and hence some AFR pools have adopted the pool-in-pool design wherein the fuel pool is enclosed by a secondary outer pool filled with clean water. In the dual-pool design, any leakage of water from the contaminated pool will occur into the outer pool, which serves as the barrier against ground water contamination. The dual pool design, however, has several unattractive aspects, viz.: (1) the structural capacity of the storage system is adversely affected by two reinforced concrete containers separated from each other except for springs and dampers that secure their spacing; (2) there is a possibility that the outer pool may leak along with the inner pool, defeating both barriers and allowing for contaminated water to reach the external environment; and (3) the dual-pool design significantly increases the cost of the storage system. Prompted by the deficiencies in the present designs, a novel design of a spent nuclear fuel pool that would guarantee complete confinement of pool's water and monitoring of the entire liner structure including seams and base metal areas is desirable. The present invention provides an environmentally sequestered spent fuel pool system having a dual impervious liner system and leak detection/evacuation system configured to collect and identify leakage in the interstitial space formed between the liners. The internal cavity of the pool has not one but two liners layered on top of each other, each providing an independent barrier to the out-migration (emigration) of pool water. Each liner encompasses the entire extent of the water occupied space and further extends above the pool's “high water level.” The top of the pool may be equipped with a thick embedment plate (preferably 2 inches thick minimum in one non-limiting embodiment) that circumscribes the perimeter of the pool cavity at its top extremity along the operating deck of the pool. Each liner may be independently welded to the top embedment plate. The top embedment plate features at least one telltale hole, which provides direct communication with the interstitial space between the two liner layers. In one implementation, a vapor extraction system comprising a vacuum pump downstream of a one-way valve is used to draw down the pressure in the inter-liner space through the telltale hole to a relatively high state of vacuum. The absolute pressure in the inter-liner space (“set pressure”) preferably should be such that the pool's bulk water temperature is above the boiling temperature of water at the set pressure as further described herein. In one embodiment, an environmentally sequestered nuclear spent fuel pool system includes: a base slab; a plurality of vertical sidewalls extending upwards from and adjoining the base slab, the sidewalls forming a perimeter; a cavity collectively defined by the sidewalls and base slab that holds pool water; a pool liner system comprising an outer liner adjacent the sidewalls, an inner liner adjacent the outer liner and wetted by the pool water, and an interstitial space formed between the liners; a top embedment plate circumscribing the perimeter of the pool at a top surface of the sidewalls adjoining the cavity; and the inner and outer sidewalls having top terminal ends sealably attached to the embedment plate. In another embodiment, an environmentally sequestered nuclear spent fuel pool with leakage detection system includes: a base slab; a plurality of vertical sidewalls extending upwards from and adjoining the base slab, the sidewalls forming a perimeter; a cavity collectively defined by the sidewalls and base slab that holds pool water; at least one fuel storage rack disposed in the cavity that holds a nuclear spent fuel assembly containing nuclear fuel rods that heat the pool water; a pool liner system comprising an outer liner adjacent the sidewalls and base slab, an inner liner adjacent the outer liner and wetted by the pool water, and an interstitial space formed between the liners; a top embedment plate circumscribing the perimeter of the pool, the embedment plate embedded in the sidewalls adjoining the cavity; the inner and outer liners attached to the top embedment plate; a flow plenum formed along the sidewalls that is in fluid communication with the interstitial space; and a vacuum pump fluidly coupled to the flow plenum, the vacuum pump operable to evacuate the interstitial space to a negative set pressure below atmospheric pressure. A method for detecting leakage from a nuclear spent fuel pool is provided. The method includes: providing a spent fuel pool comprising a plurality of sidewalls, a base slab, a cavity containing cooling water, and a liner system disposed in the cavity including an outer liner, an inner liner, and an interstitial space between the liner; placing a fuel storage rack in the pool; inserting at least one nuclear fuel assembly into the storage rack, the fuel assembly including a plurality of spent nuclear fuel rods; heating the cooling water in the pool to a first temperature from decay heat generated by the spent nuclear fuel rods; drawing a vacuum in the interstitial space with a vacuum pump to a negative pressure having a corresponding boiling point temperature less than the first temperature; collecting cooling water leaking from the pool through the liner system in the interstitial space; converting the leaking cooling water into vapor via boiling; and extracting the vapor from the interstitial space using the vacuum pump; wherein the presence of vapor in the interstitial space allows detection of a liner breach. The method may further include discharging the vapor extracted by the vacuum pump through a charcoal filter to remove contaminants. The method may further include: monitoring a pressure in the interstitial space; detecting a first pressure in the interstitial space prior to collecting cooling water leaking from the pool through the liner system in the interstitial space; and detecting a second pressure higher than the first pressure after collecting cooling water leaking from the pool through the liner system in the interstitial space; wherein the second pressure is associated with a cooling water leakage condition. All drawings are schematic and not necessarily to scale. Parts shown and/or given a reference numerical designation in one figure may be considered to be the same parts where they appear in other figures without a numerical designation for brevity unless specifically labeled with a different part number and described herein. References herein to a figure number (e.g. FIG. 1) shall be construed to be a reference to all subpart figures in the group (e.g. FIGS. 1A, 1B, etc.) unless otherwise indicated. The features and benefits of the invention are illustrated and described herein by reference to exemplary embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Referring to FIGS. 2-6, an environmentally sequestered spent fuel pool system includes a spent fuel pool 40 comprising a plurality of vertical sidewalls 41 rising upwards from an adjoining substantially horizontal base wall or slab 42 (recognizing that some slope may intentionally be provided in the upper surface of the bottom wall for drainage toward a low point if the pool is to be emptied and rinsed/decontaminated at some time and due to installation tolerances). The base slab 42 and sidewalls 41 may be formed of reinforced concrete in one non-limiting embodiment. The fuel pool base slab 42 may be formed in and rest on the soil sub-grade 26 the top surface of which defines grade G. In this embodiment illustrated in the present application, the sidewalls are elevated above grade. In other possible embodiments contemplated, the base slab 42 and sidewalls 41 may alternatively be buried in sub-grade 26 which surrounds the outer surfaces of the sidewalls. Either arrangement may be used and does not limit of the invention. In one embodiment, the spent fuel pool 40 may have a rectilinear shape in top plan view. Four sidewalls 41 may be provided in which the pool has an elongated rectangular shape (in top plan view) with two longer opposing sidewalls and two shorter opposing sidewalls (e.g. end walls). Other configurations of the fuel pool 40 are possible such as square shapes, other polygonal shapes, and non-polygonal shapes. The sidewalls 41 and base slab 42 of the spent fuel pool 40 define a cavity 43 configured to hold cooling pool water W and a plurality of submerged nuclear spent fuel assembly storage racks 27 holding fuel bundles or assemblies 28 each containing multiple individual nuclear spent fuel rods. The storage racks 27 are disposed on the base slab 42 in typical fashion. With continuing reference to FIGS. 1-6, the spent fuel pool 40 extends from an operating deck 22 surrounding the spent fuel pool 40 downwards to a sufficient depth D1 to submerge the fuel assemblies 28 (see, e.g. FIG. 6) beneath the surface level S of the pool water W for proper radiation shielding purposes. In one implementation, the fuel pool may have a depth such that at least 10 feet of water is present above the top of the fuel assembly. A nuclear fuel assembly storage rack 27 is shown in FIGS. 2 and 3, and further described in commonly assigned U.S. patent application Ser. No. 14/367,705 filed Jun. 20, 1014, which is incorporated herein by reference in its entirety. The storage rack 27 contains a plurality of vertically elongated individual cells as shown each configured for holding a fuel assembly 28 comprising a plurality of individual nuclear fuel rods. An elongated fuel assembly 28 is shown in FIG. 6 holding multiple fuel rods 28a and further described in commonly assigned U.S. patent application Ser. No. 14/413,807 filed Jul. 9, 2013, which is incorporated herein by reference in its entirety. Typical fuel assemblies 28 for a pressurized water reactor (PWR) may each hold over 150 fuel rods in 10×10 to 17×17 fuel rod grid arrays per assembly. The assemblies may typically be on the order of approximately 14 feet high weighing about 1400-1500 pounds each. The substantially horizontal operating deck 22 that circumscribes the sidewalls 41 and pool 40 on all sides in one embodiment may be formed of steel and/or reinforced concrete. The surface level of pool water W (i.e. liquid coolant) in the pool 40 may be spaced below the operating deck 22 by a sufficient amount to prevent spillage onto the deck during fuel assembly loading or unloading operations and to account to seismic event. In one non-limiting embodiment, for example, the surface of the operating deck 22 may be at least 5 feet above the maximum 100 year flood level for the site in one embodiment. The spent fuel pool 40 extending below the operating deck level may be approximately 40 feet or more deep (e.g. 42 feet in one embodiment). The fuel pool is long enough to accommodate as many spent fuel assemblies as required. In one embodiment, the fuel pool 40 may be about 60 feet wide. There is sufficient operating deck space around the pool to provide space for the work crew and for staging necessary tools and equipment for the facility's maintenance. There may be no penetrations in the spent fuel pool 40 within the bottom 30 feet of depth to prevent accidental draining of water and uncovering of the spent fuel. According to one aspect of the invention, a spent fuel pool liner system comprising a double liner is provided to minimize the risk of pool water leakage to the environment. The liner system is further designed to accommodate cooling water leakage collection and detection/monitoring to indicate a leakage condition caused by a breach in the integrity of the liner system. The liner system comprises a first outer liner 60 separated from a second inner liner 61 by an interstitial space 62 formed between the liners. An outside surface of liner 60 is disposed against or at least proximate to the inner surface 63 of the fuel pool sidewalls 41 and opposing inside surface is disposed proximate to the interstitial space 62 and outside surface of liner 61. The inside surface of liner 61 is contacted and wetted by the fuel pool water W. It bears noting that placement of liner 60 against liner 61 without spacers therebetween provides a natural interstitial space of sufficient width to allow the space and any pool leakage there-into to be evacuated by a vacuum system, as further described herein. The natural surface roughness of the materials used to construct the liners and slight variations in flatness provides the needed space or gap between the liners. In other embodiments contemplated, however, metallic or non-metallic spacers may be provided which are distributed in the interstitial space 62 between the liners if desired. The liners 60, 61 may be made of any suitable metal which is preferably resistant to corrosion, including without limitation stainless steel, aluminum, or other. In some embodiments, each liner may be comprised of multiple substantially flat metal plates which are seal welded together along their peripheral edges to form a continuous liner system encapsulating the sidewalls 41 and base slab 42 of the spent fuel pool 40. The inner and outer liners 61, 60 may have the same or different thicknesses (measured horizontally or vertically between major opposing surfaces of the liners depending on the position of the liners). In one embodiment, the thicknesses may be the same. In some instances, however, it may be preferable that the inner liner 61 be thicker than the outer liner 60 for potential impact resistant when initially loading empty fuel storage racks 27 into the spent fuel pool 40. The outer and inner liners 60, 61 (with interstitial space therebetween) extend along the vertical sidewalls 41 of the spent fuel pool 40 and completely across the horizontal base slab 42 in one embodiment to completely cover the wetted surface area of the pool. This forms horizontal sections 60b, 61b and vertical sections 60a, 61a of the liners 60, 61 to provide an impervious barrier to out-leakage of pool water W from spent fuel pool 40. The horizontal sections of liners 60b, 61b on the base slab 42 may be joined to the vertical sections 60a, 61a along the sidewalls 41 of the pool 40 by welding. The detail in FIG. 4 shows one or many possible constructions of the bottom liner joint 64 comprising the use of seal welds 65 (e.g. illustrated fillet welds or other) to seal sections 60a to 60b along their respective terminal edges and sections 61a to 61b along their respective terminal edges as shown. Preferably, the joint 64 is configured and arranged to fluidly connect the horizontal interstitial space 64 between horizontal liner sections 60b, 61b to the vertical interstitial space 64 between vertical liner sections 60a, 61a for reasons explained elsewhere herein. The top liner joint 65 in one non-limiting embodiment between the top terminal edges 60c, 61c of the vertical liner sections 60a, 61a is shown in the detail of FIG. 5. The top of the spent fuel pool 40 is equipped with a substantially thick metal embedment plate 70 which circumscribes the entire perimeter of the fuel pool. The embedment plate 70 may be continuous in one embodiment and extends horizontally along the entire inner surface 63 of the sidewalls 41 at the top portion of the sidewalls. The embedment plate 70 has an exposed portion of the inner vertical side facing the pool which extends above the top terminal ends 60c, 61c of the inner and outer liners 60, 61. The opposing outer vertical side of the plate 70 is embedded entirely into the sidewalls 41. A top surface 71 of the embedment plate 70 that faces upwards may be substantially flush with the top surface 44 of the sidewalls 41 to form a smooth transition therebetween. In other possible implementations, the top surface 71 may extend above the top surface 44 of the sidewalls. The embedment plate 70 extends horizontal outward from the fuel pool 40 for a distance into and less than the lateral width of the sidewalls 41 as shown. The embedment plate 70 has a horizontal thickness greater than the horizontal thickness of the inner liner 61, outer liner 60, and in some embodiments both the inner and outer liners combined. The top embedment plate 70 is embedded into the top surface 44 of the concrete sidewalls 41 has a sufficient vertical depth or height to allow the top terminal edges 60c, 61c of liners 60, 61 (i.e. sections 60a and 61a respectively) to be permanently joined to the plate. The top terminal edges of liners 60, 61 terminate at distances D2 and D1 respectively below a top surface 71 of the embedment plate 70 (which in one embodiment may be flush with the top surface of the pool sidewalls 41 as shown). Distance D1 is less than D2 such that the outer liner 60 is vertical shorter in height than the inner liner 61. In one embodiment, the embedment plate 70 has a bottom end which terminates below the top terminal edges 60i cl , 61i c l of the liners 60, 61 to facilitate for welding the liners to the plate. In various embodiments, the embedment plate 70 may be formed of a suitable corrosion resistant metal such as stainless steel, aluminum, or another metal which preferably is compatible for welding to the metal used to construct the outer and inner pool liners 60, 61 without requiring dissimilar metal welding. As best shown in FIG. 5, the top terminal edges 60c, 61c of inner and outer liners 60, 61 may have a vertically staggered arranged and be separately seal welded to the top embedment plate 70 independently of each other. A seal weld 66 couples the top terminal edge 61c of liner 61 to the exposed portion of the inner vertical side of the embedment plate 70. A second seal weld 67 couples the top terminal edge 60c of liner 60 also to the exposed portion of the inner vertical side of the embedment plate 70 at a location below and spaced vertical apart from seal weld 66. This defines a completely and hermetically sealed enclosed flow plenum 68 that horizontal circumscribes the entire perimeter of the spent fuel pool 40 in one embodiment. The flow plenum 68 is in fluid communication with the interstitial space 62 as shown. One vertical side of the flow plenum is bounded by a portion of inner liner 61 and the opposing vertical side of the plenum is bounded by the inner vertical side of the top embedment plate 70. The top flow plenum 68 may be continuous or discontinuous in some embodiments. Where discontinuous, it is preferable that a flow passageway 105 in the top embedment plate 70 be provided for each section of the separate passageways. Seal welds 66 and 67 may be any type of suitable weld needed to seal the liners 60, 61 to the top embedment plate 70. Backer plates, bars, or other similar welding accessories may be used to make the welds as needed depending on the configuration and dimensions of the welds used. The invention is not limited by the type of weld. In one embodiment, the outer and inner liners 60, 61 are sealably attached to the spent fuel pool 40 only at top embedment plate 70. The remaining portions of the liners below the embedment plate may be in abutting contact with the sidewalls 41 and base slab 42 without means for fixing the liners to these portions. It bears noting that at least the inner liner 61 has a height which preferably is higher than the anticipated highest water level (surface S) of the pool water W in one embodiment. If the water level happens to exceed that for some reason, the top embedment plate 70 will be wetted directly by the pool water and contain the fluid to prevent overflowing the pool onto the operating deck 22. According to another aspect of the invention, a vapor extraction or vacuum system 100 is provided that is used to draw down the air pressure in the interstitial space between the outer and inner liners 60, 61 to a relatively high state of vacuum for leakage control and/or detection. FIG. 7 is a schematic diagram of one embodiment of a vacuum system 100. Referring to FIGS. 5 and 7, vacuum system 100 generally includes a vacuum pump 101 and a charcoal filter 102. Vacuum pump 101 may be any suitable commercially-available electric-driven vacuum pump capable of creating a vacuum or negative pressure within the interstitial space 62 between the pool liners 60 and 61. The vacuum pump 101 is fluidly connected to the interstitial space 68 via a suitable flow conduit 103 which is fluidly coupled to a telltale or flow passageway 105 extending from the top surface 71 of the top embedment plate 70 to the top flow plenum 68 formed between the pool liners 60 and 61. Flow conduit 103 may be formed of any suitable metallic or non-metallic tubing or piping capable of withstanding a vacuum. A suitably-configured fluid coupling 104 may be provided and sealed to the outlet end of the flow passageway 105 for connecting the flow conduit 103. The inlet end of the flow passageway penetrates the inner vertical side of top embedment plate 70 within the flow plenum 68. The flow passageway 105 and external flow conduit 103 provides a contiguous flow conduit that fluidly couples the flow plenum 68 to the vacuum pump 101. A one-way check valve is disposed between the flow plenum 105 and the suction inlet of the vacuum pump 101 to permit air and/or vapor to flow in a single direction from the liner system to the pump. The absolute pressure maintained by the vacuum system 100 in the interstitial space 62 between the liners 60, 61 (i.e. “set pressure”) preferably should be such that the bulk water temperature of the spent fuel pool 40 which is heated by waste decay heat generated from the fuel rods/assemblies is above the boiling temperature of water at the set pressure. The table below provides the boiling temperature of water at the level of vacuum in inches of mercury (Hg) which represent some examples of set pressures that may be used. . Pressure in inch, HgABoiling Temp, deg F.1792101311541255133 Any significant rise in pressure would indicate potential leakage of water in the interstitial space 62 between the liners 60, 61. Because of sub-atmospheric conditions maintained by the vacuum pump 101 in the interstitial space, any water that may leak from the pool into this space through the inner liner 61 would evaporate, causing the pressure to rise which may be monitored and detected by a pressure sensor 104. The vacuum pump 101 preferably should be set to run and drive down the pressure in the interstitial space 62 to the “set pressure.” In operation as one non-limiting example, if the vacuum pump 101 is operated to create a negative pressure (vacuum) in the interstitial space 62 of 2 inches of Hg, the corresponding boiling point of water at that negative pressure is 101 degrees Fahrenheit (degrees F.) from the above Table. If the bulk water temperature of pool water W in the spent fuel pool 40 were at any temperature above 101 degrees F. and leakage occurred through the inner pool liner 61 into the interstitial space 62, the liquid leakage would immediately evaporate therein creating steam or vapor. The vacuum pump 101 withdraws the vapor through the flow plenum 68, flow passageway 105 in the top embedment plate 70, and flow conduit 103 (see, e.g. directional flow arrows of the water vapor in FIGS. 5 and 7). Pressure sensor 104 disposed on the suction side of the pump 101 would detect a corresponding rise in pressure indicative of a potential leak in the liner system. In some embodiments, the pressure sensor 104 may be operably linked to a control panel of a properly configured computer processor based plant monitoring system 107 which monitors and detects the pressure measured in the interstitial space 62 between the liners on a continuous or intermittent basis to alert operators of a potential pool leakage condition. Such plant monitoring systems are well known in the art without further elaboration. The extracted vapor in the exhaust or discharge from the vacuum pump 101 is routed through a suitable filtration device 102 such as a charcoal filter or other type of filter media before discharge to the atmosphere, thereby preventing release of any particulate contaminants to the environment. Advantageously, it bears noting that if leakage is detected from the spent fuel pool 40 via the vacuum system 100, the second outer liner 60 encapsulating the fuel pool provides a secondary barrier and line of defense to prevent direct leaking of pool water W into the environment. It bears noting that there is no limit to the number of vapor extraction systems including a telltale passageway, vacuum pump, and filter combination with leakage monitoring/detection capabilities that may be provided. In some instances, four independent systems may provide adequate redundancy. In addition, it is also recognized that a third or even fourth layer of liner may be added to increase the number of barriers against leakage of pool water to the environment. A third layer in some instances may be used as a palliative measure if the leak tightness of the first inter-liner space could not, for whatever reason, be demonstrated by a high fidelity examination in the field such as helium spectroscopy. While the foregoing description and drawings represent exemplary embodiments of the present disclosure, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made within the scope of the present disclosure. One skilled in the art will further appreciate that the embodiments may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the disclosure, which are particularly adapted to specific environments and operative requirements without departing from the principles described herein. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive. The appended claims should be construed broadly, to include other variants and embodiments of the disclosure, which may be made by those skilled in the art without departing from the scope and range of equivalents. |
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039986942 | summary | BACKGROUND OF THE INVENTION This invention relates to deflection instruments and their use for the measurement of various kinds of forces; in particular, it relates to an improved method and means for the continuous monitoring and recording of the phenomena measured by such instruments. In certain respects, this application is similar to a former publication, U.S. Pat. No. 2,986,697. However, it contains substantial improvements and innovations beyond those described in the earlier patent. In many deflection instruments, the forces deflecting the movable element are subject to a field of force of one kind or another: where the relationship between the field and the forces or between the forces themselves are known, measurements may be undertaken. For example, when an electrostatic field of force controls the movement of a member that is collecting an electric charge, an electric current may be measured. Known physical laws relate the average current and the rate of deflection of the movable element. More in particular, it is often desirable to use a quartz fiber electrometer in conjunction with an ionization chamber for radiation measurements. A well-designed electrometer is sufficiently accurate to be employed as a secondary standard but requires operator observation of the rate of deflection of the electrometer needle. The need for continuous operator attention has inhibited use of an electrometer for routine measurements despite its many desirable characteristics. It is therefore an object of this invention to provide an improved method for the automatic monitoring of the phenomena measured by such deflection instruments. It is also the object of this invention to provide other types of deflection instruments which are useful in various manufacturing, production or controlled process operations. These new types of deflection instruments have similar properties to the automatically operated electrometer which will first be described. The electrometer is a deflection instrument having a movable element in an electric field of force. Since the field is electric the forces will be produced by electric charges (of opposite sign) supplied to the moving element. Also, associated with this instrument is a source of illumination and an optical system adjusted to focus the real image of the deflection element (usually provided in the form of a needle clamped at one end and free to vibrate at the other) upon an optical mask having one slit thereon. A photoelectric cell is either mounted behind the slit or is connected to the slit by means of fiber optics so that signals may be generated when the deflection element is in a null condition. The signals generated through the instrumentality of the real-image-optical-slit-photocell arrangement are then employed both to operate a feedback control for the instrument itself, as well as to control digital time counters and recorders, thus to store in memory digital information of the rate of change of the potential of the needle due to the charge accumulating thereon. In employing such a unit of invention in connection with an ion chamber and a quartz fiber needle electrometer for the measurement of radiation, the "rate of charge" or "drift" method is employed and currents of less than about 10-.sup.16 ampere may be measured. In the device first to be described, time is the dependent variable and is the principal quantity under measurement. Independent variables, such as voltage, may be set by control nobs or are variables to be measured, as is the electric current with quartz fiber electrometers. Constants are either built into the device or are set by controls. As a result, all measurements appear as digital, time-interval readings. This makes it natural for automatic readout as on a tape or card together with visual display of the reading if desired. The arrangement thus generally described provides for fully automatic operation of the scaler or digitizer eliminating the necessity for attendance by a trained operator. As a digitizer the instrument makes possible the change of information from analog to digital form. Another important object of this invention is to describe an automatic method of taking current measurements with an electrometer on a continuous basis. This is possible since the time required for sensitivity measurements is completely eliminated and the dead time between measurements may be a minimum, constant time interval, (less than a second). Thus, all measurements can be made in real time. A further object of this invention is to apply this digitizing, flux-measuring equipment to the control of a nuclear reactor. Not only does such a digitizer afford a means of obtaining reactor flux levels at several places simultaneously in the core lattice, and at frequent, almost continuous intervals, but these readings may be further used to obtain automatic control of the reactor's power output. Another object of this invention is to employ rotating arm or deflecting arm instruments as optical, analog-to-digital converters. In these instruments, the real image of the deflecting arm may be used to derive time interval measurements in an optical slit-phototransducer system; or the moving arm itself may cause the interruption of optical flux incident upon the optical-slit-phototransducer arrangement. The following description and accompanying drawings will more fully describe the purpose of this invention: |
description | This application claims priority to Chinese Patent Application No. 201611052178.6 filed on Nov. 24, 2016, the entire content of which is incorporated herein by reference. The present disclosure relates to an accelerator system and a method of controlling the accelerator system. A medical electronic linear accelerator may include a multi-leaf collimator (MLC) including a plurality of movable leaves. In a process of tumor radiation therapy, the leaves of the multi-leaf collimator may be formed into a specified shape according to a treatment plan so as to implement conformal radiation therapy or intensity modulated radiation therapy. Thus, tumor treatment effect may be improved and ionizing radiation for a subject may be decreased. In the entire process of tumor radiation therapy, the position of the respective leaves of the multi-leaf collimator shall be accurately controlled. When a leaf position is abnormal, the electronic linear accelerator may be damaged relatively easily, and undesirable ionizing radiation may be brought to the subject. NEUSOFT MEDICAL SYSTEMS CO., LTD. (NMS), founded in 1998 with its world headquarters in China, is a leading supplier of medical equipment, medical IT solutions, and healthcare services. NMS supplies medical equipment with a wide portfolio, including CT, Magnetic Resonance Imaging (MRI), digital X-ray machine, ultrasound, Positron Emission Tomography (PET), Linear Accelerator (LINAC), and biochemistry analyser. Currently, NMS' products are exported to over 60 countries and regions around the globe, serving more than 5,000 renowned customers. NMS's latest successful developments, such as 128 Multi-Slice CT Scanner System, Superconducting MRI, LINAC, and PET products, have led China to become a global high-end medical equipment producer. As an integrated supplier with extensive experience in large medical equipment, NMS has been committed to the study of avoiding secondary potential harm caused by excessive X-ray irradiation to the subject during the CT scanning process. FIG. 1 illustrates a diagram of an accelerator system 10 according to an example of the present disclosure. In an example, the accelerator system 10 may include a medical accelerator system, such as, an electronic linear accelerator, which is not limited herein. The accelerator system 10 may include a fixing gantry 11, a rotating gantry 12, a radiation head 13, a multi-leaf collimator 14 and a treatment bed 15. The rotating gantry 12 may be located at one side of the fixing gantry 11. The rotating gantry 12 may be rotated about an axis M of the fixing gantry 11. The radiation head 13 may be connected to a top of the rotating gantry 12 and opposite to the treatment bed 15. The radiation head 13 may include a ray source (not shown in FIG. 1) to emit an imaging ray beam and a treatment ray beam. The imaging ray beam may include an X-ray beam and the treatment ray beam may include a β-ray beam. The multi-leaf collimator 14 may be connected to the radiation head 13. The multi-leaf collimator 14 is a mechanical assembly to produce a conformal radiation field for conformal radiation therapy or intensity modulated radiation therapy. The treatment bed 15 may be used to support a subject 151. The treatment bed 15 may be rotated about an axis N to adjust a position of the subject 151 relative to the radiation head 13, so that the ray beam emitted from the ray source can be irradiated to a particular part of the subject 151. FIG. 2 illustrates a structural diagram of a multi-leaf collimator 14 according to an example of the present disclosure. The multi-leaf collimator 14 includes a fixing support 142 and a plurality of leaves 141 which are disposed on the fixing support 142 and arranged in pairs. Each of the leaves 141 may be driven by a corresponding miniature motor to move. Each of the leaves 141 may be formed into a particular shape at a moment according to a treatment plan (also referred to as a field), for example, an opening, such as a sub-field 143 may be formed in a central region of the multi-leaf collimator 14. The ray beam emitted from the ray source may be irradiated to a part to be examined of the subject after passing through the sub-field 143. FIG. 3 illustrates a module diagram of an accelerator system 10 according to an example of the present disclosure. The accelerator system 10 includes a ray source 16, a multi-leaf collimator 14, a multi-leaf collimator controller 17 and a leaf position determining device 18. The multi-leaf collimator 14 includes two groups of leaves 141 arranged symmetrically. The multi-leaf collimator controller 17 may be configured to control the respective leaves 141 to move according to a predetermined position of the respective leaves 141. The leaf position determining device 18 may be configured to determine a sub-field shape and a sub-field size of the multi-leaf collimator 14, determine an actual position of the respective leaves 141 according to the sub-field shape and the sub-field size, and obtain an error value for the respective leaves by comparing the actual position with the predetermined position for the respective leaves so as to control operation of the accelerator system 10. In an example, the leaf position determining device 18 includes a three-dimensional scanner 181, a reflector 182, an image reconstructing module 183 and a position processing module 184. The three-dimensional scanner 181 may be configured to obtain the three-dimensional image of the multi-leaf collimator 14 by performing a three-dimensional scan on the multi-leaf collimator 14 with an emitted structured light beam. The reflector 182 may be configured to reflect the structured light beam emitted from the three-dimensional scanner 181 to the multi-leaf collimator 14 to perform the three-dimensional scan. In an example, the reflector 182 may be located between the ray source 16 and the multi-leaf collimator 14. A reflecting layer of the reflector 182 may face the sub-field 143 of the multi-leaf collimator 14. The three-dimensional scanner 181 may be configured to obtain a three-dimensional coordinate and a reflected light intensity of each of the leaves 141, generate scanning data according to the three-dimensional coordinate and the reflected light intensity of each of the leaves 141, and output the scanning data to the image reconstructing module 183. The image reconstructing module 183 may be configured to receive the scanning data from the three-dimensional scanner 181 and reconstruct the three-dimensional image of the multi-leaf collimator 14 according to the scanning data. The position processing module 184 may be configured to determine the sub-field shape and the sub-field size of the multi-leaf collimator 14 according to the three-dimensional image, and determine the actual position of each of the leaves 141 according to the sub-field shape and the sub-field size. FIG. 4 illustrates a module diagram of an accelerator system 40 according to another example of the present disclosure. The accelerator system 40 shown in FIG. 4 includes all assemblies of the accelerator system 10 shown in FIG. 3. Compared with the accelerator system 10 shown in FIG. 3, the accelerator system 40 shown in FIG. 4 further includes a predetermined position obtaining module 20 and a system controller 19. In an example, the predetermined position obtaining module 20 may be configured to output the predetermined position of each of the leaves 141 in a sub-field to the multi-leaf collimator controller 17 according to the treatment plan. At this case, the multi-leaf collimator controller 17 may be configured to receive the predetermined position of each of the leaves 141 from the predetermined position obtaining module 20 and control the corresponding motor to drive each of the leaves 141 according to the predetermined position of the each of the leaves 141. In an example, when each of the leaves 141 stops moving, the multi-leaf collimator controller 17 may be configured to notify the system controller 19 that the leaves 141 of the multi-leaf collimator 14 are ready. The leaf position determining device 18 may be configured to determine the sub-field shape and the sub-field size of the multi-leaf collimator 14, determine the actual position of each of the leaves 141 according to the sub-field shape and the sub-field size, and obtain an error value for each of the leaves 141 by comparing the actual position and the predetermined position for each of the leaves 141, so as to control the operation of the accelerator system 10. In an example, the leaf position determining device 18 may be configured to obtain the three-dimensional image of the multi-leaf collimator 14, and determine the sub-field shape and the sub-field size according to the three-dimensional image. The leaf position determining device 18 includes a three-dimensional scanner 181, a reflector 182, an image reconstructing module 183 and a position processing module 184. The three-dimensional scanner 181 may be configured to obtain the three-dimensional image of the multi-leaf collimator 14 by performing a three-dimensional scan on the multi-leaf collimator 14 with an emitted structured light beam. In an example, the three-dimensional scanner 181 may include a three-dimensional laser scanner to emit a laser beam to scan the multi-leaf collimator 14 at a relatively higher scanning speed. In an example, when the system controller 19 receives a notification signal indicating that the respective leaves 141 is ready from the multi-leaf collimator controller 17, it may instruct the three-dimensional scanner 181 to emit the structured light beam for starting scanning the multi-leaf collimator 14. An area of the structured light beam emitted from the three-dimensional scanner 181 passing through the multi-leaf collimator 14 is as the same as an area of the ray beam emitted from the ray source 16 passing through the multi-leaf collimator 14. In other words, to some extent, the position of the three-dimensional scanner 181 may be equivalent to the position of the ray source 16. The structured light beam emitted from the three-dimensional scanner 181 may be reflected to the multi-leaf collimator 14 by the reflector 182 so that the shape and the size of the sub-field 143 may be obtained. In an example, the three-dimensional scanner 181 may be out of an area in which there is the ray beam emitted from the ray source 16, so as to avoid blocking the ray beam. The reflector 182 may be configured to reflect the structured light beam emitted from the three-dimensional scanner 181 to the multi-leaf collimator 14 to perform a three-dimensional scan. The positions of the three-dimensional scanner 181 and the reflector 182 may be adjusted such that the area of the structured light beam emitted from the three-dimensional scanner 181 passing through the multi-leaf collimator 14 after being reflected by the reflector 182 is the same as the area of the ray beam emitted from the ray source 16 passing through the multi-leaf collimator 14. For example, a propagation path of the structured light beam after being reflected by the reflector 182 may be consistent with that of the ray beam after penetrating the reflector 182. In an example, the three-dimensional scanner 181 may be parallel to the multi-leaf collimator 14, and the reflector 182 may be at 45 degrees with respect to an axis of the structured light beam emitted from the three-dimensional scanner 181. A distance from the three-dimensional scanner 181 to the reflector 182 may be equal to a distance from the ray source 16 to the reflector 182, and an area on which the structured light beam emitted from the three-dimensional scanner 181 is irradiated to the reflector 182 may be the same as an area through which the ray beam emitted from the ray source 16 penetrates the reflector 182. In this way, the actual position of each of the leaves 141 may be obtained more accurately. In another example, according to actual demands, the three-dimensional scanner 181 may also be placed at an inclination with respect to the multi-leaf collimator 14, and an angle of the reflector 182 relative to the axis of the structured light beam emitted from the three-dimensional scanner 181 may also be adjusted accordingly. It is noted that the reflecting layer of the reflector 182 may be disposed close to one side of the three-dimensional scanner 181 and face the multi-leaf collimator 14. In this way, the structured light beam emitted from the three-dimensional scanner 181 may be directly reflected by the reflecting layer when arriving at the reflector 182, rather than being refracted when arriving at the reflector 182 in the thickness direction of the reflector 182. The three-dimensional scanner 181 may be configured to obtain the three coordinate and the reflected light intensity of each of the leaves 141, generate scanning data according to the three coordinate and the reflected light intensity of the each of the leaves 141, and output the scanning data to the image reconstructing module 183. The image reconstructing module 183 may be configured to receive the scanning data from the three-dimensional scanner 181, and reconstruct the three-dimensional image of the multi-leaf collimator 14 according to the scanning data. The position processing module 184 may be configured to determine the sub-field shape and the sub-field size shown by the multi-leaf collimator 14 according to the three-dimensional image. For example, the position processing module 184 may be configured to obtain a sectional image of the multi-leaf collimator 14 in a cross-sectional direction of the multi-leaf collimator 14 according to the three-dimensional image of the multi-leaf collimator 14. In an example, as shown in FIG. 5, the position processing module 184 may be configured to generate the sectional image of the multi-leaf collimator 14 in the cross sectional direction along an axis 145 of each of the leaves 141. FIG. 5 illustrates a schematic diagram of a sectional image. The actual multi-leaf collimator 14 may include more than one hundred leaves. The sectional image may show each of the leaves 141 and the sub-field 143 of the multi-leaf collimator 14. A tip 144 of each of the leaves 141 is shown in the sectional image, and the tip 144 of the leaves 141 may be formed into an edge contour of the sub-field 143, e.g. the sub-field shape. In another example, the position processing module 184 may be configured to generate the sectional image of the multi-leaf collimator 14 on a cross section that is parallel to the axis 145 and below the axis 145 (e.g., in the direction away from the ray source 16). The position processing module 184 may be further configured to determine the sub-field shape and the sub-field size from the sectional image. FIG. 6 illustrates a sub-field shape obtained from the sectional image shown in FIG. 5. The sub-field shape is a shape of an actual sub-field formed by each of the leaves 141 after being actually moved. The position processing module 184 may be configured to determine the actual position of each of the leaves 141 according to the sub-field shape and the sub-field size. The position of the tip 144 of each of the leaves 141, e.g., a position to which each of the leaves 141 is controlled by the multi-leaf collimator controller 17 to actually move may be determined according to the sub-field shape and the sub-field size. The position of each of the leaves 141 may be accurately and directly determined according to the actual sub-field shape. The position processing module 184 may be further configured to obtain an error value for each of the leaves 141 by comparing the actual position with the predetermined position for each of the leaves 141 to control the operation of the accelerator system 10. The position processing module 184 may be further configured to obtain the predetermined position of each of the leaves 141 from the predetermined position obtaining module 20, and determine the error value for each of the leaves 141 between the actual position and the corresponding predetermined position for each of the leaves 141. If the error value for each of the leaves 141 is within a threshold range, e.g., the error value corresponding to each of the leaves 141 is no greater than an error threshold, such as, 1 mm, the accelerator system 10 may be operated normally and proceed with treatment. In an example, the position processing module 184 may be configured to transmit instructions indicating that the position of each of the leaves 141 are normal to the system controller 19, and the system controller 19 may be configured to control the accelerator system 10 to operate normally according to the instructions. If the error value for any one of the leaves 141 is out of the threshold range, e.g., at least one error value is greater than the threshold range, the accelerator system 10 may be stopped to terminate the treatment. In an example, the position processing module 184 may be configured to transmit instructions indicating the positions of the leaves 141 are abnormal to the system controller 19, and the system controller 19 may be configured to control the accelerator system 10 to stop according to the instructions. The position processing module 184 may be configured to determine the at least one abnormal error value and the respective leaves 141 corresponding to the at least one abnormal error value for eliminating a failure. In this way, the subject may be prevented from being injured by the undesirable ionizing radiation caused by the failure of the multi-leaf collimator 14. The image reconstructing module 183, the position processing module 184, the predetermined position obtaining module 20, the multi-leaf collimator controller 17 and/or the system controller 19 of the accelerator system 10 may be implemented by software, hardware or a combination thereof. The image reconstructing module 183, the position processing module 184, the predetermined position obtaining module 20, the multi-leaf collimator controller 17 and/or the system controller 19 may be a plurality of independent modules and may also be integrated into one module, for example, functions corresponding to these modules may be implemented on a control software platform of the accelerator system. The accelerator system 10 may further include other assemblies not shown in the drawings, such as a memory, a display and an input device. The system embodiments described above are merely illustrative, where the modules described as separate members may be or not be physically separated, and the members displayed as modules may be or not be physical modules, i.e., may be located in one place, or may be distributed to a plurality of network modules. Part or all of the modules may be selected according to actual requirements to implement the objectives of the solutions in the embodiments. Those of ordinary skill in the art may understand and carry out them without creative work. FIG. 7 illustrates a flow diagram of a method 70 of controlling an accelerator system according to an example of the present disclosure. The flow includes the following blocks 701-706. At block 701, each of the leaves of a multi-leaf collimator in the accelerator system is controlled to move according to a predetermined position of each of the leaves. At block 702, a three-dimensional image of the multi-leaf collimator is obtained. At block 703, a sub-field shape and a sub-field size of the multi-leaf collimator are determined according to the three-dimensional image. At block 704, the actual position of each of the leaves is determined according to the sub-field shape and the sub-field size. At block 705, an error value for each of the leaves is obtained by comparing the actual position with the predetermined position for each of the leaves. At block 706, operation of the accelerator system is controlled according to the error value for each of the leaves. In an example, obtaining the three-dimensional image of the multi-leaf collimator may include obtaining the three-dimensional image of the multi-leaf collimator with a leaf position determining device in the accelerator system, and the leaf position determining device comprises a three-dimensional scanner and a reflector. In an example, obtaining the three-dimensional image of the multi-leaf collimator may include controlling the three-dimensional scanner to emit a structured light beam, reflecting the structured light beam to the multi-leaf collimator by the reflector, collect reflected light information from each of the leaves of the multi-leaf collimator, where the reflected light information comprises a three-dimensional coordinate and a reflected light intensity of each of the leaves of the multi-leaf collimator, and reconstruct the three-dimensional image of the multi-leaf collimator according to the reflected light information. In an example, controlling the operation of the accelerator system according to the error value for each of the leaves may include continuously operating the accelerator system when the error value for each of the leaves is within a preset threshold range, and stopping the accelerator system when the error value for any one of the leaves is out of the threshold range. In an example, a distance for the structured light beam travelling to the reflector is equal to a distance for a ray beam emitted by a ray source in the accelerator system travelling to the reflector. An area on which the structured light beam is irradiated to the reflector is the same as an area through which the ray beam penetrates the reflector. In an example, the reflector is located between the ray source and the multi-leaf collimator. At this case, a reflecting layer of the reflector faces the multi-leaf collimator. In an example, the three-dimensional scanner is parallel to the multi-leaf collimator. The reflector is at 45 degrees with respect to an axis of the structured light beam. In an example, the flow further includes adjusting the position of the three-dimensional scanner such that an area of the structured light beam passing through the multi-leaf collimator is the same as an area of a ray beam emitted from the ray source passing through the multi-leaf collimator. In an example, determining the sub-field shape and the sub-field size of the multi-leaf collimator according to the three-dimensional image may include obtaining a sectional image of the multi-leaf collimator in a cross sectional direction of the multi-leaf collimator according to the three-dimensional image, and determining the sub-field shape and the sub-field size according to the sectional image. The actual position of each of the leaves may be directly determined according to the sub-field shape and the sub-field size by this method. The method can be relatively stable and reliable. The implementation process of the above blocks may be referred to the implementation process corresponding to the above accelerator system 10, which is not described redundantly. Since method embodiments substantially correspond to the system embodiments, reference may be made to a partial description of the system embodiments for related parts. The foregoing description is merely illustrative of embodiments of the present disclosure but not intended to limit the present disclosure, and any modifications, equivalent substitutions, adaptations thereof made within the spirit and principles of the disclosure shall be encompassed in the scope of protection of the present disclosure. |
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description | 1. Field of the Invention The present invention relates to a method of resetting a substrate processing apparatus, a storage medium storing a program for implementing the method, and a substrate processing apparatus, and in particular relates to a method of resetting a substrate processing apparatus after maintenance, and a storage medium storing a program for implementing the method. 2. Description of the Related Art Generally, a substrate processing system that carries out predetermined processing such as film formation or etching on semiconductor wafers (hereinafter referred to as “wafers”) as substrates is comprised of a processing chamber (substrate processing apparatus, hereinafter referred to as “P/C”) in which a wafer is housed and subjected to the predetermined processing, an atmospheric system transferring apparatus that removes wafers from a wafer cassette, which is a sealed container housing a predetermined number of wafers, and a load lock chamber that is disposed between the atmospheric system transferring apparatus and the P/C and transfers wafers in and out between the atmospheric system transferring apparatus and the P/C. In such a substrate processing system, the P/C has a cylindrical chamber (hereinafter referred to as “chamber”), a wafer being subjected to the desired processing such as etching in the chamber using a plasma or the like. However, the plasma in the chamber during the etching does not only etch the wafer, but also causes wear of component parts of the chamber, and furthermore causes production of reaction products such as a deposit. The reaction products become attached to surfaces of component parts, and hence every time a predetermined processing time period has elapsed, it is necessary to open a lid that separates the inside of the chamber from the outside, and carry out maintenance such as replacing worn out component parts in the chamber or cleaning component parts on which reaction products have become attached. Once the maintenance has been completed, the lid is closed, and resetting work involving reducing the pressure in the chamber and so on is carried out on the P/C. During the P/C resetting work, the state of transfer of wafers and the wafer surface etch rate are checked. In the case, for example, that the etch rate exhibits an abnormal value, it is necessary to carry out a chamber leakage check, or open the lid of the chamber and recheck inside the chamber, checking for example whether or not a component part is missing or out of place, whether or not there is a component part installation defect, or whether or not there is a component part cleaning defect. There has thus been a problem that the P/C resetting work takes much time. In recent years, there has thus been developed a method in which, because the state of plasma generation in the chamber becomes unstable if a component part installation defect arises in the chamber, the output of a high-frequency power source that applies high-frequency electrical power in the chamber is monitored so as to detect abnormalities in the P/C. In this method, it is detected whether the high-frequency power source has reached a stable state by comparing results of multivariate analysis on a plurality of types of measured data for the high-frequency power source measured when the state of application of the electrical power by the high-frequency power source has become stable (hereinafter referred to as the “normal model”) with results of multivariate analysis on the same measured data measured upon start-up of the P/C (see, for example, Japanese Laid-open Patent Publication (Kokai) No. 2002-18274). According to this method, mistakes in the installation of component parts in the chamber and so on can be detected without opening the lid of the chamber, and hence compared with conventional P/C resetting work, the time required for resetting the P/C can be reduced. However, according to the method described above, in which the output of the high-frequency power source is monitored so as to detect abnormalities in the P/C, multivariate analysis must be carried out. A method of selecting the measured data to be used in the normal model has not been established, and hence there is a problem that when setting thresholds for comparison, universality of the normal model cannot be secured, and the P/C abnormality judgment cannot be carried out accurately. Moreover, in the multivariate analysis, each piece of measured data is standardized, and hence the multivariate analysis results are not absolute values. When setting a given multivariate analysis result as a threshold, because the multivariate analysis result is not an absolute value, it is difficult for workers to understand the effect that a variation in the measured data will have on the multivariate analysis results. As a result, there is a problem that the workers' subjectivity enters into the setting of thresholds, and hence again the universality of the normal model cannot be secured, and the P/C abnormality judgment cannot be carried out accurately. Furthermore, the environment in the chamber changes between before and after maintenance, and hence the normal model must be reset every time maintenance is carried out. There is a problem that this takes much time, and hence a decrease in the utilization ratio of the P/C still cannot be prevented. It is an object of the present invention to provide a method of resetting a substrate processing apparatus, a storage medium storing a program for implementing the method, and a substrate processing apparatus, which are capable of carrying out abnormality judgment on the substrate processing apparatus accurately without causing a decrease in the utilization ratio of the substrate processing apparatus. To attain the above object, in a first aspect of the present invention, there is provided a method of resetting a substrate processing apparatus having a chamber, the method comprising an evacuating step of evacuating the chamber, a temperature setting step of setting a temperature in the chamber, an abnormality judgment step of judging whether or not there is an abnormality in the chamber, and a seasoning step of stabilizing an atmosphere in the chamber so as to conform to predetermined processing conditions, wherein the abnormality judgment step comprises measuring at least one selected from data that change in response to a change in a state inside the chamber, and comparing the measured data with reference data that corresponds to the measured data for a normal state in the chamber. According to the construction of the first aspect described above, in the abnormality judgment, at least one selected from data that change in response to a change in the state inside the chamber is measured, and the measured data is compared with reference data that corresponds to the measured data for a normal state in the chamber. That is, the abnormality judgment is carried out based on data that changes in response to a change in the state inside the chamber, without using results of multivariate analysis. The abnormality judgment for the substrate processing apparatus can thus be carried out accurately. Moreover, there is no need to reset a normal model every time maintenance is carried out. The utilization ratio of the substrate processing apparatus can thus be prevented from decreasing. Preferably, in the temperature setting step, the temperature in the chamber is set to a temperature different to a temperature in the chamber during ordinary substrate processing. According to the construction of the first aspect described above, in the temperature setting, the temperature in the chamber is set to a temperature different to the temperature in the chamber during ordinary substrate processing. As a result, the atmosphere in the chamber can be made to conform to processing conditions enabling only the minimum substrate processing accuracy required in the resetting of the substrate processing apparatus to be secured. The resetting of the substrate processing apparatus can thus be carried out quickly. Preferably, in the abnormality judgment step, a processing gas that does not cause production of a reaction product in the chamber is introduced while a substrate in the chamber is subjected to a predetermined process. According to the construction of the first aspect described above, in the abnormality judgment, a processing gas that does not cause production of a reaction product in the chamber is introduced during substrate processing. As a result, a reaction product is not deposited in the chamber during resetting of the substrate processing apparatus, and hence transition to ordinary substrate processing after the resetting of the substrate processing apparatus can be carried out smoothly. More preferably, the processing gas comprises only oxygen. According to the construction of the first aspect described above, the gas introduced into the chamber during the abnormality judgment comprises only oxygen. As a result, production of a reaction product in the chamber can be prevented reliably. Preferably, the measured data comprises a log showing a state of at least one component part of the substrate processing apparatus. According to the construction of the first aspect described above, the measured data comprises a log showing a state of at least one component part of the substrate processing apparatus. As a result, the measured data can be obtained simultaneously with the processing, and hence the abnormality judgment can be carried out quickly. More preferably, the log is of an impedance of a matcher that adjusts high-frequency electrical power applied to a lower electrode disposed in the chamber. According to the construction of the first aspect described above, the log used as the measured data is of the impedance of a matcher that adjusts high-frequency electrical power applied to a lower electrode disposed in the chamber. The impedance of the matcher changes in accordance with the state of plasma generation in the chamber. The abnormality judgment for the substrate processing apparatus can thus be carried out reliably. Alternatively, the log is of a voltage between a lower electrode disposed in the chamber and a matcher that adjusts high-frequency electrical power applied to the lower electrode. According to the construction of the first aspect described above, the log used as the measured data is of the voltage between a lower electrode disposed in the chamber and a matcher that adjusts high-frequency electrical power applied to the lower electrode. The voltage changes in accordance with the state of plasma generation in the chamber. The abnormality judgment for the substrate processing apparatus can thus be carried out reliably. Alternatively, the log is of an opening extent of a control valve that controls a pressure in the chamber. According to the construction of the first aspect described above, the log used as the measured data is of the opening extent of a control valve that controls the pressure in the chamber. The opening extent changes in accordance with the state of plasma generation in the chamber. The abnormality judgment for the substrate processing apparatus can thus be carried out reliably. Preferably, the measured data is processed substrate light emission data. According to the construction of the first aspect described above, the measured data is processed substrate light emission data. The light emission data can be measured easily, and moreover changes in accordance with the state of plasma generation in the chamber. The abnormality judgment for the substrate processing apparatus can thus be carried out reliably, without causing a decrease in the utilization ratio of the substrate processing apparatus. More preferably, the light emission data relates to a light intensity ratio. According to the construction of the first aspect described above, the light emission data used as the measured data relates to a light intensity ratio. The light intensity ratio is dimensionless, and hence there is no effect from the magnitude of light intensities used in calculation of the light intensity ratio. The abnormality judgment for the substrate processing apparatus can thus be carried out accurately. Preferably, the measured data is data relating to a high-frequency power source that supplies high-frequency electrical power applied to a lower electrode disposed in the chamber. According to the construction of the first aspect described above, the measured data is data relating to a high-frequency power source that supplies high-frequency electrical power applied to a lower electrode disposed in the chamber. As a result, the data can be measured easily, and hence the abnormality judgment can be carried out easily. Preferably, in the evacuating step, the temperature in the chamber is raised. According to the construction of the first aspect described above, in the evacuation, the temperature in the chamber is raised. By raising the temperature in the chamber, evaporation of moisture that has become attached to component parts of the chamber is promoted. The resetting of the substrate processing apparatus can thus be carried out quickly. Preferably, in the seasoning step, stability of the atmosphere in the chamber is detected based on a difference in light emission data between two consecutively processed substrates. According to the construction of the first aspect described above, in the seasoning, the stability of the atmosphere in the chamber is detected based on the difference in light emission data between two consecutively processed substrates. Once the atmosphere in the chamber has become stable, the amount of light emitted from the substrates also stops varying and becomes stable. Judgment of the stability of the atmosphere in the chamber can thus be carried out easily based on the difference in light emission data between two consecutively processed substrates as described above. More preferably, in the seasoning step, the stability of the atmosphere in the chamber is detected based on a derivative of the difference in the light emission data. According to the construction of the first aspect described above, in the seasoning, the stability of the atmosphere in the chamber is detected based on the derivative of the difference in the light emission data between two consecutively processed substrates. The derivative of the difference in the light emission data can decrease affection by the magnitude of the variation of the light intensity. The judgment of the stability of the atmosphere in the chamber can thus be carried out more accurately. Preferably, in the abnormality judgment step, a leak in the chamber is detected based on a ratio of light emission amounts at different wavelengths for light emission from a processed substrate. According to the construction of the first aspect described above, in the abnormality judgment, a leak in the chamber is detected based on a ratio of light emission amounts at different wavelengths for light emission from a processed substrate. If a leak occurs in the chamber, then gas flows in from the outside, and the state of light emission by the plasma changes in accordance with the type of the gas. Moreover, the ratio of light emission amount is dimensionless, and hence there is no effect from the magnitude of the light emission amount at each of the wavelengths used in calculation of the ratio of light emission amount. The abnormality judgment for the substrate processing apparatus can thus be carried out more accurately. To attain the above object, in a second aspect of the present invention, there is provided a computer-readable storage medium storing a program for causing a computer to implement a method of resetting a substrate processing apparatus having a chamber, the program comprising an evacuating module for evacuating the chamber, a temperature setting module for setting a temperature in the chamber, an abnormality judgment module for judging whether or not there is an abnormality in the chamber, and a seasoning module for stabilizing an atmosphere in the chamber so as to conform to predetermined processing conditions, wherein the abnormality judgment module measures at least one selected from data that change in response to a change in a state inside the chamber, and compares the measured data with reference data that corresponds to the measured data for a normal state in the chamber. To attain the above object, in a third aspect of the present invention, there is provided a substrate processing apparatus comprising a chamber, an evacuating device that evacuates the chamber, a temperature setting device that sets a temperature in the chamber, an abnormality judgment device that judges whether or not there is an abnormality in the chamber, and a seasoning device that stabilizes an atmosphere in the chamber so as to conform to predetermined processing conditions, wherein the abnormality judgment device measures at least one selected from data that change in response to a change in a state inside the chamber, and compares the measured data with reference data that corresponds to the measured data for a normal state in the chamber. The above and other objects, features, and advantages of the invention will become more apparent from the following detailed description taken in conjunction with the accompanying drawings. The present invention will now be described in detail with reference to the drawings showing preferred embodiments thereof. First, a substrate processing apparatus according to an embodiment of the present invention will be described. FIG. 1 is a sectional view schematically showing the construction of a processing chamber, which is the substrate processing apparatus according to the present embodiment. As shown in FIG. 1, the processing chamber (hereinafter referred to as “P/C”) 2, which is constructed as an etching apparatus that subjects semiconductor wafers to etching, has a cylindrical chamber 10 made of a metal such as aluminum or stainless steel. A lower electrode 11 is disposed in the chamber 10 as a stage on which is mounted a semiconductor wafer having a diameter of, for example, 200 mm. An exhaust path 12 that acts as a flow path through which gas above, the lower electrode 11 is exhausted to the outside of the chamber 10 is formed between a side wall of the chamber 10 and the lower electrode 11. An annular evacuation plate (partitioning plate) 13 is disposed part way along the exhaust path 12, and a space in the exhaust path 12 downstream of the evacuation plate 13 is communicated with an automatic pressure control valve (hereinafter referred to as the “APC valve”), not shown, which is a variable butterfly valve. The APC valve is connected to a turbo-molecular pump (TMP) and a dry pump (DP), which are exhausting pumps for evacuation. The APC valve is used for controlling the pressure in the chamber 10, and also for reducing the pressure in the chamber 10 down to a substantially vacuum state using the TMP and the DP. A lower high-frequency power source 18 is connected to the lower electrode 11 via a lower matcher 19. The lower high-frequency power source 18 applies predetermined high-frequency electrical power to the lower electrode 11. The lower matcher 19 reduces reflection of the high-frequency electrical power from the lower electrode 11 so as to maximize the efficiency of input of the high-frequency electrical power into the lower electrode 11. An electrostatic chuck (hereinafter referred to as “ESC”) 20 for attracting the semiconductor wafer by electrostatic attraction is disposed in an upper portion of the lower electrode 11. A DC power source (not shown) is electrically connected to the ESC 20. The semiconductor wafer is attracted to and held on an upper surface of the lower electrode 11 through a Johnsen-Rahbek force or a Coulomb force generated by a DC voltage applied to the ESC 20 from the DC power source. Moreover, an annular focus ring 24 made of silicon (Si) or the like is disposed at a periphery of the upper portion of the lower electrode 11. The focus ring 24 focuses a plasma generated above, the lower electrode 11 toward the semiconductor wafer. Moreover, the focus ring 24 is surrounded by an annular cover ring 14 a surface of which is coated with Y2O3 or the like. An annular coolant chamber 25 that extends, for example, in a circumferential direction of the lower electrode 11 is provided inside the lower electrode 11. A coolant, for example cooling water, at a predetermined temperature is circulated through the coolant chamber 25 via piping 26 from a chiller unit (not shown). A processing temperature of the semiconductor wafer on the lower electrode 11 and the temperature of the lower electrode 11 are controlled through the temperature of the coolant. A support 15 provided so as to extend from a lower portion of the lower electrode 11 downward is disposed below the lower electrode 11. The support 15 supports the lower electrode 11. The lower electrode 11 is raised and lowered by rotating a ball screw, not shown. Moreover, the support 15 is surrounded by a bellows cover 16 so as to be cut off from the atmosphere in the chamber 10. A shower head 33 is disposed in a ceiling portion of the chamber 10. The shower head 33 has a disk-shaped upper electrode plate (CEL(ceiling electrode)) 35 that faces into the chamber 10 and has therein a large number of gas-passing holes 34, and an electrode support 36 that is disposed above, the upper electrode plate 35 and on which the upper electrode plate 35 is detachably supported. According to the P/C 2, when a semiconductor wafer is transferred into or out from the chamber 10, the lower electrode 11 is lowered down to a semiconductor wafer transferring in/out position; when the semiconductor wafer is being etched, the lower electrode 11 is raised up to a semiconductor wafer processing position, and moreover the ESC 20 attracts and thus holds the semiconductor wafer. An upper high-frequency power source 17 is connected to the upper electrode plate 35 via an upper matcher 21. The upper high-frequency power source 17 applies predetermined high-frequency electrical power to the upper electrode plate 35. The upper matcher 21 reduces reflection of the high-frequency electrical power from the upper electrode plate 35 so as to maximize the efficiency of input of the high-frequency electrical power into the upper electrode plate 35. A buffer chamber 37 is provided inside the electrode support 36. A processing gas introducing pipe 38 is connected from a processing gas supply apparatus 40, described below, to the buffer chamber 37. Moreover, a coolant chamber (not shown) that is connected to a chiller unit (not shown) is disposed in a lower portion of the electrode support 36. The electrode support 36 thus functions as a cooling plate for the upper electrode plate 35, whereby the temperature of the upper electrode plate 35 is controlled. A valve 41 is disposed part way along the processing gas introducing pipe 38, and the processing gas supply apparatus 40 is disposed upstream of the valve 41. The processing gas supply apparatus 40 has a silicon tetrafluoride (SiF4) supply unit 42 that supplies silicon tetrafluoride, a silicon tetrahydride (SiH4) supply unit 43 that supplies silicon tetrahydride, an oxygen gas (O2) supply unit 44 that supplies oxygen gas, an argon gas (Ar) supply unit 45 that supplies argon gas, and MFC's (mass flow controllers) 46 to 49 provided in correspondence with the supply units 42 to 45 respectively. The processing gas supply apparatus 40 mixes together silicon tetrafluoride, silicon tetrahydride, oxygen gas, and argon gas in predetermined flow rate proportions, and supplies the resulting mixture into the chamber 10 as a processing gas. At this time, the MFC's 46 to 49 control the flow rates of the respective components of the processing gas, and thus, in collaboration with the APC valve, control the pressure in the chamber 10 to a desired value. The processing gas supplied by the processing gas supply apparatus 40 may also have carbon tetrafluoride (CF4) mixed therein. An annular holder 54 that holds the shower head 33 is disposed surrounding the shower head 33. A shield ring 55 that protects a gap between the holder 54 and the shower head 33 from the plasma, described below, is disposed on a lower surface of the holder 54. Moreover, a cylindrical flow-adjusting exhaust ring 56 that projects out downward is disposed at a periphery of the shield ring 55. The flow-adjusting exhaust ring 56 prevents diffusion of the plasma generated in a space between the lower electrode 11 and the upper electrode plate 35, confining the plasma within this space. Component parts in the chamber 10 such as the upper electrode plate 35 and the shield ring 55 described above are assembled in the chamber 10 using screws 60 as shown in FIG. 2. As shown in FIG. 2, the screws 60 are each cut off from the atmosphere in the chamber 10 by a screw cover 58 or a screw cap (not shown). Two transmission windows 50 and 51 made of quartz glass are disposed in the side wall of the chamber 10 either side of the lower electrode 11 at a height corresponding to the semiconductor wafer processing position. Moreover, a light source 52 that emits laser light is disposed on the opposite side of the transmission window 50 to the chamber 10, and a light-receiving sensor 53 that receives laser light that has been emitted from the light source 52 and passed through the chamber 10 is disposed on the opposite side of the transmission window 51 to the chamber 10. Here, the flow-adjusting exhaust ring 56 described above has slits formed in portions thereof facing respectively the light source 52 and the light-receiving sensor 53, so that the laser light emitted from the light source 52 passes through one slit in the flow-adjusting exhaust ring 56, passes over the semiconductor wafer mounted on the lower electrode 11, passes through the other slit in the flow-adjusting exhaust ring 56, and reaches the light-receiving sensor 53. Moreover, a heater (not shown) is built into the side wall of the chamber 10. The heater controls the temperature of the side wall when the semiconductor wafer is being subjected to plasma processing. When the semiconductor wafer is being subjected to the plasma processing, due to the plasma, the atmosphere above, the semiconductor wafer emits light in accordance with the concentration and pressure of the processing gas. The state of plasma generation can thus be monitored by measuring the state of emission of light. According to the P/C 2, laser light passes over the semiconductor wafer as described above, and hence the state of plasma generation can be monitored using the light source 52 and the light-receiving sensor 53. In the chamber 10 of the P/C 2, high-frequency electrical power is applied to the lower electrode 11 and the upper electrode plate 35 as described above. A high-density plasma is generated from the processing gas in the space between the lower electrode 11 and the upper electrode plate 35 by the applied high-frequency electrical power, whereby ions and radicals are produced. The produced ions and radicals are focused onto the surface of the semiconductor wafer by the focus ring 24, whereby the surface of the semiconductor wafer is physically and chemically etched. Furthermore, the P/C 2 has a control unit 57 that controls operation of the component parts, and records the state of operation of the component parts as an apparatus log. When, for example, a semiconductor wafer is subjected to the plasma processing, the control unit 57 controls the heater in the side wall of the chamber and the chiller unit so as to control the temperature in the chamber 10 to a desired temperature, and furthermore controls operation of the processing gas supply apparatus 40, the upper high-frequency power source 17 and the lower high-frequency power source 18 so as to generate a desired amount of plasma between the lower electrode 11 and the upper electrode plate 35, this being in accordance with a recipe that indicates the processing procedure. Moreover, when a semiconductor wafer is subjected to the plasma processing, the control unit 57 records the impedance of the lower matcher 19 and the upper matcher 21, the voltage between the lower electrode 11 and the lower matcher 19 (Vpp), or the opening extent of the APC valve (APC angle) as an apparatus log, and furthermore records the state of emission of light from the atmosphere above, the semiconductor wafer as measured using the light source 52 and the light-receiving sensor 53 (light emission data), the current and voltage of the lower high-frequency power source 18 and so on as measured data. Next, a method of resetting the substrate processing apparatus according to the present embodiment will be described. FIG. 3 is a flowchart of an automatic setup process, which is the method of resetting the substrate processing apparatus according to the present embodiment. The process is carried out by the control unit 57 in accordance with a program, described below, after maintenance of the P/C 2 as described above. After the automatic setup process, the P/C 2 can carry out predetermined etching on semiconductor wafers for mass production in accordance with a predetermined recipe. As shown in FIG. 3, first, the APC valve is opened, and the pressure in the chamber 10 is reduced down to a substantially vacuum state using the TMP and the DP (evacuating step) (step S31), then the heater in the side wall of the chamber and the chiller unit control the inside of the chamber 10 to a predetermined temperature (temperature setting step) (step S32), and then an APC server, described below, judges whether or not there is an abnormality in the P/C 2, for example whether or not a component part is missing or out of place, or whether or not there is a leak in the chamber 10, based on measured data or an apparatus log for the component parts (abnormality judgment step) (step S33). After that, a plurality of dummy wafers are processed, whereby stabilization is carried out such that the chiller unit, the heater, the APC valve, the processing gas supply apparatus 40, the lower high-frequency power source 18, the upper high-frequency power source 17 and so on set the atmosphere in the chamber 10 to conform to predetermined processing conditions as stipulated in a predetermined recipe (seasoning step) (step S34). The APC server then indicates the state with regard to abnormalities of the P/C 2, leakage in the chamber 10 and so on a monitor (not shown) of the P/C 2 or the APC server. A worker can check the details displayed on the monitor, and decide whether to carry out the predetermined processing on semiconductor wafers for mass production forthwith, or whether to suspend the resetting work and open the lid of the chamber 10 and carry out maintenance once again. According to the automatic setup process shown in FIG. 3, a worker can detect an abnormality of the P/C 2 using the APC server without opening the lid of the chamber 10, and hence, for example, resetting work that has conventionally taken 3 hours can be carried out in 2 hours, i.e. the resetting work of the P/C 2 can be carried out quickly, and thus the utilization ratio of the P/C 2 can be improved. The respective steps of the automatic setup process shown in FIG. 3 will now be described in detail. First, the evacuating step of step S31 will be described. During the evacuation of the chamber in a conventional resetting process, the temperature in the chamber has been set to be the same as the temperature during the etching of semiconductor wafers for mass production (hereinafter referred to as the “mass production etching”, and hence moisture that has got into the chamber and become attached to component parts during maintenance, or alcohol that has become attached to component parts upon wet cleaning has evaporated gradually, and as a result the evacuation has taken much time. In contrast with this, in step S31 of the automatic setup process according to the present invention, the temperature in the chamber 10 is set to be higher than the temperature during the mass production etching. Specifically, during the evacuation, conventionally the temperatures of the upper electrode, the side wall of the chamber, and the lower electrode have been set to be constant at 60, 50, and 40° C. respectively, whereas in step S31 of the automatic setup process according to the present invention, using the heater in the side wall of the chamber and the chiller unit, the temperatures of the upper electrode, the side wall of the chamber, and the lower electrode are first set to 80, 80, and 40° C. respectively, and then after 1.5 hours has elapsed, are reset to 60, 50, and 40° C. respectively. According to the evacuation of step S31, the temperature in the chamber 10 is made to be higher than conventionally. By raising the temperature in the chamber 10, evaporation of moisture or alcohol attached to component parts of the chamber 10 is promoted. The resetting of the P/C 2 can thus be carried out quickly. FIG. 4 is a graph showing a comparison of the evacuating time period for the automatic setup process shown in FIG. 3 and a conventional resetting process. In FIG. 4, the leak rate (pressure release rate) from the inside of the chamber for the conventional resetting process is shown by a dashed line, whereas the leak rate from the inside of the chamber 10 for the automatic setup process of the present invention is shown by a full line. As shown in FIG. 4, to reach 0.13 Pa/min (1 mTorr/min), which is a threshold indicating that the evacuation has been completed, 3.5 hours is required for the conventional resetting process, whereas only approximately 3 hours is required for the automatic setup process of the present invention. According to the automatic setup process of the present invention, the resetting of the P/C 2 can thus be carried out quickly, and hence a decrease in the utilization ratio of the P/C 2 can be prevented. In step S31 described above, the temperatures of the upper electrode, the side wall of the chamber, and the lower electrode are controlled using the heater in the side wall of the chamber and the chiller unit. However, the temperature in the chamber may instead be raised using only the heater in the side wall of the chamber, or the by applying high-output high-frequency electrical power to the lower electrode 11 or the upper electrode plate 35 (high-power seasoning). Next, the chamber temperature setting step of step S32 will be described. In a conventional resetting process, although the accuracy of the etching in the automatic setup process may be lower than the accuracy in the mass production etching, regarding the setting of the temperature in the chamber, the dummy wafer etching has been carried out under the same processing conditions as for the mass production etching, and as a result the dummy wafer etching has been carried out at an etch rate lower than necessary. The dummy wafer etching in the automatic setup process of the conventional resetting process has thus required much time. In contrast with this, in step S32 of the automatic setup process according to the present invention, to set the etch rate to be fast while securing the minimum required etching accuracy, the dummy wafer etching is carried out under special processing conditions (a special recipe), for example with the temperature in the chamber 10 set to a temperature different to the temperature in the chamber in the conventional resetting process. Specifically, in step S32, the pressure in the chamber 10 is set to 6.67 Pa (50 mT), the flow rate of oxygen gas supplied by the processing gas supply apparatus 40 is set of 600 sccm, the pressure of helium (He) gas used as a heat-transmitting gas supplied from the ESC 20 toward a rear surface of the dummy wafer is set to 6.67×102 Pa (5 torr) at a central portion of the rear surface and 3.33×103 Pa (25 torr) at a peripheral portion of the rear surface, and the temperatures of the upper electrode, the side wall of the chamber, and the lower electrode are set to 60, 50, and 40° C. respectively. According to the chamber temperature setting of step S32, based on a special recipe exclusively for the automatic setup process, the atmosphere in the chamber 10 is made to conform to processing conditions enabling the minimum required etching accuracy in the automatic setup process to be secured, and the dummy wafer etching is carried out at a faster etch rate than conventionally. The resetting of the P/C 2 can thus be carried out quickly. Moreover, generally, in the case of carrying out mass production etching in a plurality of P/C's, a different recipe is used for each P/C in accordance with differences between the P/C's. However, according to the automatic setup process of the present invention, because a special recipe exclusively for the automatic setup process is set, in the case of carrying out the automatic setup process in a plurality of P/C's, the same special recipe is used for each P/C. As a result, light emission data or data relating to the high-frequency power source can be measured under the same conditions for each P/C, and hence the abnormality judgment, described below, can be carried out stably. Next, the abnormality judgment step of step S33 will be described. In a conventional resetting process, in the abnormality judgment, the dummy wafer etching is carried out with the same processing gas as in the mass production etching, for example silicon tetrafluoride, introduced into the chamber. However, when silicon tetrafluoride is converted into a plasma, reaction products are readily produced, and hence deposit accumulates on surfaces of component parts of the chamber during the resetting process, which causes the production of particles during the mass production etching. In contrast with this, in step S33 of the automatic setup process according to the present invention, to prevent production of such a reaction product, a special processing gas exclusively for the automatic setup process is used, for example oxygen gas, which does not produce reaction products upon being converted into a plasma. According to the abnormality judgment of step S33 described above, when carrying out the dummy wafer etching, oxygen gas, which does not produce reaction products, is introduced into the chamber 10 as the processing gas. Deposit thus does not accumulate in the chamber 10 during the resetting of the P/C 2, and hence transition to mass production etching after the resetting of the P/C 2 can be carried out smoothly. Moreover, in the abnormality judgment step, the APC (advanced process control) server, which is an external controller connected to the P/C 2, collects measured data or an apparatus log recorded by the control unit 57, and detects component part installation defects, component part cleaning defects, and leak in the chamber 10, as described below, based on the collected apparatus log or measured data. Here, component part installation defects include forgetting to install a component part, a component part being missing or out of place, a component part being installed in the wrong position, and so on. First, detection of a component part installation defect in the abnormality judgment step will be described. As described above, if a component part installation defect arises in the chamber, then the state of plasma generation in the chamber becomes unstable. When carrying out the dummy wafer etching, a component part installation defect can thus be detected by monitoring the state of plasma generation in the chamber. In the present embodiment, a component part installation defect is detected based on an apparatus log or measured data that is affected by the state of plasma generation in particular. In the case of detecting a component part installation defect based on an apparatus log, the apparatus log is recorded by the control unit 57 while the dummy wafer etching is carried out, and hence the APC server or the like can obtain the apparatus log simultaneously with the dummy wafer etching being carried out. The abnormality judgment can thus be carried out quickly. Apparatus logs that can be used in component part installation defect detection will now be described in detail. First, the case that the impedance of the lower matcher 19 is used for an apparatus log for component part installation defect detection will be described. When the state of plasma generation in the chamber changes, the high-frequency electrical power applied from the lower high-frequency power source 18 changes so as to maintain the plasma in a desired state, and accompanying this the impedance of the lower matcher 19 changes. A change in the state of plasma generation in the chamber, and hence a component part installation defect, can thus be detected by monitoring the impedance of the lower matcher 19. FIG. 5 is a graph showing the relationship between the presence/absence of a component part installation defect and the impedance of the lower matcher 19. In FIG. 5, the axis of abscissas shows the lot number for which the automatic setup process was carried out, and the axis of ordinates shows the measured value of the impedance for each lot. The measured value of the impedance varies over each lot because 25 dummy wafers were etched in each lot and the measured value of the impedance was plotted on the graph for each wafer in each lot. Each measured value of the impedance is the average value during the etching of the dummy wafer in question. The state of installation of the component parts for each lot was as shown in Table 1 below. TABLE 1Lot No.Details1Reference2No flow-adjusting exhaust ring3Reference4No screw cover5Reference6No shield ring7Reference8No cover ring9Reference10No screw cap (ESC side)11Reference12No cover ring or focus ring13Reference14No screw cap (cooling plate side) For example, in lot 2, the dummy wafer etching was carried out with the flow-adjusting exhaust ring 56 removed. Moreover, for the reference of lot 3 and so on, all of the component parts were installed properly. Thresholds for the component part installation defect judgment were set based on the measured value of the impedance for these reference lots. As shown in FIG. 5, the maximum value of the measured value of the impedance for the reference lots was set as an upper threshold (reference value (max)), and the minimum value of the measured value of the impedance for the reference lots was set as a lower threshold (reference value (min)). As shown in FIG. 5, the measured value of the impedance for lot 2 clearly exceeded the upper threshold. The flow-adjusting exhaust ring 56 being missing or out of place can thus be detected by monitoring the impedance. In FIG. 5 described above, the average value during the etching of each dummy wafer is used as the measured value of the impedance. However, the minimum value or the maximum value may be used instead, in which case a component part other than the flow-adjusting exhaust ring 56 being missing or out of place can also be detected. Moreover, the impedance of the upper matcher 21 also changes in accordance with the state of plasma generation in the chamber, and hence a component part being missing or out of place can also be detected using the measured value of the impedance of the upper matcher 21. According to the abnormality judgment of step S33 described above, the impedance of the lower matcher 19 is used for an apparatus log for component part installation defect detection. The impedance of the lower matcher 19 changes in accordance with changes in the state of plasma generation in the chamber 10, and hence changes in accordance with whether or not a component part is missing or out of place. Detection of a component part being missing or out of place in the P/C 2 can thus be carried out reliably. Next, the case that the voltage between the lower electrode 11 and the lower matcher 19 (Vpp) (hereinafter referred to as the “lower voltage”) is used for an apparatus log for component part installation defect detection will be described. When the state of plasma generation in the chamber changes, the high-frequency electrical power applied from the lower high-frequency power source 18 changes so as to maintain the plasma in a desired state, and accompanying this the lower voltage changes. A change it the state of plasma generation in the chamber, and hence a component part installation defect, can thus be detected by monitoring the lower voltage. FIG. 6 is a graph showing the relationship between the presence/absence of a component part installation defect and the voltage between the lower electrode and the lower matcher (the lower voltage). In FIG. 6, the axis of abscissas shows the lot number for which the automatic setup process was carried out, and the axis of ordinates shows the measured value of the lower voltage for each lot. Moreover, the state of installation of the component parts for each lot was as shown in Table 1 described above as for FIG. 5. The lower voltage varies over each lot for the same reason as for FIG. 5. In FIG. 6, as for FIG. 5, each measured value of the lower voltage is the average value during the etching of the dummy wafer in question. Moreover, as shown in FIG. 6, the maximum value of the measured value of the lower voltage for the reference lots was set as an upper threshold (reference value (max)), and the minimum value of the measured value of the lower voltage for the reference lots was set as a lower threshold (reference value (min)). As shown in FIG. 6, the measured value of the lower voltage for lot 12 clearly exceeded the upper threshold, and furthermore the measured values of the lower voltage for lot 8 and lot 14 clearly dropped below the lower threshold. The cover ring 14 and the focus ring 24 being missing or out of place, the cover ring 14 only being missing or out of place, or the screw cap on the cooling plate side being missing or out of place can thus be detected by monitoring the lower voltage. In FIG. 6 described above, the average value during the etching of each dummy wafer is used as the measured value of the lower voltage. However, the minimum value or the maximum value may be used instead, in which case a component part other than the above component parts being missing or out of place can also be detected. Moreover, the voltage between the upper electrode plate 35 and the upper matcher 21 also changes in accordance with the state of plasma generation in the chamber, and hence a component part being missing or out of place can also be detected using the measured value of the voltage between the upper electrode plate 35 and the upper matcher 21. According to the abnormality judgment of step S33 described above, the lower voltage is used for an apparatus log for component part installation defect detection. The lower voltage changes in accordance with changes in the state of plasma generation in the chamber 10, and hence changes in accordance with whether or not a component part is missing or out of place. Detection of a component part being missing or out of place in the P/C 2 can thus be carried out reliably. Next, the case that the opening extent of the APC valve (the APC angle) is used for an apparatus log for component part installation defect detection will be described. When the state of plasma generation in the chamber changes, the pressure in the chamber 10 must be changed to maintain the plasma in a desired state, and accompanying this the opening extent of the APC valve changes. A change in the state of plasma generation in the chamber, and hence a component part installation defect, can thus be detected by monitoring the opening extent of the APC valve. FIG. 7 is a graph showing the relationship between the presence/absence of a component part installation defect and the opening extent of the APC valve. In FIG. 7, the axis of abscissas shows the lot number for which the automatic setup process was carried out, and the axis of ordinates shows the measured value of the opening extent of the APC valve for each lot. Moreover, the state of installation of the component parts for each lot was as shown in Table 1 described above as for FIG. 5. The opening extent of the APC valve varies over each lot for the same reason as for FIG. 5. In FIG. 7, as for FIG. 5, each measured value of the opening extent of the APC valve is the average value during the etching of the dummy wafer in question. Moreover, as shown in FIG. 7, the maximum value of the measured value of the opening extent of the APC valve for the reference lots was set as an upper threshold (reference value (max)), and the minimum value of the measured value of the opening extent of the APC valve for the reference lots was set as a lower threshold (reference value (min)). As shown in FIG. 7, the measured value of the opening extent of the APC valve for lot 2 clearly exceeded the upper threshold. The flow-adjusting exhaust ring 56 being missing or out of place can thus be detected by monitoring the opening extent of the APC valve. In FIG. 7 described above, the average value during the etching of each dummy wafer is used as the measured value of the opening extent of the APC valve. However, the minimum value or the maximum value may be used instead, in which case a component part other than the above component parts being missing or out of place can also be detected. According to the abnormality judgment of step S33 described above, the opening extent of the APC valve is used for an apparatus log for component part installation defect detection. The opening extent of the APC valve changes in accordance with changes in the state of plasma generation in the chamber 10, and hence changes in accordance with whether or not a component part is missing or out of place. Detection of a component part being missing or out of place in the P/C 2 can thus be carried out reliably. In the abnormality judgment of step S33 described above, the cases that the impedance of the lower matcher 19, the voltage between the lower electrode 11 and the lower matcher 19, and the opening extent of the APC valve are used for apparatus logs for component part installation defect detection have been described. However, apparatus logs that can be used in the abnormality judgment are not limited thereto, but rather any apparatus log that changes in accordance with changes in the state of plasma generation in the chamber can be used. The state of plasma generation in the chamber also changes in accordance with the type of the P/C, and hence it is preferable to investigate the relationship between the state of plasma generation in the chamber and component parts being missing or out of place for each type of P/C in advance, and select the apparatus log to be used for component part installation defect detection based on the results of the investigation. On the other hand, in the case of detecting a component part installation defect based on measured data such as data on light emission above, the semiconductor wafer or data relating to a high-frequency power source, the measured data can easily be measured by the control unit 57 or the like during dummy wafer etching, and hence the APC server or the like can obtain the measured data easily. The abnormality judgment can thus be carried out easily. Measured data that can be used in component part installation defect detection will now be described in detail. First, the case that data on light emission above, the semiconductor wafer is used as measured data for component part installation defect detection will be described. If a component part installation defect arises in the chamber, then the state of plasma generation in the chamber changes. For example, if the focus ring 24 is missing or out of place, then the plasma is not focused over the semiconductor wafer; if the flow-adjusting exhaust ring 56 is missing or out of place, then the plasma diffuses out from the space between the lower electrode 11 and the upper electrode plate 35. When the state of plasma generation in the chamber changes, the dummy wafer light emission data also changes. A change in the state of plasma generation in the chamber, and hence a component part installation defect, can thus be detected by monitoring the light emission data. FIG. 8 is a graph showing the relationship between the presence/absence of a component part installation defect and light emission data. In FIG. 8, the axis of abscissas shows the component part that is missing, and the axis of ordinates shows the ratio between the light emission data when the component part is missing (hereinafter referred to as the “test data”) and the light emission data for a state in which all of the component parts have been installed properly (a reference state) (hereinafter referred to as the “reference data”). First, the ratio between the test data and the reference data (hereinafter referred to as the “light intensity ratio”) will be described. First, the reference data is measured over a predetermined wavelength region, for example 200 to 800 nm, and the test data is also measured over the predetermined wavelength region, for example 200 to 800 nm. Next, the ratio Ai of the test data to the reference data at each wavelength i is calculated as shown in equation (1) below.Ai=ai (test data)/ai (reference data),i=200 to 800 nm (1) The average value Aave of the calculated Ai's is then calculated. Next, as shown in equation (2) below, the absolute value of the difference between Ai and Aave at each wavelength i is calculated, and the sum B of the calculated absolute values of the differences is calculated. The calculated sum B is taken as the light intensity ratio in FIG. 8. B = ∑ i = 200 800 A i - A ave ( 2 ) Next, setting of thresholds shown in FIG. 8 will be described. First, n dummy wafers or the like are prepared, and the reference data is measured for each dummy wafer in the reference state. Then, one measured test data is selected as the reference data in equation (1) above and another measured reference data as the reference data in equation (1) from the measured reference data for the n dummy wafers. Next, the ratio Ai is calculated at each wavelength i in accordance with equation (1) above, and then the light intensity ratio Bk is calculated in accordance with equation (2) above. This sequence of calculations described above is repeated n times, whereby n light intensity ratios B are calculated. Next, the average value (Bave) of the n light intensity ratios B and the standard deviation (Bsigma) (σ) are calculated. Bave+6σ is then set as an upper threshold (reference value (max)), and Bave−6σ is set as a lower threshold (reference value (min)). As shown in FIG. 8, in the case that the flow-adjusting exhaust ring was missing, the light intensity ratio clearly exceeded the upper threshold. The flow-adjusting exhaust ring 56 being missing or out of place can thus be detected by monitoring the light emission data. According to the abnormality judgment of step S33 described above, data on light emission above, the semiconductor wafer is used as measured data for component part installation defect detection. The data on light emission above, the semiconductor wafer can easily be measured, and moreover changes in accordance with changes in the state of plasma generation in the chamber 10, and hence changes in accordance with whether or not a component part is missing or out of place. Detection of a component part being missing or out of place in the P/C 2 can thus be carried out reliably without causing a decrease in the utilization ratio of the P/C 2. Moreover, according to the abnormality judgment of step S33 described above, the light intensity ratio is calculated from the light emission data. The light intensity ratio is dimensionless, and hence there is no effect from the magnitude of the test data and the reference data used in calculation of the light intensity ratio. Detection of a component part being missing or out of place in the P/C 2 can thus be carried out accurately. Next, the case that data relating to a high-frequency power source (hereinafter referred to as “high frequency data”) is used as measured data for component part installation defect detection will be described. If a component part installation defect arises in the chamber, then the state of plasma generation in the chamber changes. Upon the state of plasma generation in the chamber changing, the high-frequency electrical power applied from the lower high-frequency power source 18 changes so as to maintain the plasma in a desired state. A change in the state of plasma generation in the chamber, and hence a component part installation defect, can thus be detected by monitoring high frequency data, for example the voltage, the current, the phase, or the impedance. FIG. 9 is a graph showing the relationship between the presence/absence of a component part installation defect and high frequency data. In FIG. 9, the axis of abscissas shows the number of the dummy wafer etched (the water count), and the axis of ordinates shows the measured value of the voltage applied in the chamber 10. Moreover, the component part names shown in FIG. 9 indicate the component part removed from the chamber 10 for the wafer count in question. The measured value of the voltage varies with the wafer count because the measured value of the voltage was plotted on the graph for each dummy wafer. Each measured value of the voltage is the average value during the etching of the dummy wafer in question. Moreover, the average value of the measured value of the voltage and the standard deviation (σ) were calculated. Average value+6σ was set as an upper threshold (reference value (max)), and average value−6σ was set as a lower threshold (reference value (min)). As shown in FIG. 9, in the case that the flow-adjusting exhaust ring was missing, and the case that the cover ring and the focus ring were missing, the measured value of the voltage clearly exceeded the upper threshold, and furthermore in the case that the cover ring only was missing, the measured value of the voltage clearly dropped below the lower threshold. The flow-adjusting exhaust ring 56 being missing or out of place, the cover ring 14 and the focus ring 24 being missing or out of place, or the cover ring 14 only being missing or out of place can thus be detected by monitoring the voltage applied in the chamber 10. In FIG. 9 described above, the average value during the etching of each dummy wafer is used as the measured value of the voltage applied in the chamber 10. However, the minimum value or the maximum value may be used instead, in which case a component part other than the above component parts being missing or out of place can also be detected. FIG. 10 is another graph showing the relationship between the presence/absence of a component part installation defect and high frequency data. In FIG. 10, the axis of abscissas shows the lot number for which the automatic setup process was carried out, and the axis of ordinates shows the measured value of the voltage applied in the chamber 10. The measured value of the voltage varies over each lot because 25 dummy wafers were etched in each lot and the measured value of the voltage was plotted on the graph for each wafer in each lot. Each measured value of the voltage is the average value during the etching of the dummy wafer in question. The state of installation of the component parts for each lot was as shown in Table 2 below. TABLE 2Lot No.Details1Reference2Reference3Reference4Reference5Reference6Reference7Reference8Reference9Reference10Reference11Upper electrode plate screw loose12Shield ring askew13No rubber on rear surface of focus ring For example, in lot 11, the dummy wafer etching was carried out with the tightening torque of a screw for installing the upper electrode plate 35 reduced. Moreover, for the reference of lot 1 and so on, all of the component parts were installed properly. Thresholds for the component part installation defect judgment were set based on the measured value of the voltage for these reference lots. As shown in FIG. 10, the average value of the measured value of the voltage and the standard deviation (σ) were calculated for the reference lots, and average value+6σ was set as an upper threshold (reference value (max)), and average value−6σ was set as a lower threshold (reference value (min)). As shown in FIG. 10, the measured value of the voltage in the case that there was no heat-insulating rubber on the rear surface of the focus ring clearly exceeded the upper threshold. The rubber on the rear surface of the focus ring being missing, i.e. a component part installation defect, can thus be detected by monitoring the voltage applied in the chamber 10. According to the abnormality judgment of step S33 described above, high frequency data is used as measured data for component part installation defect detection. The high frequency data can easily be measured, and moreover changes in accordance with changes in the state of plasma generation in the chamber 10, and hence changes in accordance with whether or not there is a component part installation defect. Detection of a component part installation defect in the P/C 2 can thus be carried out reliably without causing a decrease in the utilization ratio of the P/C 2. In FIGS. 9 and 10 described above, the voltage applied in the chamber 10 is used as the high frequency data. However, apart from the voltage, the current, the phase, or the impedance may be used. Detection of a component part cleaning defect in the abnormality judgment step will now be described. If a component part cleaning defect arises in the chamber, then the state of plasma generation in the chamber becomes unstable due to deposit remaining accumulated after cleaning. When carrying out the dummy wafer etching, a component part cleaning defect can thus be detected by monitoring the state of plasma generation in the chamber. Here, a “component part cleaning defect” means a state in which deposit on the surface of a component part has not been removed through wet cleaning or the like but rather remains accumulated thereon. The case that data on light emission above, the semiconductor wafer is used as measured data for component part cleaning defect detection will now be described. If a component part cleaning defect arises in the chamber, then the state of plasma generation in the chamber changes. Upon the state of plasma generation in the chamber changing, the dummy wafer light emission data also changes. A component part cleaning defect can thus be detected by monitoring the light emission data. FIG. 11 is a graph showing the relationship between the presence/absence of a component part cleaning defect and light emission data. In FIG. 11, the axis of abscissas shows the number of the dummy wafer etched (the wafer count), and the axis of ordinates shows the rate of change of the light intensity ratio for the dummy wafer. Moreover, the component part names shown in FIG. 11 indicate the component part having deposit attached to the surface thereof in the dummy wafer etching corresponding to the wafer count in question. Here, the rate of change of the light intensity ratio means the rate of change, from the atmosphere above, the dummy wafer starting to emit light up until 10 seconds has elapsed, of the ratio, out of the light emission data measured over a predetermined wavelength region for the same dummy wafer, between the light emission data at a wavelength at which the light intensity changes sensitively in response to a component part cleaning defect (e.g. 656.5 nm), and the light emission data at a wavelength at which the light intensity does not change regardless of whether or not a component part cleaning defect has arisen (e.g. 374 nm). The reason that a wavelength at which the light intensity changes sensitively in accordance with a component part cleaning defect is 656.5 nm is that CF reaction products emit light strongly at a wavelength of 656.5 nm. Regarding a threshold shown in FIG. 11, the average value of the rate of change of the light intensity ratio in the case that component parts having deposit attached to the surface thereof were not disposed in the chamber 10 was calculated, and the standard deviation (σ) of the rate of change of the light intensity ratio in this case was further calculated, and then the threshold was set to “average value−6σ”. As shown in FIG. 11, the rate of change of the light intensity ratio in the case that deposit was attached to the surface of the upper electrode (CEL) plate, or the case that deposit was attached to the surface of the ESC was clearly below the threshold. An upper electrode cleaning defect or an ESC cleaning defect can thus be detected by monitoring the light emission data. According to the abnormality judgment of step S33 described above, data on light emission above, the semiconductor wafer is used as measured data for component part cleaning defect detection. The data on light emission above, the semiconductor wafer can easily be measured, and moreover changes in accordance with changes in the state of plasma generation in the chamber 10, and hence changes in accordance with whether or not a component part cleaning defect has arisen. Detection of a component part cleaning defect in the P/C 2 can thus be carried out reliably without causing a decrease in the utilization ratio of the P/C 2. Moreover, according to the abnormality judgment of step S33 described above, the rate of change of the light intensity ratio is calculated over 10 seconds from immediately after the light emission starts based on the light emission data. CF reaction products rapidly react with oxygen gas and so on in the chamber 10 and are thus converted into other products, whereby the amount of light at a wavelength of 656.5 nm rapidly decays. By calculating the rate of change of the light intensity ratio, the presence of CF reaction products can thus be detected reliably, and hence component part cleaning defect detection can be carried out more reliably. Next, detection of a leak in the chamber 10 in the abnormality judgment step will be described. If a leak in the chamber 10 arises and hence air from the outside flows into the chamber 10, then in etched dummy wafer light emission data, the light emission amount for light due to constituent gases of air increases. A leakage check can thus be carried out by monitoring the light emission data. Specifically, the leakage check can be carried out by monitoring the light emission data at a wavelength at which the light emission amount changes sensitively upon the influx of, for example, nitrogen gas (N2) out of the constituent gases of air. If nitrogen gas flows into the chamber 10, then the light emission amount for the light emission data at a wavelength of 745 nm increases greatly, whereas the light emission amount for the light emission data at a wavelength of 560 nm hardly changes. In the abnormality judgment of step S33, the ratio of the light emission amount for the light emission data at a wavelength of 745 nm to the light emission amount for the light emission data at a wavelength of 560 nm (hereinafter referred to as the “light emission ratio”) is thus calculated, and a leakage check is carried out by comparing the calculated light emission ratio with a threshold, described below. FIG. 12 is a graph showing the relationship between the chamber leak rate and the light emission ratio. In FIG. 12, the axis of abscissas shows the leak rate as the amount of leak in the chamber 10, and the axis of ordinates shows the light emission ratio. Moreover, regarding the threshold, the average value of the light emission ratio for the chamber 10 in a normal state was calculated, and the standard deviation (σ) of the light emission ratio in this state was further calculated, and then average value+6σ was used as the threshold. Each point in FIG. 12 represents the measured value of the light emission ratio for a leak rate in question, and the alternate long and short dash line shows the relationship between the leak rate and the light emission ratio predicted based on the measured values of the light emission ratio. As shown in FIG. 12, the leak rate when the alternate long and short dash line reaches the threshold is a leak rate when it is judged as “NO GOOD” using the measured value of the light emission ratio, and 0.05 Pa/min (0.4 mT/min). It can thus be detected whether or not there is a leak of rate less than the general permissible value of 0.13 Pa/min (1.0 mT/min). According to the abnormality judgment of step S33 described above, it is detected whether or not there is a leak in the chamber 10 based on the light emission ratio for the etched dummy wafer light emission data, specifically the ratio of the light emission amount for the light emission data at a wavelength of 745 nm to the light emission amount for the light emission data at a wavelength of 560 nm. If a leak in the chamber 10 arises, then nitrogen gas from the outside flows in, and hence the light emission amount for light due to the nitrogen gas increases. Moreover, the light emission ratio is dimensionless, and hence there is no effect from the magnitude of the light emission amount at each of the two wavelengths used in calculation of the ratio of light emission amount. The leakage check can thus be carried out more accurately. As described above, according to the process shown in FIG. 3, in the abnormality judgment, at least one selected from data such as measured data and apparatus logs that change in response to a change in the state inside the chamber 10 is measured, and the measured data is compared with reference data that corresponds to the measured data for a normal state. That is, the abnormality judgment is carried out based on at least one selected from data that change in response to a change in the state inside the chamber 10, without using results of multivariate analysis on the measured data and reference data in the abnormality judgment. The judgment of abnormalities in the P/C 2 can thus be carried out accurately. Moreover, there is no need to reset a normal model every time maintenance is carried out, and hence the utilization ratio of the P/C 2 is not caused to decrease. Next, the seasoning step of step S34 will be described. In seasoning in a conventional resetting process, judgment of whether or not the atmosphere in the chamber conforms to predetermined processing conditions stipulated in a predetermined recipe, i.e. whether or not seasoning has been completed, has been carried out through whether or not a predetermined number of dummy wafers, for example 50 dummy wafers, have been etched. There has thus been needless etching, resulting in the seasoning taking much time. In contrast with this, in step S34 of the automatic setup process according to the present invention, light emission data is monitored. Once the atmosphere in the chamber 10 has become stable so as to conform to predetermined processing conditions stipulated in a predetermined recipe, the state of plasma generation in the chamber becomes stable, and hence the state of emission of light from the atmosphere above, the dummy wafer also becomes stable. It can thus be judged whether the seasoning has been completed by monitoring the light emission data. Specifically, if the difference in the light emission data between two consecutively etched dummy wafers has become low, then it can be judged that the atmosphere in the chamber 10 has become stable and hence the seasoning has been completed. FIG. 13 is a graph showing PDC (posterior data calibration) data, which is light emission data used in the seasoning. As shown in FIG. 13, the PDC data is calculated by first measuring, over a predetermined wavelength region, light emission data L(n) for the nth dummy wafer and light emission data L(n−1) for the (n−1)th dummy wafer, respectively, these dummy wafers being consecutively etched, and then summing the absolute value of the difference between the light emission data at each wavelength over the predetermined wavelength region, for example over i=200 to 800 nm, in accordance with equation (3) below. PDC ( n ) = ∑ i = 200 800 L ( n ) i - L ( n - 1 ) i ( 3 ) Furthermore, in step S34, the PDC data is differentiated, and it is judged whether or not the seasoning has been completed based on the derivative of the PDC data (the PDC differentiated value). Specifically, the point at which the derivative of the PDC data during the seasoning first changes sign from negative to positive is taken as the point at which the seasoning has been completed. FIG. 14 is a graph showing the relationship between the derivative of the PDC data and the etch rate. In FIG. 14, the axis of abscissas shows the number of dummy wafers processed in the seasoning, and the axis of ordinates shows the derivative of the PDC data on the left and the etch rate on the right. As shown in FIG. 14, once six dummy wafers have been processed in the seasoning, the etch rate becomes stable, i.e. the atmosphere in the chamber 10 becomes stable; corresponding to this, as shown by the arrow in FIG. 14, the derivative of the PDC data first changes sign from negative to positive during the processing of the 6th dummy wafer. It can thus be judged whether or not the atmosphere in the chamber 10 has become stable, and hence whether or not the seasoning has been completed, based on the derivative of the PDC data. According to the seasoning step of step S34, it is judged whether or not the seasoning has been completed based on the difference in light emission data between two consecutively etched dummy wafers (PDC data). Once the state of plasma generation in the chamber becomes stable, the dummy wafer light emission data also becomes stable, and hence the difference in the light emission data between two consecutively etched dummy wafers becomes small. Judgment of when the seasoning has been completed can thus be carried out easily. Moreover, according to the seasoning step of step S34, the derivative of the PDC data is calculated, and it is judged whether or not the seasoning has been completed based on the derivative. The derivative of the PDC data can decrease affection by the magnitude of variation of the light emission amount. The judgment of when the seasoning has been completed can thus be carried out more accurately. It is to be understood that the object of the present invention can also be attained by supplying to a system or apparatus (e.g. the APC server) a storage medium storing program code of software that realizes the functions of the embodiment described above, and then causing a computer (or CPU, MPU, etc.) of the system or apparatus, or a controller connected to the apparatus, to read out and execute the program code stored in the storage medium. In this case, the program code itself read out from the storage medium realizes the functions of the embodiment, and hence the program code and the storage medium storing the program code constitute the present invention. The storage medium for supplying the program code may be, for example, a floppy (registered trademark) disk, a hard disk, a magnetic-optical disk, an optical disk such as a CD-ROM, a CD-R, a CD-RW, a DVD-ROM, a DVD-RAM, a DVD-RW, or a DVD+RW, a magnetic tape, a nonvolatile memory card, or a ROM. Alternatively, the program code may be downloaded via a network. Moreover, it is to be understood that the functions of the embodiment described above can be realized not only by executing program code read out by a computer, but also by causing an OS (operating system) or the like operating on the computer to perform part or all of the actual processing based on instructions of the program code. Furthermore, it is to be understood that the functions of the embodiment described above can also be realized by writing the program code read out from the storage medium into a memory provided on an expansion board inserted into the computer or in an expansion unit connected to the computer, and then causing a CPU or the like provided on the expansion board or in the expansion unit to perform part or all of the actual processing based on instructions of the program code. In the embodiment described above, the APC server collects measured data or an apparatus log recorded by the control unit 57. However, as described above, the state of plasma generation in the chamber varies with the type of the P/C, and hence which apparatus log or measured data is closely related to the state of plasma generation in the chamber also varies depending on the type of the P/C. To cope with this, in the present invention, it may be made to be such that the control unit or the like of each P/C has a list of which apparatus log or measured data is closely related to the state of plasma generation for that P/C, and when a given P/C is connected to the APC server, the APC server receives the list for that P/C, and collects an apparatus log or measured data based on the list. As a result, the APC server will not collect unnecessary apparatus logs or measured data, and hence the automatic setup process can be carried out quickly. Moreover, there is no need for the APC server to manage the lists, and hence the burden on the APC server can be reduced. In the embodiment described above, component part installation defects and component part cleaning defects are detected in the abnormality judgment in the automatic setup process. However, the abnormal states detected are not limited to these, but rather, for example, it may also be detected whether or not there is an abnormal electrical discharge in the chamber 10, or whether or not there is leakage of helium gas from between the semiconductor wafer and the ESC 20. In the embodiment described above, the case that the substrate processing apparatus is an etching apparatus has been described. However, substrate processing apparatuses to which the present invention can be applied are hot limited thereto, but rather the substrate processing apparatus may be, for example, a coating/developing apparatus, a substrate cleaning apparatus, a heat treatment apparatus or a wet etching apparatus. Moreover, it is not necessary for the substrate processing apparatus to have both an upper electrode and a lower electrode. Moreover, in the embodiment described above, the substrates processed are semiconductor wafers. However, the substrates processed are not limited thereto, but rather may be, for example, LCD (liquid crystal display) or FPD (flat panel display) glass substrates. |
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abstract | A detection apparatus is usable to detect the neutron absorption capability of a control element of a nuclear installation and includes a neutron radiograph apparatus and a robot apparatus. The neutron radiograph apparatus includes a neutron emission source of variable strength, a detector array, a mask apparatus and a positioning robot all under the control of a central processor and data acquisition unit. The neutron emission source is advantageously switchable between an ON state and OFF state with variable source strength in the ON state, which avoids any need for shielding beyond placing the neutron emission source in an inspection pool at the nuclear plant site including but not limited to the spent fuel or shipping cask laydown pools. The neutron emission source is situated at one side of a wing of the control element and generates a neutron stream, the detector array is situated on an opposite side of a wing, and the neutron emission source and detector array are robotically advanced along the wing. The detector array is monitored in real time, and various masks of the mask apparatus can be positioned between the neutron emission source and the detector array to more specifically identify the position on the blade where the neutrons are passing through. |
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claims | 1. An acceleration system for charged nuclear particles with mass number equal or greater than 1, comprising:a rotating mechanical gantry structure;at least one radiofrequency (RF) linear accelerator configured for producing a pulsed ion beam,said accelerator comprising at least one linac section mounted on the rotating gantry structure, the accelerator configured for rotating around an axis to irradiate a target or patient from more than one direction with the produced pulsed ion beam, the produced pulsed ion beam configured for treating a patient tumor or malformation; andradio-frequency (RF) power generators located outside the rotating gantry structure and connected to the at least one linac section via rotating wave-guide devices. 2. A system for ion acceleration according to claim 1, further comprising:an ion source producing an ion beam; anda particle accelerator, configured as pre-accelerator to impart energy to particles within the ion beam, produced by the ion source, before injecting the ion beam into the at least one linac section mounted on the rotating gantry structure. 3. A system for ion acceleration according to claim 2, wherein the pre-accelerator is another linac section. 4. A system for ion acceleration according to claim 2, wherein the pre-accelerator is a cyclotron. 5. A system for ion acceleration according to claim 1, a system for ion beam delivery mounted on the rotating gantry structure. 6. A system for ion acceleration according to claim 1, wherein,the at least one linac section comprises plural linac sections, andeach linac section is configured to run at a different frequency. 7. A system for ion acceleration according to claim 1, wherein,the linac section comprises plural successive accelerating modules mounted on the rotating gantry structure, andoutput energy of the pulsed ion beam is modulated by varying an input (RF) power to each of the successive accelerating modules which constitute the linac section. 8. A system for ion acceleration according to claim 1, wherein the pulsed ion beam is made of protons. 9. A system for ion acceleration according to claim 8, wherein the pulsed ion beam further is made of carbon ions. 10. A system for ion acceleration according to claim 1, wherein the pulsed ion beam comprises 230 MeV protons. 11. A method of producing radiopharmaceuticals, comprising a step of producing a pulsed ion beam using the system of claim 1. 12. A method of providing cancer radiation therapy, comprising the step of producing a pulsed ion beam using the system of claim 1. 13. An acceleration system for charged nuclear particles with mass number equal or greater than 1, comprising:a rotating mechanical gantry structure;at least one radiofrequency (RF) linear accelerator configured for producing a pulsed ion beam,said accelerator comprising at least one linac section mounted on the rotating gantry structure, the accelerator configured for rotating around an axis to irradiate a target or patient from more than one direction with the produced pulsed ion beam, the produced pulsed ion beam configured for treating a patient tumor or malformation;an ion source configured to produce an ion beam; anda particle accelerator, configured as pre-accelerator to impart energy to particles within the ion beam, produced by the ion source, before injecting the ion beam into the linac section mounted on the rotating gantry structure,wherein at least one of i) the ion source and ii) the pre-accelerator are mounted on the rotating gantry structure. 14. A system for ion acceleration according to claim 13, further comprising radio-frequency (RF) power generators mounted on the rotating gantry structure and directly connected to the at least one linac section. 15. An acceleration system for charged nuclear particles with mass number equal or greater than 1, comprising:a rotating mechanical structure rotating around an axis; andat least one radiofrequency (RF) linear accelerator configured to produce a pulse ion beam,said at least one radiofrequency (RF) linear accelerator comprising at least one linac section mounted on the rotating mechanical structure, the accelerator configured to irradiate a target or patient from more than one direction with the produced pulsed ion beam, the produced ion beam configured for treating a patient tumor or malformation; anda particle accelerator, configured as pre-accelerator and arranged to impart energy to particles within an ion beam before injecting the ion beam into the linac section, wherein the the pre-accelerator is one of i) a cyclotron, and ii) a Fixed Field Alternating Gradient(FFAG) accelerator. 16. A system for ion acceleration according to claim 15, wherein the pre-accelerator is a cyclotron. 17. A system for ion acceleration according to claim 15, wherein the pre-accelerator is a FFAG accelerator. |
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abstract | Improvements to atom interferometers. An improved atom interferometer has a single polarization-preserving fiber, coupled for propagation of beams of two Raman frequencies, and a parallel displacement beamsplitter for separating the laser beams into respective free-space-propagating parallel beams traversing at least one ensemble of atoms. A reflector generates one or more beams counterpropagating through the ensemble of atoms. Other improvements include interposing a beam-splitting surface common to a plurality of parallel pairs of beams counterpropagating through the ensemble of atoms, generating interference fringes between reflections of the beams to generate a detector signal; and processing the detector signal to derive at least one of relative phase and relative alignment between respective pairs of the counterpropagating beams. |
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claims | 1. A container for storing and/or transporting spent nuclear fuel comprising:a body having an outer surface and an inner surface defining an internal cavity;a plurality of blind holes formed into the body, each of the blind holes defined by a floor and a sidewall extending from the floor to an opening in the outer surface of the body;a plurality of trunnions coupled to the body, each of the trunnions comprising:a first component located within one of the blind holes, the first component extending from a first end to a second end and having an inner surface defining a hollow interior; anda second component at least partially located within the hollow interior of the first component, the second component extending from a first end to a second end along a longitudinal axis; andwherein for each of the plurality of trunnions, the second component is axially slidable relative to the first component between: (1) a protruded state in which a portion of the second component protrudes from the outer surface of the body; and (2) a retracted state in which the second component does not protrude from the outer surface of the body. 2. The container according to claim 1 wherein for each of the plurality of trunnions, the second component is welded to the first component to provide the second component of the trunnion with a limited axial load bearing capacity. 3. The container according to claim 1 wherein for each of the plurality of trunnions, an outer surface of the second component is in surface contact with the inner surface of the first component thereby forming an interface pressure between the first and second components to provide the second component of the trunnion with a limited axial load bearing capacity. 4. The container according to claim 1 wherein for each of the plurality of trunnions, the second component comprises a body portion and a flange portion extending radially from the body portion at the second end of the second component, the portion of the second component that protrudes from the outer surface of the body in the protruded state comprising the flange portion. 5. The container according to claim 4 wherein for each of the plurality of trunnions, the second end of the first component comprises a first portion that is flush with the outer surface of the body and a second portion that is recessed relative to the outer surface of the body thereby forming a nesting groove in the second end of the first component, and wherein when the second component is in the retracted state the flange portion of the second component nests within the nesting groove of the first component and is flush with the first portion of the second end of the first component and the outer surface of the body. 6. The container according to claim 1 wherein for each of the plurality of trunnions, the first end of the first component is in surface contact with the floor of the blind hole, an outer surface of the first component is in surface contact with the sidewall of the blind hole, and the second end of the first component is flush with the outer surface of the body. 7. The container according to claim 1 wherein the second component of each of the plurality of trunnions has an axial load bearing capacity and wherein the second component moves from the protruded state into the retracted state when an axial force greater than the axial load bearing capacity is applied to the second end of the second component in a direction towards the body. 8. The container according to claim 1 wherein for each of the plurality of trunnions, in the protruded state a first portion of the second component is located within the hollow interior of the first component and a second portion of the second component protrudes from the hollow interior of the first component, and wherein in the retracted state both of the first and second portions of the second component are located within the hollow interior of the first component. 9. The container according to claim 1 wherein for each of the plurality of trunnions, an entirety of the first component is located within the blind hole with the second end of the first component either flush with or recessed relative to the outer surface of the body when the second component is in both the protruded and retracted states. 10. The container according to claim 1 wherein for each of the plurality of trunnions, in the protruded state the first end of the second component is spaced apart from the floor of the blind hole by a gap and in the retracted state the first end of the second component moves into the gap. 11. The container according to claim 10 wherein axially sliding the second component from the protruded state to the retracted state comprises sliding the second component relative to the first component towards the body in a direction of the longitudinal axis of the second component. 12. The container according to claim 1 wherein in the protruded state, a portion of the second component that is located within the hollow interior of the first component has a length and a diameter, a ratio of the diameter to the length being between 1:1 and 2:1. 13. The container according to claim 1 wherein the second component has a length measured from the first end to the second end, and wherein when the second component is in the protruded state at least two-thirds of the length of the second component is located within the hollow interior of the first component. 14. The container according to claim 1 wherein in the retracted state the second end of the second component is flush with the outer surface of the body. 15. A container for storing and/or transporting spent nuclear fuel comprising:a body having an outer surface and an inner surface defining an internal cavity configured to hold spent nuclear fuel;a hole formed into the body, the hole defined by a floor and a sidewall extending from the floor to an opening in the outer surface of the body;a trunnion at least partially located within the hole, the trunnion comprising:a first component welded to the body within the hole, the first component having a first end that is in contact with the floor of the hole, a second end that is flush with or recessed relative to the outer surface of the body, an outer surface that is in contact with the sidewall of the hole, and an inner surface that defines a hollow interior; anda second component located within the hollow interior of the first component, the second component extending from a first end to a second end along a longitudinal axis, the second component comprising a first portion that is located within the hollow interior of the first component and spaced from the floor of the hole by a gap and a second portion that protrudes from the outer surface of the body; andwherein upon application of an axial force that exceeds a predetermined threshold onto the second end of the second component, the second component slides relative to the first component in an axial direction into the gap. 16. The container according to claim 15 wherein the second component is welded to the first component and wherein the axial force is sufficient to break the weld between the first and second components to permit the second component to slide relative to the first component. 17. The container according to claim 15 wherein an outer surface of the second component is in surface contact with the inner surface of the first component thereby creating an interface pressure between the first and second components that prevents the second component from sliding relative to the first component until the axial force that exceeds the predetermined threshold is applied to the second end of the second component. 18. The container according to claim 15 wherein upon application of the axial force, the second component slides in the axial direction until the second end of the second component is flush with the outer surface of the body. 19. The container according to claim 15 wherein upon application of the axial force, the second component slides in the axial direction until a flange at the second end of the second component abuts the second end of the first component. 20. A container for storing and/or transporting spent nuclear fuel comprising:a body having an outer surface and an inner surface defining an internal cavity;a plurality of blind holes formed into the body, each of the blind holes defined by a floor and a sidewall extending from the floor to an opening in the outer surface of the body;a plurality of trunnions, each of the trunnions located within one of the blind holes and extending from a first end to a second end along a longitudinal axis; andwherein at least one of the plurality of trunnions is axially slidable relative to the body between: (1) a protruded state in which a portion of the trunnion protrudes from the outer surface of the body; and (2) a retracted state in which the trunnion does not protrude from the outer surface of the body. |
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description | The present application claims the benefit of U.S. Provisional Patent Application No. 61/327,394, filed Apr. 23, 2010, the entirety of which is hereby incorporated by reference. The present invention relates generally to the field of reducing environmental contamination in nuclear power plants, and specifically to reducing the migration of tritium from spent nuclear fuel pools. In the operation of water cooled nuclear reactors, tritium is a direct product of nuclear fission. Tritium has a very short biological half life in the human body of 7 to 14 days which reduces the total effects of single-incident ingestion and precludes long term bioaccumulation of tritium from the environment. As tritium is not a strong beta emitter, it is not dangerous externally or simply through contacting a person's skin. However, tritium can be a radiation hazard when it is inhaled, ingested via food stuff, ingested in water, or absorbed through the skin. In the nuclear industry, tritium is a direct product of the operation of nuclear power plants. When a nuclear fuel rod becomes spent through use in the nuclear reactor, it is stored for up to five years or longer in what is known as a spent fuel pool in order to cool the spent fuel rod and rid the spent fuel rod of its radioactive components. Tritium control and recovery needs to be considered in the nuclear plant fuel cycle since production is not a goal of the reactor. Tritium does represent a personnel and environmental radiological hazard in sufficient concentrations. Because tritium concentration in the nuclear storage pools can be relevantly high, a very small amount of tritiated water can affect a large amount of ground water. Further, tritiated water has the same chemical characteristics as regular water, which means that the tritiated water can spread through evaporative loss and redeposit via subcooling condensation on the external surfaces within the spent fuel pool room. Therefore, it is an object of this invention to reduce the migration of tritiated water from the spent nuclear fuel pool. The present invention is directed to systems and methods for reducing the migration of tritium from spent nuclear fuel pools to the internal atmosphere of and surfaces within a containment structure housing a spent nuclear fuel pool. In accordance with certain aspects of the present invention, this reduction in tritium migration can be accomplished via: (1) containing tritiated water vapor that evaporates from the exposed surface of the body of titiated water within the spent nuclear fuel pool within a volume of a controlled atmosphere; (2) inducing condensation of the tritiated water vapor onto control surfaces within the internal atmosphere of the containment structure that houses spent nuclear fuel pool; (3) controlling the thermodynamic properties of the internal atmosphere of the containment structure that houses the spent nuclear fuel pool to minimize evaporation from the exposed surface of the body of tritiated water; and/or (4) preventing escape of tritiated water vapor from the internal atmosphere of the containment structure that houses the spent nuclear fuel pool. In one embodiment, the invention can be a system for reducing tritium migration comprising: a spent nuclear fuel pool comprising a body of tritiated water having an exposed surface; and a cover movable between an open state in which the cover does not obstruct access to the exposed surface of the body of tritiated water and a closed state in which the cover hermetically seals the exposed surface of the body of tritiated water. In another embodiment, the invention can be a system for reducing tritium migration comprising: a spent nuclear fuel pool comprising a body of tritiated water having an exposed surface; and a tent structure comprising a frame and a vapor impermeable membrane, the tent structure movable between an open state in which the tent structure does not obstruct access to the exposed surface of the body of tritiated water and a closed state in which the tent structure hermetically seals the exposed surface of the body of tritiated water. In yet another embodiment, the invention can be a method of reducing tritium migration from a spent nuclear fuel pool containing a body of tritiated water having an exposed surface, the method comprising hermetically sealing the exposed surface of the body of tritiated water with a cover movable between an open-state and a close-state In still another embodiment, the invention can be a method of reducing tritium migration from a spent nuclear fuel pool containing a body of tritiated water having an exposed surface, the method comprising: monitoring a temperature of an internal environment of a containment structure housing the spent nuclear fuel pool; monitoring a temperature of the body of tritiated water of spent nuclear fuel pool; and controlling the temperature of the internal environment of the containment structure so as to be substantially equal to the temperature of the body of tritiated water of the spent nuclear fuel pool. The temperature of the body of tritiated water can be an average temperature or a temperature taken at a location adjacent the exposed surface of the body of tritiated water. A proper control system, including temperature sensors and processor, can be utilized to repetitively measure and compare the temperatures of both the internal atmosphere of the containment structure and the body of tritiated water, and to adjust the temperature of the internal atmosphere of the containment structure when a threshold temperature differential is identified via the comparison. In a further embodiment, the invention can be a method of reducing tritium migration from a spent nuclear fuel pool containing a body of tritiated water having an exposed surface, the method comprising: positioning at least one dehumidifier within an internal environment of a containment structure housing the spent nuclear fuel pool; condensing tritiated water vapor from the internal environment using the dehumidifier; and flowing the condensed tritiated water vapor in liquid form back into the spent nuclear fuel pool or into a waste management sub-system. In a yet further embodiment, the invention can be a system or method of reducing tritium migration from a spent nuclear fuel pool that utilizes humidity traps having exposed surfaces that are configured to induce condensation of tritiated water vapor in an internal environment of a containment structure housing the spent nuclear fuel pool. In one such embodiment, the humidity traps can comprise sacrificial collection plates comprising metal fins, tubes or plates embedded in a desiccate adsorption material. Other surfaces within the room can be coated with a material that is less conducive to causing condensation thereon as compared to the exposed outer surfaces of the humidity traps. Once such coating is a polymer coating. In an even further embodiment, the invention can be a system of method of reducing tritium migration from a spent nuclear fuel pool by humid air from entering and/or escaping the internal environment of a containment structure housing the spent nuclear fuel pool. In one such embodiment, entrance, points to the internal environment will comprise air locked chambers which are heated to drive moisture out, thereby further reducing moisture enriched air from entering or leaving the internal environment. The air-lock chamber can be operated such that only one egress is open at a time and sufficient time between egress openings is permitted to lapse that allows the heating system to suppress moisture within the chamber. In another embodiment, the invention can be a method of reducing tritium migration from a spent nuclear fuel pool by coating and sealing surfaces within an internal environment of a containment structure housing the spent nuclear fuel pool that are not intended to have evaporative tritiated water condense thereon. Sealing of floor gaps and other seams will inhibit any tritiated water from migrating and/or leaking into those gaps and seams before detection and clean-up. It has been discovered by the present inventor that one of the primary mechanisms by which tritium is transferred from the interior coolant system to the internal atmosphere of the containment structure and, hence, to the external atmosphere, may be through evaporative losses of the body of tritiated water of the spent nuclear fuel pool. Evaporative loss within the spent nuclear fuel pool increases with a greater difference between the pool saturation vapor pressure and the vapor pressure of the internal chamber of the containment structure at room temperature. The water of the spent nuclear fuel pool is heated by the decay heat generated by the spent nuclear fuel. While variable, the heating load in the spent nuclear fuel pool is substantial and further drives evaporative loss. When the body of tritiated water of the spent nuclear fuel pool evaporates, it takes the tritium with it, thereby carrying tritium (in the form of tritiated water vapor) into the atmosphere. Once the tritiated water vapor reforms in liquid form as condensate, the condensed tritiated water vapor can migrate to the outside environments via a variety of transport mechanisms. One aspect of the present invention is to monitor and control the temperatures in the pool and the room surrounding the pool so as to keep them as close to the same as possible. By maintaining the pool temperature close to the room temperature, the opportunity for evaporation of the water in the spent fuel pool is decreased. Therefore, the opportunity for tritiated water to migrate to the internal environment of the building or to the external environment outside of the building is also decreased. In another embodiment, to reduce humid air from entering (or exiting) the internal atmosphere of the containment structure housing the spent nuclear fuel pool, aggress points to the internal atmosphere of the containment structure will have air-locked type chambers which are heated to drive moisture out, thereby further reducing the introduction and/or escape of moisture enriched air into and/or out of the internal atmosphere of the containment structure. The air-lock chamber should be operated such that only one door is open at a time and sufficient time between door openings is permitted to lapse that allows the heating system to suppress moisture within the enclosure. Additionally, surfaces within the internal atmosphere of the containment structure should be re-evaluated for coating on surfaces and sealing for floor areas. In other words, all surfaces that are not intended to have evaporative tritiated water condense thereon should be coated. The coating will assist in prevention of tritiated water surface sub-cooling on those surfaces and increase the benefit of the collection plates which remain uncoated. Sealing of floor gaps and other seams will inhibit any tritiated water from migrating and/or leaking into those gaps and seams before detection and clean-up. In a typical nuclear power plant, the spent nuclear fuel pools are not covered, but rather remain open and exposed to the internal atmosphere of the containment structure (which is typically a building) within which the spent nuclear fuel pool is housed. In addition to the aforementioned safeguards, the system of the present invention provides one or more covers (in the exemplified embodiment, there are two covers) that reduce the migration of tritium by sealing the exposed surface of the spent nuclear fuel pool. Referring first to FIGS. 1, 2 and 7 concurrently, a system 1000 for reducing tritium migration is disclosed (hereinafter the “TMR system”). The TMR system 1000 generally comprises a spent nuclear fuel pool 100, a tent structure 200, a tarp assembly 300, a condenser 400, and a plurality of humidity traps 500. Of course, in alternate embodiments of the present invention, one or more of the aforementioned components can be omitted. For example, in certain embodiment, only one of the tenet structure 200 or the tarp assembly 300 can be utilized in the TMR system 1000. Moreover, the condenser 400 and the humidity traps 500 can be omitted as desired. Both the tent structure 200 and the tarp assembly 300 act as covers for hermetically sealing the exposes surface 101 of the body of tritiated water 102. As discussed in greater detail below, each of the tent structure 200 and the tarp assembly 300 are movable between an open state in which access to the exposed surface 101 of the body of tritiated water 102 is unobstructed and a closed state in which the exposed surface 101 of the body of tritiated water 102 is hermetically sealed. While the exemplified embodiment of the TMR system 100 comprises both the tent structure 200 and the tarp assembly 300, one of these covers can be omitted as desired in alternate embodiments of the invention. Moreover, while both the tent structure 200 and the tarp assembly 300 hermetically seal the exposes surface 101 of the body of tritiated water 102 in the exemplified embodiment of the present invention, it is possible that one of these covers will not form a seal that is hermetic in nature in alternate embodiments. While not illustrated, the entire spent nuclear fuel pool 100 is housed within an internal atmosphere of containment structure (not illustrated). The spent nuclear fuel pool 100 is formed in a radiation shielding body 110, which in the exemplified embodiment is a concrete monolith. The top surface 111 of the radiation shielding body 110 forms the floor surface of the containment structure. As is known in the art, spent nuclear fuel 150 is loaded within the spent nuclear fuel pool 100 for a period of time after removal from the reactor vessel for cooling. The spent nuclear fuel 150 is supported at the bottom of the spent nuclear fuel pool in racks or other structures. The tent structure 200 is illustrated in the closed state in FIG. 1 and in the open state in FIG. 2 (also in FIG. 3). When in the open state, the tent structure 200 is removed from its position over the spent nuclear fuel pool 100 and placed in a position such that the tent structure does not obstruct access to the spent nuclear fuel pool 100. For example, in the open state, the tent structure 200 may be positioned on the floor surface 211 of the containment structure away form the spent nuclear fuel 100 or on some other supporting structure. The tent structure 200 generally comprises a frame 210 and a vapor impermeable membrane 220. The frame 210 comprises a base 211 (formed of four beams 211A-D) and a plurality of uprights 212A-D that are connected to and extend upward from the base 211, thereby forming a dome-like structure 213 in the shape of a four-sided pyramid. Of course, the frame 210 can take on a wide variety of structural shapes, sizes and geometries, none of which are limiting of the present invention unless specifically claimed. Similarly, while the base 211 forms a rectangular shape, the invention is not so limited and the base 211 can take on other shapes, such as circular, oval, triangular, or other polygonal shapes. The exact size and shape of the base 211 will be dictated by the size and shape of the spent nuclear fuel pool 110. The frame 210 is formed by a plurality of interconnected structural beams 211A-D, 212A-D. While the structural beams 211A-D, 212A-D are hollow tubes in the exemplified embodiment, the structural beams 211A-D, 212A-D can take on a wide variety of embodiments, including without limitation I-beams, T-beams, angle-beams, solid rods, and combinations thereof. The interconnectivity between the structural beams 211A-D, 212A-D can be achieved by welding, clamping, fasteners, integral formation or combinations thereof as is known in the art. The frame 210 (including its structural beams 211A-D, 212A-D) is designed to be suitably rigid and robust so as to be capable of maintaining its three-dimensional geometry when the vapor impermeable membrane 220 is coupled thereto. The structural beams 211A-D, 212A-D may be formed of a metal, plastic, or composite material in certain embodiments. The frame 210 of the tent structure 200 further comprises an eyelet 217 located at an apex 214 of the dome-like structure 213 to facilitate lifting and handling of the tent structure 200 by a crane or other lifting device within the containment structure/building. The eyelet 217 provides a structure by which the tent structure 200 can be grasped for repetitive movement between the closed state (FIG. 1) and the open state (FIG. 2) of the tent structure 200. The vapor impermeable membrane 220 is coupled to the frame 210 so that an inner surface 216 of the vapor impermeable membrane 220 forms a cavity 215 having an open bottom end defined by the base 211 of the frame 210. With the exception of the open bottom end, the vapor impermeable membrane 220 forms a hermetically sealed structure. Thus, if any seams exist, they are preferably fluid tight. However, it is preferable that the vapor impermeable membrane 220 be made as a single, unitary piece of material with seamless construction in certain embodiments. The vapor impermeable membrane 220 is preferably a thin flexible sheet of material that is impervious to water vapor. Of course, in alternate embodiments, the vapor impermeable membrane 220 could be rigid. However, flexibility may be preferred in embodiments where the frame 210 may be collapsible. The vapor impermeable membrane 220 can be a single-layer construct or a multi-layer laminate. If the vapor impermeable membrane 220 is a single-layer construct, the single layer is made of a rubber, polymer, metal or other material that is impervious to water vapor. In one embodiment wherein the vapor impermeable membrane 220 is a multi-layered laminate, the vapor impermeable membrane 220 can comprise at least one vapor impermeable layer 221, at least one radiation shielding layer 222, and at least one thermal insulating layer 223 (see FIG. 6). The vapor impermeable layer 221, in one embodiment, can be constructed of a rubber, a thermoplastic elastomer, a polymer, a metal or other material that is impervious to water vapor. The radiation shielding layer 222 can be constructed of materials that are known in the art to absorb/shield neutron radiation and/or gamma radiation. The thermal insulating layer 223 can be formed of, without limitation, blankets of alumina-silica fire clay (Kaowool Blanket), oxides of alumina and silica (Kaowool S Blanket), alumina-silica-zirconia fiber (Cerablanket), and alumina-silica-chromia (Cerachrome Blanket). Constructed out of material such as a Kaewool blanket. In the specific embodiment exemplified, the upper-most and lower-most layers of the vapor impermeable membrane 220 are vapor impermeable layers 221 wherein the radiation shielding layer 222 and the thermal insulating layer 223 are sandwiched therebetween. Referring now to FIGS. 1-2 and 6-7 concurrently, when the tent structure 200 is positioned in the closed state, the base 211 of the tent structure 200 forms a hermetically seal that forms a perimeter that surrounds the spent nuclear fuel pool 100, thereby hermetically sealing the cavity 215 from the remainder of the internal atmosphere of the containment structure/building. This hermetic seal can be formed in a wide variety of ways. In the exemplified embodiment, a gasket 230 is provided on the floor surface 211 that forms a closed perimeter about the spent nuclear fuel pool 100. The gasket 230 is sized and shaped to correspond to the size and shape of the base 210 of the tent structure 200. Thus, when the tent structure 200 is in the closed state, the base 210 of the tent structure 200 rests atop and is in surface contact with the gasket 230, thereby forming a hermetically sealed interface that forms a closed-perimeter about the spent nuclear fuel pool 100. A groove 231 is provided in the top surface of the gasket 230. The groove 231 provides a depression in which the base 211 of the tent structure 200 can nest, thereby improving the quality of hermetic seal and providing a means by which proper positioning of the tent structure 200 can be ensured. While not illustrated, one or more clamps can be provided that would clamp the base 210 of the tent structure 200 to the floor surface 211, thereby further compressing the gasket 230 and ensuring a hermetic seal. In one embodiment, at least one clamp is provided on each side of the tent structure 200. In alternate embodiments, the hermetic seal between the base 211 of the tent structure 200 can be accomplished by mere surface contact between the base 211 and the floor surface 211 without the use of a gasket or other sealing structure. In still other embodiments, a compressible gasket (similar to gasket 230) could be connected directly to the base 211 rather than to the floor surface 211. As with the base 211 of the tent structure 200, the gasket 230 can take on a wide variety of shapes but is preferably selected to correspond in size and shape to the base 211. As mentioned above, when the tent structure 200 is in the closed state, the cavity 215 becomes hermetically sealed. As a result, tritiated water vapor coming from the exposed surface 101 of the body of tritiated water 101 is contained within the cavity 215 (even if the tarp assembly 300 is omitted as in FIG. 8). The tritiated water vapor within the cavity 215 may condense on the inner surface 216 of the vapor impermeable membrane 220 where it is either re-directed back into the spent nuclear fuel pool 100 or to waste collection sub-system. The condensed tritiated water vapor can be captured and directed by an internal funnel 240 (discussed in detail below with respect to FIG. 8). Furthermore, because tritiated water vapor may build-up within the hermetically sealed cavity 215 that is formed between the vapor impermeable membrane 220 of the tent structure 200 and the exposed surface 101 of the body of tritiated water 102 over time, a condenser 400 can be operably coupled to the hermetically sealed cavity 215 to dehumidify the air within the hermetically sealed cavity 215 prior to moving the tent structure 200 into the open state and/or periodically as desired. The condenser 400 is a cross-flow heat exchanger having a primary coolant circuit comprising an incoming cool leg 401 and outgoing warm leg 402. The condenser 400 further comprises a wet air intake 403 and a dry air exhaust 404. Both the wet air intake 403 and the dry air exhaust 404 are operably coupled to be in fluid communication with the hermetically sealed cavity 215 (see FIG. 2 for positioning of wet air intake 403 and dry air exhaust 404). As such, during operation, the condenser 400 will draw wet air (i.e., air containing high levels of tritiated water vapor) into the condenser 400 and into contact with the primary coolant circuit. Upon contacting the tubes of the primary coolant circuit, the tritiated water vapor will condense and gather within the condenser 400 as tritiated water in liquid form. This liquid form of tritiated water will then be directed to either a waste management sub-system or back into the spent nuclear fuel pool 100 via a drain line 406. After the wetted air is cooled (and the tritiated water vapor is condensed out therefrom), the resulting dried air is reintroduced back into the hermetically sealed cavity 215 via the dry air exhaust line 404. In the exemplified embodiment, the wet air intake 403 and the dry air exhaust 404 extend through the floor surface 111 of the internal atmosphere of the containment structure so as to avoid interfering with the hermetic seal formed about the perimeter of the spent nuclear fuel pool 100 by the cooperation of the tent structure 200 and the gasket 230. However, in other embodiments, the tent structure 200 can be provided with connection ports to which the wet air intake 403 and the dry air exhaust 404 can be permanently coupled detachably coupled. Furthermore, in certain other embodiments, the wet air intake 403 and the dry air exhaust 404 may be operably coupled to the hermetically sealed cavity 315 formed between the tarp assembly 300 and the exposed surface 101 of the body of tritiated water 102 (discussed in greater detail below). Operable coupling to the hermetically sealed cavity 315 can be instead of or in addition to the coupling to the hermetically sealed cavity 215. Referring now to FIGS. 2, 3 and 7 concurrently, the tarp assembly 300 will be described in greater detail. As mentioned above, the tarp assembly 300 can be used instead of or in addition to the tent structure 200. Similar to the tent structure 200, the tarp assembly 300 is movable between an open state (FIG. 3) in which the tarp assembly 300 does not obstruct access to the exposed surface 101 of the body of tritiated water 102 and a closed state (FIGS. 2 and 7) in which the tarp assembly 300 hermetically seals the exposed surface 101 of the body of tritiated water 102. The tarp assembly 300 generally comprises a vapor impermeable membrane 320, two seal roller assemblies 340A,B, two channel guides 350, a feed roller assembly 360 and an anchor assembly 370. The vapor impermeable membrane 320 of the tarp assembly 300 is similar in construction and materials as described above with respect to the vapor impermeable membrane 320 of the tent structure 200. The vapor impermeable membrane 320 of the tarp assembly 300 is preferably a thin flexible sheet of material that is impervious to water vapor. Of course, in alternate embodiments, the vapor impermeable membrane 320 could be rigid. In such an embodiment, the vapor impermeable membrane 320 could take the form of one or more platen structures that are retractable. However, flexibility is preferred to facilitate wrapping about the feed roller assembly 360. The vapor impermeable membrane 320 can be a single-layer construct or a multi-layer laminate. If the vapor impermeable membrane 320 is a single-layer construct, the single layer is made of a rubber, a thermoplastic elastomer, a polymer, a metal or other material that is impervious to water vapor. In one embodiment wherein the vapor impermeable membrane 320 is a multi-layered laminate, the vapor impermeable membrane 320 can comprise at least one vapor impermeable layer 321, at least one radiation shielding layer 322, and at least one thermal insulating layer 323 (see FIGS. 4 and 5). The vapor impermeable layer 321, in one embodiment, can be constructed of a rubber, polymer, metal or other material that is impervious to water vapor. The radiation shielding layer 322 can be constructed of materials that are known in the art to absorb/shield neutron radiation and/or gamma radiation. The thermal insulating layer 323 can be formed of, without limitation, blankets of alumina-silica fire clay (Kaowool Blanket), oxides of alumina and silica (Kaowool S Blanket), alumina-silica-zirconia fiber (Cerablanket), and alumina-silica-chromia (Cerachrome Blanket). Constructed out of material such as a Kaewool blanket. In the specific embodiment exemplified, the upper-most and lower-most layers of the vapor impermeable membrane 320 are vapor impermeable layers 321 wherein the radiation shielding layer 322 and the thermal insulating layer 323 are sandwiched therebetween. When the tarp assembly 300 is in the closed state, the vapor impermeable membrane 320 seals the exposed surface 101 of the body of tritiated water 102, thereby forming a hermetically sealed cavity 315 between an undersurface of the vapor impermeable membrane 320 and the exposed surface 101 of the body of tritiated water 102. The hermetic nature of the sealed cavity 315 can be formed by forming a hermetic seal along a perimeter region of the vapor impermeable membrane 320. In the exemplified embodiment, this is accomplished by creating hermetic seals between the lateral edges of the vapor impermeable membrane 320 and the guide channels 350A, B (see FIG. 4), a hermetic seal between a distal portion of the vapor impermeable membrane 320 and the seal roller assembly 340A (see. FIG. 5), and a hermetic seal between a proximal portion of the vapor impermeable membrane 320 and the seal roller assembly 340B (identical to that shown in FIG. 5). Of course, the invention is not so limited and the vapor impermeable membrane 320 can form a hermetic seal about the perimeter of the spent nuclear fuel in a variety of manners, including the use of a heavy mat, bar clamps, suction, a tight fit, wrapping, or combinations thereof. Still other methods of ensuring sufficient contact between the vapor impermeable membrane 320 and the floor surface 111 can be used, such as, for example, taping or hook-and-loop type fasteners. The feed roller assembly 360 and the seal roller assembly 340B are positioned on a first side 104 of the spent nuclear fuel pool 100 while the anchor assembly 370 and seal roller assembly 340A are positioned on a second side 103 of the spent nuclear fuel pool 100 that is opposite the first side 104. Similarly, the guide channels 350A, 350B are located on opposite lateral sides 106, 105 of the spent nuclear fuel pool 100 respectively. All of these assemblies are mounted to the floor surface 111. In alternate embodiments, the aforementioned components and assemblies 340A-B, 350A-B, 360, 370 can be positioned in alternate locations relative to one another and/or the spent nuclear fuel pool 100. For example, in one embodiment, the guide channels 350A, 350B (and other structures) can be mounted on the vertical walls of the spent nuclear fuel pool 100 above the water level. As discussed below, the guide channels 350A, 350B function to form a hermetic seal with the vapor impermeable membrane 320 along the lateral edges of the vapor impermeable membrane 320. However, the guide channels 350A-B also function to mechanically secure the lateral edges of the vapor impermeable membrane 320 within the guide channels 350A-B so as to prevent the vapor impermeable membrane 320 from sagging into contact with the exposed surface 101 of the body of tritiated water 102 (see FIG. 4). The feed roller assembly generally comprises a pair of mounting brackets 363, a roller 361, and a drive motor 363. The roller 361 is rotatably mounted between the mounting brackets 363 so that the roller 361 can be rotated both clockwise and counterclockwise. The drive motor 362 is operably coupled to the roller 361 so that the roller can be rotated as desired automatically. A proximal end portion of the vapor impermeable membrane 320 is coupled to the roller 361 so that rotation of the roller 361 in one direction causes the vapor impermeable membrane 320 to wrap around the roller and become retracted from the closed state to the open state. Rotation of the roller 361 in the opposite direction allows the vapor impermeable membrane 320 to become unwrapped from the roller 361 and become protracted from the open state to the closed state. The anchor assembly 370 comprises a first anchor 371 and a second anchor 372. The first and second anchors 371, 371 comprises a retractable cord, in the form of cables 373, that are secured to the distal edge of the vapor impermeable membrane 320. The first and second anchors are spring-loaded so as to keep tension on the vapor impermeable membrane 320 at all times, thereby keeping the vapor impermeable membrane 320b taut during movement between the closed-state to the open state and vice-versa. The spring force of the anchors 371, 372 is selected so as to be overcome by the activation of the motor 362 to retract the vapor impermeable membrane 320 without tearing the vapor impermeable membrane 320. When the vapor impermeable membrane 320 is in the open state, the motor only need to be put into neutral and the spring force of the anchors 371, 372 can automatically pull the vapor impermeable membrane 320 into the closed-state. Of course, the location of the drive motor can be switched along with the biasing spring force in alternate embodiments. For example, in one particular embodiment, the vapor impermeable membrane 320 will always be biased into the open state, thus requiring activation of the drive motor and subsequent locking to maintain the vapor impermeable membrane 320 in the closed state. Such an arrangement can ensure venting of the spent nuclear fuel pool 100 in the event of power loss. In such an embodiment, the roller 360 of the feed roller assembly 360 may be spring-loaded while the drive motors are located within the anchor assembly 370. The anchors 371, 372 are positioned so that when the vapor impermeable membrane 320 is in the open state, the cables 373 do not extend over the exposes surface 101 of the body of tritiated water 102. As a result, the cables 373 do not present any potential of interfering with the loading and unloading of spent nuclear fuel from the spent nuclear fuel pool 100. Thought of another way, the retractable cables 373 are located outboard of a perimeter (formed by sides 103-106) of the spent nuclear fuel pool 100 when the vapor impermeable membrane 320 is in the open-state. In the exemplified embodiment, the cables 373 extend into the channel guides 350A-B when the vapor impermeable membrane 320 is in the open-state, thereby further preventing any danger of accident or obstruction. Referring now to FIG. 2-4 concurrently, the guide channels 350A-B are mounted to the floor surface 111 but can be mounted elsewhere as discussed above. Each of the guide channels 350A-B is an elongated member and is of the C-channel type. Of course, other cross-sections can be used to form the guide channels 350A-B as desired. Each of the guide channels 350A-B comprises an open slot 351 in which a flanged lateral edge 325 of the vapor impermeable membrane 320 nests. The flanged later edges 325 of the vapor impermeable membrane 320 can slide longitudinally within the open slots 351 of the guide channels 350A-B, thereby affording the vapor impermeable membrane 320 to be moved between the open state and the closed state as desired. However, due to their flanged nature, the lateral edges 325 of the vapor impermeable membrane 320 can not be transversely pulled from the slots 351 of the guide channels 350A-B due to the interference fit. In alternate embodiments, the lateral edges 325 of the vapor impermeable membrane 320 may be provided with rollers to facilitate ease of movement within the guide channels 350A-B instead of or in addition to the flanged profile. Furthermore, in certain instances, it may be desirable to add linear gaskets 352, 353 to the longitudinal edges of the longitudinal openings of the slots 351. Such linear gaskets 352, 353 serve not only to increase the integrity of the hermetic seals but can also protect the vapor impermeable membrane 320 from sharp edges. Moreover, the materials of the linear gaskets 352, 353 can be selected to provide reduced frictional resistance of sliding between the vapor impermeable membrane 320 and the guide channels 350A-B. Referring now to FIGS. 2-3 and 5, each of the seal roller assemblies 340A-B comprises a pair of mounting brackets 343 and pair of compressible rollers 341, 342. The compressible rollers 341, 342 are rotatably mounted for free rotation to the mounting brackets 343. The tolerance between the compressible rollers 341, 342 is selected so that the vapor impermeable membrane 320 can fit between and be freely pulled through the compressible rollers 341, 342. Each pair of the compressible rollers 341, 342 form a hermetic seal with the vapor impermeable membrane 320 along their length. The seal roller assemblies 340A-B are located adjacent the ends of the guide channels 350A-B to minimize any gaps/tolerances through which air can flow. While the seal roller assemblies 340A-B are exemplified with two rollers 341, 342, in other embodiments, other structures may be used to create the desired hermetic seals along the proximal and distal perimeter portions of the vapor impermeable membrane 320. Such structures can be used instead of or in addition to one or more of the rollers 341, 342. For examples, compression bars and gaskets may be used along with other known techniques for creating a hermetic seal. All gaskets and rollers described herein can be constructed of a vapor impermeable and compressible material, such as a thermoplastic elastomer, rubber or polymers. In one embodiment, the vapor impermeable membrane 320 is made with a color that contrasts with the pool color so that if the cover is introduced into the pool, in whole or in part, the vapor impermeable membrane 320 can be easily identified and retrieved. This color contrast will avoid and prevent foreign materials from being lost in critical plant systems. Referring now to FIG. 8, an alternate embodiment of the TMR system 1000 is exemplified. The TMR system 1000 of FIG. 8 is similar to that of the TMR system of FIG. 1 with the exception that the tarp assembly 300 is omitted and an internal funnel 240 has been added to the tent structure 200. The internal funnel 240 is also made of a material that is impervious to water vapor and liquid, such as those materials discussed above for the membranes 220, 320. The internal funnel 240 directs tritiated water vapor that condenses on the inner surface 216 of the vapor impermeable membrane 220 either into a waste management subsystem or back into the body of tritiated water 102 of the spent nuclear fuel pool 100. In the exemplified embodiment, the internal funnel 240 directs the condensed tritiated water vapor back into the body of tritiated water 102 of the spent nuclear fuel pool 100. Thus, the internal funnel 240 comprises a drip edge 241 that is located above the exposed surface 101 of the body of tritiated water 102 of the spent nuclear fuel pool 100. As a result, the tritiated water vapor that condenses on the inner surface 216 of the vapor impermeable membrane 220 drips back into the body of tritiated water 102 of the spent nuclear fuel pool 100. The internal funnel 240 is sloped in the opposite direction of the vapor impermeable membrane 220 of the tent structure 200. In alternate embodiments, the internal funnel 240 can be adapted to collect the condensate that forms on the vapor impermeable membrane 220. The internal funnel 240 is preferably sized and configured to permit air communication between the exposed surface 101 of the body of tritiated water 102 and the apex 214 of the tent structure 200, while still covering a large subsurface area under the vapor impermeable membrane 220. The internal funnel 240 is conically formed downward. In alternate embodiments, a drainage tube can be supplied at the bottom of the funnel 240. The drainage tube can be connected to a drain at the bottom of the conically formed funnel 240 and can transport the condensation caught in the funnel 240 either back to the spent fuel pool or to a separate waste system. While not illustrated, either of the TMR systems 1000 of FIG. 1 or 8 can further comprise industrial dehumidifiers positioned within the containment structure adjacent the spent nuclear fuel pool 100 to reduce humidity. The dehumidifiers can include an enclosed collection basin to prevent re-evaporation of the condensate. In a preferable embodiment, the dehumidifiers are hard piped to an appropriate waste collection tank to transfer any water caught in the dehumidifiers directly to the waste collection tank. The dehumidifiers can be especially important during periods of operation where the pool needs to be uncovered due to work in the pool. Often, at the start of nuclear power plant refueling operations, the spent fuel pool has increased heat loading and therefore elevated temperature and humidity in the pool room. The increase in pool temperature drives greater evaporative losses and hence greater tritiated water distribution. The dehumidifiers seek to mitigate the extent of the evaporative losses that may occur, if not eliminate them altogether, during plant refueling operations. In addition, the TMR system 1000 further comprises a plurality of humidity traps 500. The humidity traps 500 induce sub-cooling of any evaporated water onto controlled surfaces. In other words, the traps 500 are controlled surfaces within the containment structure that are specifically designed so that tritiated water evaporation is attracted to them. Specifically, sacrificial collection plates 501 comprising metal fins, tubes or plates are embedded in a desiccate adsorption material 502. The humidity traps 500 are positioned throughout the containment structure. A desiccant is a hygroscopic, or water attracting substance that induces or sustains a state of extreme dryness. These sacrificial collection plates 501 further attract and capture any water vapor that escapes the tent structure 200 and/or the tarp assembly 300. Moisture which sub-cools on the metal sacrificial collection plates 501 will drip onto the desiccate material 502 thereby trapping the moisture and associated tritiated water by adsorption and capillary condensation. The uncoated metal collectors 501 are more attractive targets to the tritiated water than coated materials in the room. Thus, standalone desiccate moisture traps 500 should be introduced at strategic locations within and around the pool enclosure. The invention is not limited to having all of the above discussed attributes combined or in use at the same time. Rather, any one of the membrane cover, tent enclosure with catch basin, dehumidifiers, sacrificial collection plates, sealed entrance points, coated room surfaces or temperature monitoring may be used exclusive of the others. Furthermore, any combination of the aspects of this invention may be used in conjunction to increase the likelihood of preventing and/or controlling tritium migration from nuclear fuel storage pools. However, creating a system, method and/or apparatus that incorporates all of the inventive features described herein will provide the greatest likelihood of preventing and/or controlling tritium migration from nuclear fuel storage pools. While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. |
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056065868 | abstract | An exposure method using X-rays from a synchrotron radiation source includes determining a relationship between an X-ray intensity distribution and an exposure amount distribution in an exposure area; and effecting exposure operation while controlling a dose amount for respective positions in the exposure area using the relationship, wherein the dose amount is controlled by changing a driving profile of a movable shutter for controlling the exposure operation, and wherein the relationship is in the form of a proportional coefficient between an X-ray intensity and the exposure amount as a function of position information in the exposure area. |
claims | 1. A tanning module comprising: a housing; a tridimensional reflector disposed in the housing; and at least one discoid radiation filter; wherein the at least one discoid radiation filter covers the radiation emitting area of the reflector and is disposed on a first side of the housing, at least one opening being provided in the reflector for the installation and electrical, connection of a tanning radiator, and the reflector having its maximum cross section in the plane of the radiation emitting area, wherein the housing is configured on a second side opposite the radiation filter in the form of a quadrilateral pyramid with a rectangular base and flattened pyramid apex and that the rectangular base faces in the direction of the at least one radiation filter, wherein an intake plate is disposed between housing and reflector in which case the radiation emitting area of the reflector is shifted upward or downward from the plane of the intake plate, at least one intake opening being formed between intake plate and reflector, and the intake plate has a cut-out for the reflector which in vertical protection onto the at least one discoid radiation filter has the size of the radiation emitting area of the reflector. 2. The tanning module according to claim 1 , wherein the at least one discoid radiation filter is aligned parallel to the radiation emitting area of the reflector. claim 1 3. The tanning module according to claim 1 , wherein the base of the pyramid is aligned parallel to the at least one discoid radiation filter. claim 1 4. The tanning module according to claim 1 , wherein the flattened pyramid apex is formed by a planar part of the housing wall. claim 1 5. The tanning module according to claim 4 , wherein the planar housing wall portion is aligned parallel to the base of the pyramid. claim 4 6. The tanning module according to claim 1 , wherein the flattened pyramid apex is formed by a vaulted housing wall portion. claim 1 7. The tanning module according to claim 6 , wherein the vaulted housing wall portion is configured concavely or convexly with respect to the base of the pyramid. claim 6 8. The tanning module according to claim 1 , wherein a rectangular housing wall area adjoins the base of the pyramid. claim 1 9. The tanning module according to claim 1 , wherein the reflector is cupular or tub-shaped. claim 1 10. The tanning module according to claim 9 , wherein the bottom of the cupular or tub-shaped reflector is vaulted. claim 9 11. The tanning module according to claim 9 , wherein the bottom of the capular or tub-shaped bottom of the reflector is made plane-parallel to the at least one discoid radiation filter. claim 9 12. The tanning module according to claim 1 , wherein a perimeter of the reflector parallel to the radiation emitting area describes a circle, an ellipse, a rectangle or a polygon. claim 1 13. The tanning module according to claim 12 , wherein the reflector is formed of facets and the perimeter of the reflector parallel to the radiation emitting area describes a dodecagon. claim 12 14. The tanning module according to claim 13 , wherein the reflector has a height of 90 mm to 95 mm and the dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) in the range of 210 mm to 230 mm. claim 13 15. The tanning module according to claim 14 , wherein the reflector has a height of 93.6 mm and the dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) of 210 mm. claim 14 16. The tanning module according to claim 13 , wherein the reflector has a height ranging from 110 mm to 125 mm, and the dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) ranging from 170 mm to 200 mm. claim 13 17. The tanning module according to claim 16 , wherein the reflector has a height of 118.7 mm and the dodecagon has in the plane of radiation emitting area a maximum diameter (corner to corner) of 184 mm. claim 16 18. The tanning module according to claim 13 , wherein the reflector has a height ranging from, 75 mm to 90 mm, and The dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) ranging from 205 mm to 235 mm. claim 13 19. The tanning module according to claim 18 , wherein the reflector has a height of 118.7 mm and the dodecagon has in the plane of radiation emitting area a maximum diameter (corner to corner) of 184 mm. claim 18 20. The tanning module according to claim 1 , wherein the housing has at least one air exhaust opening in the area or the pyramid. claim 1 21. The tanning module according to claim 20 , wherein a flange is provided at the at least one air exhaust opening. claim 20 22. The tanning module according to claim 21 , wherein an air exhaust hose is connected to the flange. claim 21 23. The tanning module according to claim 20 , wherein a reducing disk is present to reduce the size of the air exhaust opening. claim 20 24. The tanning module according to claim 20 , wherein an air exhaust opening is arranged on each of three sides of the pyramid. claim 20 25. The tanning module according to claim 1 , wherein at least one mounting is disposed externally on the housing for electrical connections or components. claim 1 26. The tanning module according to claim 1 , wherein the at least one discoid radiation filter is of rectangular shape. claim 1 27. The tanning module according to claim 26 , wherein the at least one discoid radiation filter has a length and a width ranging from 215 mm 240 mm. claim 26 28. The tanning module according to claim 27 , wherein the at least one discoid radiation filter has a length of 230 mm and a width of 225 mm. claim 27 29. The tanning module according to claim 1 , wherein at least one discoid radiation filter is an interference filter. claim 1 30. The tanning module according to claim 29 , wherein a first discoid radiation filter is present, and plane-parallel thereto a second discoid radiation filter disposed between the radiation emitting area of the reflector and the first discoid radiation filter, wherein the first discoid radiation filter is an interference filter. claim 29 31. The tanning module according to claim 30 , wherein the second discoid radiation filter is an ultraviolet filter or an infrared filter. claim 30 32. The tanning module according to claim 1 , wherein at least one air intake opening is present between the at least one dicoid radiation filter and the housing. claim 1 33. The tanning module according to claim 1 , wherein to protect the at least one discoid radiation filter against breakage at least one touch contact is disposed on the housing, which rests at the at least one radiation filter. claim 1 34. The tanning module according to claim 33 , wherein the touch contact is guided through the reflector perpendicular to the radiation emitting area of the reflector. claim 33 35. The tanning module according to claim 33 , wherein the touch contact is guided through the intake plate perpendicular to the radiation emitting area of the reflector. claim 33 36. The tanning module according to claim 1 , wherein to indicate breaking of the at least one discoid radiation filter at least one touch contact is disposed on the intake plate and rests on the at least one discoid radiation filter. claim 1 37. The tanning module according to claim 1 , wherein a base is provided in the area of the at least one opening in the reflector for the mechanical and electrical connection of the tanning radiator. claim 1 38. The tanning module according to claim 1 , wherein between the at least one discoid radiation filter and the intake plate a cover plate is disposed, which is arranged at a distance from the intake plate and which has a cut-out which in vertical projection onto the at least one discoid radiation filter has the size of the radiation emitting area of the reflector. claim 1 39. The tanning module of claim 1 , wherein the at least one radiation filter is releasable from the housing through a swivelling mechanism. claim 1 40. The tanning module according to claim 1 , wherein the reflector is fastened to the housing only through the intake plate. claim 1 41. The tanning module according to claim 1 wherein at least one air intake opening is present in the housing between the at least one discoid radiation filter and the reflector. claim 1 42. A tanning module comprising: a housing: a tridimensional reflector disposed in the housing; and at least one discoid radiation filter; wherein the at least one discoid radiation filter covers the radiation emitting area of the reflector and is disposed on a first side of the housing, at least one opening being provided in the reflector for the installation and electrical connection of a tanning radiator, and the reflector having its maximum cross section in the plane of the radiation emitting area, characterized in that the housing is configured on a second side opposite the radiation filter in the form of a quadrilateral pyramid with a rectangular base and flattened pyramid apex and that the rectangular base faces in the direction of the at least one radiation filter, wherein an intake plate joins the housing and the reflector on all sides in the area of the radiation emitting area of the reflector, the intake plate having at least one intake opening and also has a cut-out for the reflector which in vertical projection onto the at least one discoid radiation filter has the size of the radiation emitting area of the reflector. 43. The tanning module according to claim 42 , wherein the at least one discoid radiation filter is aligned parallel to the radiation emitting area of the reflector. claim 42 44. The tanning module according to claim 42 , wherein the base of the pyramid is aligned parallel to the at least one discoid radiation filter. claim 42 45. The tanning module according to claim 42 , wherein a rectangular housing wall area adjoins the base of the pyramid. claim 42 46. The tanning module according to claim 42 , wherein at least one mounting is disposed externally on the housing for electrical connections or components. claim 42 47. The tanning module according to claim 42 , wherein the at least one discoid radiation filter is releasable from the housing through a swivelling mechanism. claim 42 48. The tanning module according to claim 42 , wherein at least one air intake opening is present in the housing between the at least one discoid radiation filter and the reflector. claim 42 49. The tanning module according to claim 42 , wherein to indicate breaking of the at least one discoid radiation filter at least one touch contact is disposed on the intake plate and rests on the at least one discoid radiation filter. claim 42 50. The tanning module according to claim 42 , wherein a base is provided in the area of the at least one opening in the reflector for the mechanical and electrical connection of the tanning radiator. claim 42 51. The tanning module according to claim 42 , wherein at least one air intake opening is present between the at least one discoid radiation filter and the housing. claim 42 52. The tanning module according to claim 42 , wherein the intake plate has a rectangular perimeter, that the perimeter of the reflector parallel to the radiation emitting area describes a circle, an ellipse or a polygon, and that the at least one intake opening is disposed in the area of a corner of the intake plate. claim 42 53. The tanning module according to claim 52 , wherein four intake openings are formed in the intake plate and that one each of that four intake openings is disposed in another corner of the intake plate. claim 52 54. The tanning module according to claim 42 , wherein at least one intake opening is enlarged along the sides of the intake plate. claim 42 55. The tanning module according to claim 54 , wherein the intake opening is trapezoidal, the long side of the trapeze facing toward the reflector. claim 54 56. The tanning module according to claim 55 , wherein the long side of the trapeze as well as its opposite side are curved. claim 55 57. The tanning module according to claim 42 , wherein the flattened pyramid apex is formed by a planar part of the housing wall. claim 42 58. The tanning module according to claim 57 , wherein the planar housing wall portion is aligned parallel to the base of the pyramid. claim 57 59. The tanning module according to claim 42 , wherein the flattened pyramid apex is formed by a vaulted housing wall portion. claim 42 60. The tanning module according to claim 59 , wherein the vaulted housing wall portion is configured concavely or convexly with respect to the base of the pyramid. claim 59 61. The tanning module according to claim 42 , wherein the reflector is cupular or tub-shaped. claim 42 62. The tanning module according to claim 61 , wherein the bottom of the capular or tub-shaped reflector is vaulted. claim 61 63. The tanning module according to claim 61 , wherein the bottom, of the capular or tub-shaped bottom of the reflector is made plane-parallel to the at least one discoid radiation filter. claim 61 64. The tanning module according to claim 42 , wherein the at least one discoid radiation filter is of rectangular shape. claim 42 65. The tanning module according to claim 64 , wherein the at least one discoid radiation filter has a length and a width ranging from 215 mm to 240 mm. claim 64 66. The tanning module according to claim 65 , wherein the at least one discoid radiation filter has a length of 230 mm and a width of 225 mm. claim 65 67. The tanning module according to claim 42 , wherein the at least one discoid radiation filter is an interference filter. claim 42 68. The tanning module according to claim 67 , wherein a first discoid radiation filter is present, and plane-parallel thereto a second discoid radiation filter disposed between the radiation emitting area of the reflector and the first discoid radiation filter, wherein the first discoid radiation filter is an interference filter. claim 67 69. The tanning module according to claim 68 , wherein the second discoid radiation filter is an ultraviolet filter or an infrared filter. claim 68 70. The tanning module according to claim 42 , wherein to protect the at least one discoid radiation filter against breakage at least one touch contact is disposed on the housing, which rests at the at least one radiation filter. claim 42 71. The tanning module according to claim 70 , wherein the touch contact is guided through the reflector perpendicular to the radiation emitting area of the reflector. claim 70 72. The tanning module according to claim 70 , wherein the touch contact is guided through the intake plate perpendicular to the radiation emitting area of the reflector. claim 70 73. The tanning module according to claim 42 , wherein a perimeter of the reflector parallel to the radiation emitting area describes a circle, an ellipse, a rectangle or a polygon. claim 42 74. The tanning module according to claim 73 , wherein the reflector is formed of facets and the perimeter of the reflector parallel to the radiation emitting area describes a dodecagon. claim 73 75. The tanning module according to claim 74 , wherein the reflector has a height of 90 mm to 95 mm and the dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) in the range of 210 mm to 230 mm. claim 74 76. The tanning module according to claim 75 , wherein the reflector has a height of 93.6 mm and the dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) of 210 mm. claim 75 77. The tanning module according to claim 74 , wherein the reflector has a height ranging from 110 mm to 125 mm, and the dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) ranging from 170 mm to 200 mm. claim 74 78. The tanning module according to claim 77 , wherein the reflector has a height of 118.7 mm and the dodecagon has in the plane of radiation emitting area a maximum diameter (corner to corner) of 184 mm. claim 77 79. The tanning module according to claim 74 , wherein the reflector has a height ranging from 75 mm to 90 mm, and the dodecagon has in the plane of the radiation emitting area a maximum diameter (corner to corner) ranging from 205 mm to 235 mm. claim 74 80. The tanning module according to claim 79 , wherein the reflector has a height of 118.7 mm and the dodecagon has in the plane of radiation emitting area a maximum diameter (corner to corner) of 184 mm. claim 79 81. The tanning module according to claim 42 , wherein the housing has at least one air exhaust opening in the area of the pyramid. claim 42 82. The tanning module according to claim 81 , wherein a reducing disk is present to reduce the size of the air exhaust opening. claim 81 83. The tanning module according to claim 81 , wherein an air exhaust opening is arranged on each of three sides of the pyramid. claim 81 84. The tanning module according to claim 81 , wherein a flange is provided at the at least one air exhaust opening. claim 81 85. The tanning module according to claim 84 , wherein an air exhaust hose is connected to the flange. claim 84 |
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042141672 | abstract | A device for protection of gonads in x-ray diagnosis is disclosed. The device comprises a retaining frame for fastening to the collimator of an x-ray diagnosis instrument by means of insert strips. The retaining frame receives a movable support strip of material that is permeable to x-rays and which supports a member of material that is impermeable to x-rays. Insert strips for fastening the retaining frame to a collimator of an x-ray instrument are disposed on the retaining frame so as to be movable with respect to each other so that the protective device can be fastened to different x-ray diagnosis instruments. |
abstract | The present invention is related to device for selecting one of several triggering apparatuses, which are simultaneously connectable to said device. The triggering apparatuses are arranged for producing each a triggering signal to enable/disable one or more components of a radiation treatment apparatus. The triggering signals depend on detected parameters. The device is configured to receive triggering signals from the several triggering apparatuses, when all of the apparatuses are connected to the device, to receive a selection of one of the triggering apparatuses, to generate a universal triggering signal for the one or more components on the basis of said received triggering signal from the selected triggering apparatus, and to send the universal triggering signal to the one or more components. |
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052221130 | claims | 1. An X-ray microscope for investigating a specimen, the X-ray microscope comprising: a pulsed X-ray radiation source for supplying an intensive line radiation and defining an optical axis along which the line radiation travels; reflecting condenser means for focussing the radiation on the specimen; an X-ray detector mounted on said axis downstream of said reflecting condenser means; and, zone plate means for imaging the specimen on said detector with high resolution. focussing the radiation of a pulsed X-ray radiation source on a specimen by means of a reflecting condenser; generating an image of the specimen with a triggered pulse of X-radiation; imaging the specimen on a camera with a zone plate; and, reading out the camera synchronously with the pulsed X-ray radiation source after a generated pulse of said radiation. 2. The X-ray microscope of claim 1, said reflecting condenser means including a reflecting condenser defining a reflective surface; and, a multiple layer disposed on said surface for increasing reflectivity. 3. The X-ray microscope of claim 2, said reflecting condenser being configured to focus said X-radiation onto the specimen at grazing incidence to ensure total reflection. 4. The X-ray microscope of claim 2, said reflecting condenser being a segment of an ellipsoid. 5. The X-ray microscope of claim 1, said X-ray source being a plasma-focus source. 6. The X-ray microscope of claim 1, said zone plate means being a phase zone plate. 7. The X-ray microscope of claim 1, further comprising foil means for protecting said reflecting condenser means and being permeable to said radiation. 8. The X-ray microscope of claim 1, said reflecting condenser means being adapted to image said source into the specimen. 9. The X-ray microscope of claim 1, said X-ray detector being a semiconductor camera. 10. The X-ray microscope of claim 1, further comprising electronic means for synchronizing said detector and said X-ray source so as to cause an image to be read out of said detector after a pulse of said radiation. 11. A method of generating microscopic images of high resolution in the light of X-radiation, the method comprising the steps of: |
description | This application is a continuation-in-part of, and claims domestic priority benefits under 35 U.S.C. §120 to, the following and commonly assigned U.S. patent applications: Ser. No. 10/321,441 to William Earl Russell, II, et al., filed Dec. 18, 2002 now U.S. Pat. No. 7,222,061 and entitled “Method and Arrangement for Developing Rod Patterns in Nuclear Reactors; Ser. No. 10/401,602 to William Earl Russell, II et al., filed Mar. 31, 2003 now U.S. Pat. No. 7,231,333 and entitled “Method and Arrangement for Developing Core Loading Patterns in Nuclear Reactors; and Ser. No. 10/325,831 to David J. Kropaczek et al., filed Dec. 23, 2002 now U.S. Pat. No. 7,200,541 and entitled “Method and Arrangement for Determining Nuclear Reactor Core Designs”. The entire contents of each of these applications are hereby incorporated by reference herein. 1. Field of the Invention Example embodiment(s) of the present invention are related in general to a method and system for designing a nuclear reactor core for increased power operations. 2. Description of the Related Art A core of a nuclear reactor such as boiling water reactor (BWR) or pressurized water reactor (PWR) has several hundred individual fuel bundles of fuel rods (BWR) or groups of fuel rods (PWR) that have different characteristics. These bundles (or fuel rod groups) are arranged so that interaction between rods within a fuel bundle, and between fuel bundles satisfies all regulatory and reactor design constraints, including governmental and customer-specified constraints. Additionally for a BWR, the control mechanisms, e.g. rod pattern design and core flow, must be determined so as to optimize core cycle energy. Core cycle energy is the amount of energy that a reactor core generates before the core needs to be refreshed with new fuel elements, such as is done at an outage. In the case of a BWR, for example, the number of potential fuel bundle arrangements within the core and individual fuel rod types within a bundle may be in excess of several hundred factorials. From these many different possible configurations, only a small percentage of fuel bundle designs, defined by an arrangement of fuel rod types, may satisfy all applicable design constraints for a particular core of a nuclear reactor. Further, only a small percentage of these fuel bundle designs, which do satisfy all applicable design constraints, are economical. Conventionally, a typical fuel bundle useable in a BWR core may include between about 10 to in excess of 30 different fuel rod types therein, and hence different fuel bundle assembly configurations. This is undesirable, in that the greater the number of different rod types, the greater the manufacturing complexities and cost, which may result in higher bundle costs to the consumer. Traditionally, rod pattern, fuel bundle and/or core design determinations have been made on a trial and error basis. Specifically, and based on only the past experience of the engineer or designer, in designing a particular design an initial design was identified. The initially identified design, such as a particular fuel bundle design for a core, was then simulated in a virtual core by computer. If a particular design constraint was not satisfied, then the arrangement was modified and another computer simulation was run. Many weeks of resources typically were required before an appropriate design was identified using the above-described procedure. For example, one conventional process being used is a stand-alone manual process that requires a designer to repeatedly enter reactor plant specific operational parameters into an ASCII text file, which may serve as an input file. Data entered into the input file may include the configuration of fresh and exposed fuel bundle placements, blade notch positions of control blades (if the evaluated reactor is a boiling water reactor (BWR)), soluble boric acid concentration (if a PWR, for example), core flow, core exposure (e.g., the amount of burn in a core energy cycle, measured in mega-watt days per short ton (MWD/st), etc. A Nuclear Regulatory Commission (NRC) licensed core simulation program, which may be embodied as a software program running on a suitable computer, for example, reads the resulting input file and outputs the results of the simulation to a text or binary file. A designer then evaluates the simulation output to determine if the design criteria have been met, and also to verify that no violations of margins to thermal limits have occurred. Failure to meet design criteria (i.e., violations of one or more limits) requires a manual designer modification to the input file. Specifically, the designer would manually change one or more operational parameter(s) and rerun the core simulation program. This process may be repeated until a satisfactory design was achieved. This process is time consuming. The required ASCII text files are laborious to construct, and often are error prone. The files are fixed-format and extremely long, sometimes exceeding five thousand or more lines of code. A single error in the file results in a crash of the simulator, or worse, results in a mildly errant result that may be hard to initially detect, but will profligate with time and iterations to perhaps reduce core cycle energy when placed in an actual operating nuclear reactor core. Additionally, no assistance is provided via the manual iterative process in order to guide a designer toward a more favorable solution. In the current process, the responsible designer or engineer's experience and intuition are the sole means of determining a design solution. An example embodiment is directed to a method of designing a nuclear reactor core for uprated power operations. In the method a set of constraints are inputted to be satisfied for uprated power operations, and a test reactor core design is generated based on the constraints. One or more automated tools may be selected from a set of automated tools to evaluate the test core design against the constraints. The selected tool may then be operated. Operation of the selected automated tool includes simulating reactor operation with the test core design, based on the constraints, to produce a plurality of outputs, comparing the outputs against the constraints, and providing data indicating constraints that were violated by the test core design during the simulation, based on the comparison. One or more of the automated tools are iterated until a test core design meets all constraints for uprated power operations, thereby representing an acceptable power uprate core design. Another example embodiment is directed to a system for designing a core of a nuclear reactor for increased power output. The system includes a memory for storing a test reactor core design to be evaluated for a reactor plant of interest, an interface receiving a set of constraints to be satisfied for uprated power operations in the core of the plant of interest, and a processor iterating one or more automated tools useable to evaluate the test design against the constraints. The interface enables selection of one or more of the automated tools to simulate reactor operation of the test design, based on the constraints, to produce a plurality of outputs. The processor iterates the selected automated tool to compare the outputs against the constraints and to generate data via the interface indicating any constraints violated by the test core design during the simulation, based on the comparison. Another example embodiment is directed to a core for a nuclear reactor plant which enables the plant to operate at greater than 100% of its currently-licensed power level for one or more energy cycles. The core has been loaded in accordance with a power uprate core design determined using a set of automated tools employing an optimization process to generate a core design for the plant which satisfies operating limits for operation at greater than 100% of its currently-licensed power level. The automated tools include a rod pattern design tool, a core loading pattern design tool and a fresh fuel bundle type design tool to generate the core design. As will be described in more detail below, the example embodiments are directed to a method and system for designing a nuclear reactor core for increased power operations. The system may include a graphical user interface (GUI) and a processing medium (e.g., software-driven program, processor, application(s), etc.) to enable a user of the program to virtually simulate a test initial reactor core design in an uprated power environment. Via the GUI, the user inputs constraints or limits related to requirements for operating a reactor core above 100% of its currently-licensed power level, and generates an initial test reactor core design. The user then may select one or more automated design tools to modify the test reactor core design in order to satisfy the input constraints. Each automated design tool may be embodied as a software-driven application or program to develop and/or optimize certain core features or parameters, including, but not limited to, an automated tool to modify the rod pattern (control blade pattern) for the core to meet uprated power requirements; an automated tool to modify a core loading pattern in the core to meet uprated power requirements, and an automated core design tool to optimize and/or determine desired unique fresh fuel bundle types and placement in the core, etc. For each tool, the user may initiate a reactor simulation (e.g., 3D simulation using simulation codes licensed by the NRC) of the modified core design, and view outputs from the simulation. Each automated tool utilizes an objective function to determine how closely a simulated modified core design meets the constraints. The objective function is a mathematical equation that incorporates the constraints or limits and quantifies the core design's adherence to the limits. The GUI presents a graphic illustration of which outputs violate which constraints, as determined by the objective function. Via the GUI, the user may then modify the core design. This may constitute a change in bundle design, control rod placement, exposed or fresh fuel placement, etc., and repeating the simulation to evaluating if there is any performance improvement in the modified design. The modifying, simulating and evaluating simulation outputs against the constraints may be iteratively repeated until a core design satisfies all limits (inclusive of limits for uprated power operations) so as to operate a reactor loaded based on the acceptable power uprate core design at increased power, or within a margin to given limits(s) that are acceptable to the user and/or customer. FIG. 1 illustrates a system for implementing the example method in accordance with an exemplary embodiment of the invention. Referring to FIG. 1, system 1000 may include an application server 200, which may serve as a central nexus of an accessible website, for example. The application server 200 may be embodied as any known application server software, such as Citrix MetaFrame Presentation server, for example. Application server 200 may be operatively connected to a plurality of calculation servers 400, a cryptographic server 260 and to a memory 250. Memory 250 may be embodied as a relational database server, for example. A plurality of external users 300 may communicate with application server 200 over a suitable encrypted medium such as an encrypted 128-bit secure socket layer (SSL) connection 375, although the present invention is not limited to this encrypted communication medium. An external user 300 may connect to the application server 200 over the Internet or from any one of a personal computer, laptop, and personal digital assistant (PDA), etc., using a suitable interface such as a web-based Internet browser. Further, application server 200 is accessible to internal users 350 via a suitable local area network connection (LAN 275), so that internal users 350 have access over an intranet for example. Hereafter for reasons of brevity, ‘user’ is employed generically to denote any of an external user 300, internal user 350 or other designer accessing system 1000. For example, the user may be any of a representative of a nuclear reactor plant accessing the website to determine a desired core design for his or her nuclear reactor, and/or a vendor hired by a reactor plant site to develop an uprated core design for their plant by using the example method and system of the present invention. The application server 200 provides a centralized location for users to access an application. Essentially, each user application session may be running on the server but displayed locally to the user access device (e.g., PC) allowing the user to interact with the application. However, this means of deployment is provided as an exemplary embodiment, and does not limit a given user from running the application locally on their access device. The application is responsible for directing all calculations and accessing of data in order to calculate objective function values, and for the creation of suitable graphical representations of various features of a core design that a user may desire to review. The graphical information is communicated over the 128-bit SSL connection 375 or LAN 275, to be displayed on a suitable display device of the user. FIG. 2 illustrates an application server 200 associated with the system of FIG. 1. Referring to FIG. 2, application server 200 utilizes a bus 205 to connect various components and to provide a pathway for data received from the users. Bus 205 may be implemented with conventional bus architectures such as peripheral components interconnect (PCI) bus that is standard in many computer architectures. Alternative bus architectures such as VMEBUS, NUBUS, address data bus, RDRAM, DDR (double data rate) bus, etc. could of course be utilized to implement bus 205. Users communicate information to application server 200 over a suitable connection (LAN 275 and/or via network interface 225) to communicate with application server 200. Application server 200 may also include a host processor 210, which may be constructed with conventional microprocessors such as currently available PENTIUM processors. Host processor 210 represents a central nexus from which all real time and non-real functions in application server 200 are performed, such as graphical-user interface (GUI) and browser functions, directing security functions, directing calculations such as calculation of the objective functions for various limits, etc., for display and review by the user. Accordingly, host processor 210 may include a GUI 230 which may be accessed through the use of a browser. Browsers are software devices which present an interface to, and interact with, users of the system 1000. In the example embodiment, a browser in conjunction with a Citrix ICA client (part of the commercially available Citrix MetaFrame Access Suite software) may be responsible for formatting and displaying the GUI 230. Browsers are typically controlled and commanded by the standard hypertext mark-up language (HTML). However, the application being presented or “served” to the user which allows the user to control decisions about calculations, displayed data, etc. may be implemented using C#, Java or Visual Fortran or any combination thereof. In addition, other well-known high-level languages maybe incorporated in the applications implementation (e.g., C, C++, etc.). All of these languages may be customized or adapted for the specific details of a given application or implementation, and images may be displayed in the browser using well known JPG, GIF, TIFF and other standardized compression schemes, other non-standardized languages and compression schemes may be used for the GUI 230, such as XML, ASP.NET, “home-brew” languages or other known non-standardized languages and schemes. Application server 200 through Network I/F 225 may be operatively connected to a cryptographic server 260. Accordingly, application server 200 implements all security functions by using the cryptographic server 260, so as to establish a firewall to protect the system 1000 from outside security breaches. Further, cryptographic server 260 secures external access to all personal information of registered users. Application server 200 may be also operatively connected to a plurality of calculation servers 400. The calculation servers 400 may perform some or all the calculations required to process user entered data, direct simulation of a core design, calculate values for comparison as to be described in further detail below, and to provide results which may be displayed via, the GUI 230, and presented by the application server 200. The calculation servers 400 may be embodied as WINDOWS 2000 servers, for example, although other hardware (e.g., Alpha, IA-64) and platforms (e.g., Linux, Unix) are possible. More particularly, the calculation servers 400 may be configured to perform a multitude of complex computations which may include, but are not limited to, configuring the objective function and computing objective finction values, executing a 3D simulator program to simulate reactor core operation on a particular core design to be evaluated, and to generate outputs from the simulation, providing results data for access and display by a user via GUI 230, and iterating an optimization routine as to be described in further detail below. FIG. 3 illustrates an exemplary database server 250 in accordance with an exemplary embodiment of the invention. Memory or database server 250 may be a relational database such as an Oracle relational database. Relational database server 250 may contain a number of subordinate databases that handle all the necessary data and results in order to implement the method of the present invention. For example, relational database server 250 may include storage areas which contain subordinate databases such as limits database 251, which is a database that stores all the user input limits and/or design constraints for all core designs that are evaluated for a particular nuclear reactor. There may also be a fresh fuel bundle design database 252 which may include a palette or plurality of different fresh fuel bundle designs that have been previously created, modeled and stored therein. Additionally, relational database server 250 may include a queue database 253, which stores parameters for a core design that are to be simulated in the 3D simulator, and a historical core design database 254, which includes historical reactor core loading pattern designs that may be selected in generating a reference core design that is most consistent with defined user-input limits. Further, relational database 250 may include a loading templates database 256, which may store a plurality of different loading templates for fresh and/or exposed fuel bundles, and a fuel pool database 258 which may store a plurality of different exposed and/or fresh fuel bundle metrics for each of a plurality of different exposed and/or fresh fuel bundles, for example. Simulator results may be stored in simulator results database 255. The simulator results database 255 (and limits database 251) may be accessed by the calculation servers 400 in order to calculate a number of objective function values that are applicable to a particular core design. The objective function values may be ranked for a particular core design (due to modification of one or more of the rod (control blade) pattern, core loading pattern and/or N unique fresh fuel bundle types to be used in the eventual acceptable core design) that has been inserted into a virtual core to be simulated, so as to determine whether or not the given design meets certain user-input limits or constraints, for example. These objective function values may be stored in an objective function values database 257 within relational database server 250. A 3D simulator input parameters database 259 may also be included within relational database server 250. Database 259 may include the positions of control blades and reactor operating parameters for all exposure steps. As the calculation servers 400 is operatively connected to, and may communicate with, relational database server 250, each of the subordinate databases described in FIG. 3 may be accessible to one or more calculation servers 400. FIG. 4 is a process flow diagram for illustrating a method of designing a nuclear reactor core for uprated power operations in accordance with an example embodiment of the invention. Referring to FIG. 4, each of the blocks 500 through 1100 describe processing functions of an example method of designing a nuclear reactor core for uprated power operations, beginning with an inputs block 500. A given block may represent a number of functions and/or software-driven processes that may be performed by system 1000, such as directly from input by the user via GUI 230, automatically by the host processor 210 and/or calculation servers 400, and/or under control of the host processor 210 in the application server 200 based on commands or selections made by the user. The functions performed at inputs block 500 enable a set of constraints (referred to herein also as “limits”) that need to be satisfied for uprated power operations to be inputted into the evaluation for determining an acceptable power uprate core design. These limits, including margins to thermal limits and an uprate power limits flow window which sets higher reactor power, power density and mass flow rate limits than what is rated for the plant being evaluated, may be stored in limits database 251. Core loading block 600 generates a test reactor core design based on the limits (constraints) input at block 500. As will be seen in more detail below, the functions performed by system 1000 as indicated in the core loading block 600 include determining the number of fresh fuel bundles to be used in the test core design, selecting a desired fuel loading template from the loading templates database 256, and populating the selected template with fresh fuel bundles for loading and exposed fuel for reloading so as to arrive at a ‘virtual nuclear reactor core’ that is loaded in accordance with the core loading block 600. Automated tools block 700 includes processing functionality to enable the user 300 to select and iteratively run one or more automated tools on the initial test core design generated in block 600. These automated tools include a modify rod (control blade) pattern tool, a modify core loading pattern tool, and a unique fresh fuel bundle type design tool (also known as an “N-streaming”) tool. Each of the rod pattern, core loading pattern and unique fresh fuel bundle type design tools may be iterated sequentially and/or together by the user and may provide feedback to one or both of the other tools, until all rod (control blade), exposed fuel and/or fresh fuel changes have been exhausted in the test core design and/or a given “candidate” modified test core design satisfies each of the limits or constraints set for uprated power operations. For each automated tool, operation or iteration of the selected automated tool to evaluate the test core design against the input limits or constraints includes, at a minimum, performing a simulation of reactor operation with the test core design, based on the constraints, in order to produce a plurality of simulation results (also referred to as “outputs”). The outputs are compared against the constraints. Data may be provided to the user 300 which indicates constraints that were violated by the test core design during the simulation, based on the comparison. One, some or all of the automated tools are implemented and iteratively repeated until a test core design is determined which meets all constraints for uprated power operations. This design thus represents an acceptable power uprate core design. In an example, and as to be explained in more detail below, the comparison noted above may include configuring an objective function to evaluate the outputs. By configuring an objective function to evaluate the outputs, objective function values may be generated for each output using the objective function. The objective function values representing the outputs may then be evaluated based on the constraints to determine which of the outputs violate a limit. Each automated tool thus utilizes an objective function to determine how closely a simulated test core design meets the limits or constraints. As previously discussed, the objective function is a mathematical equation that incorporates the constraints or limits and quantifies the test core design's adherence to the limits. The GUI 230 presents a graphic illustration of which outputs violate which constraints, as determined by the objective function. Via the GUI 230, the user may then modify the test core design, to create what is also known herein as a modified or ‘derivative reactor core design’. This may constitute a change in rod pattern design including control rod placement, exposed fuel placement, fresh fuel bundle placement, etc., and repeating core simulation to evaluate if there is any performance improvement in a given derivative test core design. In an example, and based on which particular constraints or limits were violated by a particular derivative core design, a specific automated tool may be selected by the user for a subsequent iteration. The modifying, simulating and comparing functions (i.e., evaluating simulation outputs against the constraints) may be iteratively repeated within a given automated tool process until a given derivative test core design satisfies all limits to operate the reactor at uprated power. Data related to the acceptable reactor core design may be reported to the user at 1100. In an example, this data may be presented in graphical format via GUI 230 to illustrate to a user how to load and run the selected reactor plant so as to satisfy the constraints and operate at the increased uprated power level. Referring again to FIG. 4, after all rod (control blade), exposed fuel and fresh fuel changes have been exhausted (output of 800 in FIG. 4 is ‘YES’), if all the limits have been satisfied (output of Refinements block 900 is ‘YES’), no further refinements need to be made to the selected exposed fuel or fresh fuel bundles the user has chosen for the test core design; this design is then reported at report block 1100 as the acceptable power uprate core design. However, if after all rod (control blade), exposed fuel and fresh fuel bundle changes have been exhausted and one or more limits from block 500 are still not satisfied, the user may be directed back to checking whether or not all exposed fuel bundles in the inventory have been used and if not, which returns the processing back to functions performed in core loading block 600, with a recommendation to modify the exposed fuel selection in the loading template. If all exposed fuel bundles from the inventory have been used and one or more limits from block 500 are still not satisfied, the user may be directed back to the processes in core loading block 600 with a recommendation to modify the fuel loading template in order to change the locations in the template for insertion of fresh fuel bundles. Once these changes have been made, a modified or derivative core design may be re-evaluated using one, some or all of the automated tools of block 700 until all limits have been satisfied and/or are within an acceptable margin thereto. Of note, the processing described in blocks 700 and 900 may be done using one or more optimization techniques to optimize certain parameters such as rod (control blade) positions, exposed and fresh fuel bundle positions or placement, core flow, and/or where to make sequence changes, so that the iterative simulation, evaluation, modification of core design including the possibility of making additional fresh and exposed fuel locations changes in the loading template may be done automatically. Once the user has received the data and design parameters corresponding to the acceptable power uprate core design at block 1100, the user or designer may optionally iterate several processes which are described in an implementation block 1200. The dotted lines of block 1200 indicate that the processes invoked by the user using system 1000 may be optional. Implementation Block As shown in block 1200 of FIG. 4, optionally for each desired energy cycle in a given reactor plant being evaluated, the user may access updated exposure accounting data (“core tracking”) directly from the reactor plant of interest for a given cycle. This may be done via an interface between system 1000 that is in communication with a process computer of the plant of interest (plant being modeled), with data being stored in database 250. This exposure accounting process is referred to as a core tracking function in 1200a. This almost real-time, actual plant data may be parsed to retrieve the desired data needed by the user or designer in order to determine whether any limits and/or margins to such limits or constraints for the power uprate operation need to be changed based on actual plant data (comparing the actual plant data to estimated data in order to determine any standard deviation (sigma) and/or bias to apply to the estimated margins to existing limits and/or to the limits themselves. Accordingly, at 1200b the user may revise thermal and/or reactivity limits, and/or margins to these limits, based on updated exposure accounting data to determine new limits for the given cycle being evaluated. Using current actual plant conditions from (1200a) and any new or revised limits or margins to limits determined from (1200b), the customer, user or designer may (1200c) perform an “online” plant optimization. This would entail setting limits and operating conditions for the power uprate core design based on (1200a) and (1200b) above, then iterating the automated rod pattern tool for the selected uprated core design reported at 1100. This iteration of the automated rod pattern tool may include optimizing rod positions, core flow, and where to make sequence changes, based on the update actual plant conditions from (1200a) and based on the new or revised limits calculated from (1200b). As discussed in block 700, the rod pattern tool may be repeatedly iterated in (1200c), inclusive of repetitive simulation, comparison and modification of the power uprate core design, to determine a modified power uprate core design which satisfies the limits for the cycle being evaluated. Once all desired energy cycles have been evaluated (output of 1300 is ‘YES’), the processes described in the optional implementation block 1200 may be terminated. In an example, data presented to the user resulting from the simulation and comparison of outputs to limits in any of the automated tools performed as described in block 700 or 1200 may include procedural recommendations. For example, such procedural recommendations may be provided as feedback to the user as an output of the modify rod pattern tool process at block 1200c, for example. As to be seen below, each automated tool may provide associated procedural recommendations to assist the user or designer with any modifications that need to be made to the test core design in order to achieve a desired core design for uprated power. Although the individual modifications may be left to the desires of the user, these procedural recommendations may be provided in the form of a pull down menu, for example, and may divided into categories such as energy beneficial moves, energy detrimental moves, and converting excessive margin (to thermal limits) into additional energy. One technique may be to address problems with meeting limits using energy beneficial moves rather than energy detrimental moves. Even if the power uprate core design meets all of the limits (client-inputted plant specific constraints, design limits, thermal limits, etc.) the user may verify that any excessive margin to a particular limit could be converted into additional energy. An overview of the example methodologies having been described, various processes within the blocks in FIG. 4 are now described in more detail. Inputs Block 500 Initially, the designer or user selects a reactor plant for evaluation from a suitable drop down menu of plants and defines a set of limits or constraints to be satisfied by the selected plant in order to operate at uprated power, i.e. a power in excess of rated capacity, such as 3600 megawatts (uprated power) for a plant that is rated at 3400 MW. In other words, limits or constraints are defined which are to be satisfied in determining an acceptable reactor core design to be used in the selected plant for uprated power operations. For example, the user may be prompted to enter uprate power thermal limits and a flow window for uprated power operations. These will be used to modify the plant operational conditions for determining the power uprate core design. These limits may be related to key aspects of the design of the particular reactor being evaluated for uprated power operations and design constraints of that reactor. The limits may be applicable to variables that are to be input for performing a simulation of a test reactor core design that is to be generated, and/or may include limits or constraints applicable only to the results of the simulation. For example, the input limits may be related to client-inputted reactor plant specific constraints and core performance criteria for uprated power operations. Limits applicable to the simulation results may be related to one or more of operational parameter limits used for reactor operation, core safety limits, margins to these to these operational and safety limits and the other client-inputted reactor plant specific constraints. FIG. 5 is a screen shot to illustrate example client-inputted plant specific constraints. These may be understood as limits or constraints on input variables to a simulation and limits or constraints on the simulation results. Referring to FIG. 5, there is listed a plurality of client-inputted plant specific constraints as indicated generally by the arrow 505. For each constraint, it is possible to assign a design value limit, as indicated by column 510. Although not shown in FIG. 5, the limits or constraints include margins to thermal limits and an uprate power limits flow window which sets higher reactor power, power density and mass flow rate limits for the plant being evaluated. Core Loading Block 600 The process of selecting a fuel loading template, selecting the fresh fuel for loading and selecting the exposed fuel to reload is described in detail in co-pending and commonly assigned U.S. patent application Ser. No. 10/678,170 to David J. Kropaczek et al., filed Oct. 6, 2003 and entitled “Method and Apparatus for Facilitating Recovery of Nuclear Fuel from a Fuel Pool”. The relevant portions describing these processes as described below in FIGS. 7-10 are incorporated in their entirety herein. However, prior to selecting the fuel loading template, a fresh fuel bundle count is determined for the initial test reactor core design. FIG. 6 is a flowchart describing the determination of a fresh fuel bundle count for the initial test reactor core design in accordance with an exemplary embodiment of the invention. The selection of a fresh fuel bundle count is done prior to the selection of the fuel loading template so as to provide the inventory of fresh fuel bundles to be used in designing a reactor core design for uprated power operations. Initially, a check is performed (602) to establish whether prior iterations on a test fresh fuel loading pattern have occurred. If this is a first iteration, e.g., no previous test fresh fuel loading pattern has been analyzed, information on past cycles or similar plants may be used to provide a basis for an initial test fresh fuel loading pattern (603). For example, an initial test fresh fuel loading pattern may be selected from a core loading pattern design used for a similar core in a previous simulation, selected based on a core loading pattern design from a reactor that is similar to the reactor being evaluated, and/or from an actual core loading pattern design used in an earlier core energy cycle in the reactor plant being evaluated, for example. These designs may be accessed from the historical fuel cycle designs database 254, for example. If past iterations have been performed (the output of 602 is “NO”) the total energy content of the core, using an established core loading pattern that conforms to the input limits, may be calculated, and a difference from a desired/required energy content may be defined (604). This may also be done using a fresh fuel loading pattern from 603, also accounting for the inputted limits, if this is the first iteration. This energy “delta” is the difference in the required energy for the next, future cycle as compared to the most recent End-of-Cycle (EOC). For additional iterations, the delta may be reduced as the difference between the actual energy and desired energy is reduced. Furthermore, negative delta energies imply that the resulting energy is greater than the desired energy and is desirable. The difference in energy should be supplied by the fresh fuel assemblies, which would also be part of the fresh fuel loading pattern for loading the core of the reactor, to be loaded at a next scheduled outage, for example. Typical rules of thumb exist that can help select the number of additional bundles needed (or number of bundles that must be removed) in order to obtain the desired target energy. For example, in a BWR reactor with 764 bundles, it is commonly believed that four (4) bundles are worth approximately 100 MWD/st of cycle length. Therefore, if the resulting energy is over 100 MWD/st longer than the desired energy, four fresh bundles could be removed. Similarly, if the resulting energy more than 100 MWD/st shorter than the desired energy, four additional fresh bundles should be added. The user should select (605) the number of fresh fuel bundles needed to make up for the energy difference. This may be done by accessing a “palette” of previously modeled and stored fresh fuel bundle designs from fresh fuel bundle design database 252, or the user may create specific fresh fuel bundles from a database of bundle types, for example. After the number or count of fresh bundles to be used in the test core design is determined, core loading symmetry is identified (606). Some plants may require quadrant loading symmetry or half-core loading symmetry, for example. GUI 230 may be used to access a plant configuration webpage, which may enable the user to select a “model size”, e.g., quarter core, half core, or full core, for evaluation in a subsequent simulation. Additionally, a user may select a core symmetry option (e.g., octant, quadrant, no symmetry) for the selected model size, by clicking on a suitable drop down menu and the like. By selecting “octant symmetry”, the user can model the reactor assuming that all eight (8) octants (where an octant is a group of fuel bundles for example) are similar to the modeled octant. Consequently, simulator time may be generally decreased by a factor of eight. Similarly, by selecting “quadrant symmetry”, the user can model the reactor assuming each of the four (4) quadrants is similar to the modeled quadrant. Hence, the simulator time may be generally decreased by a factor of four. If asymmetries in bundle properties prevent octant or quadrant symmetry, the user can also specify no symmetry. Once the fresh fuel bundle count and symmetry is determined, the user may access a desired fuel loading template 607 from the loading templates database 256. FIG. 7 illustrates a quarter-core screen shot of a partially completed loading template using a loading map editor 601 of the present invention. When the loading map editor is initially run, the user has the option via a file menu 608 to access a previously created template (from loading templates database 256) or to begin a new template. In an example, the loading map editor 601 requests the user to identify the reactor for which the template is being created. The loading map editor 601 retrieves the geometry of the identified nuclear reactor from the relational database 250 (such as loading templates database 256, and displays fuel bundle field 611 of the appropriate size based on the retrieved plant characteristics with the rows and columns numbered (such as with the fuel bundle position Row 6, Column 3 in FIG. 7). Within the fuel bundle field 611, the user may then, for example, using a suitable input device (i.e., mouse, touch pad, etc.) via his computer, GUI 230 and application processor 200 to click on the fuel bundle positions 612 in the array of possible fuel bundle positions to identify the type (fresh, reinsert (exposed fuel), or locked) and grouping of the actual fuel bundle in that position. As shown on the right side of FIG. 7, the loading map editor 601 provides several tools for performing this assignment task, including Load Type 613, Bundle Grouping 617 and Numbering Mode 618 tools. Under the Load Type 613 tool, the loading map editor 601 includes a Fresh button 614, a Reinsert button 615 and a Locked button 616. The Fresh, Reinsert and Locked buttons 614, 615 and 616 correspond to fresh, reinsert and locked fuel bundle categories. The user, for example, clicks on the desired button to choose the desired category and then clicks on the fuel bundle position 612 in the fuel bundle field 611 to assign that category to the fuel bundle position 612. The fresh fuel bundle category indicates to insert fresh fuel bundles, i.e., bundles that have not been exposed. The loading map editor then displays “F” and a number “N” at the bottom of the fuel bundle position 612. The “F” indicates the fresh fuel bundle category, and the number “N” indicates the Nth fresh bundle type 612. The loading map editor 601 maintains a count of the number of fuel bundle types assigned to the core. Multiple bundle positions can be assigned the same bundle type by specifying the same “F” and “N” value for each position. The locked fuel bundle category indicates that a fuel bundle currently occupying an associated fuel bundle position in an actual nuclear reactor core is to remain in that position in creating a nuclear reactor core loading map. The loading map editor 601 displays “L” and a number “N” in the fuel bundle position 612 when the locked fuel bundle category is assigned. The “L” indicates the locked fuel bundle category, and the number “N” indicates the Nth locked bundle group. The reinsert fuel bundle category indicates where to insert an exposed fuel bundle. The loading map editor displays only a number “N” in the fuel bundle position 612 when the reinsert fuel bundle category is assigned. The number indicates a priority of the fuel bundle position 612. In an example, the loading map editor may display the fuel bundle positions 612 in a color associated with the assigned category. For example, fresh bundles are displayed in blue, locked are displayed in yellow, and reinserted exposed fuel bundles are displayed in violet. Under the Bundle Grouping 617 heading, the loading map editor includes symmetry buttons “1”, “2”, “4” and “8”. As noted above in FIG. 6, the user may have already selected a desired core symmetry option (e.g., octant, quadrant, no symmetry, etc.) for the selected reactor, by clicking on a suitable drop down menu and the like. However, the bundle grouping tool 617 allows the user to select the symmetry for given bundle positions 612, i.e., on a fuel bundle position 612 basis. The Numbering Mode tool 618 includes an Automatic button 619 and a Manual button 620. Choosing between an automatic numbering and a manual numbering is only permitted when the Reinsert button 615 or Fresh button 616 has been selected, and is generally inapplicable if the Locked button 616 is selected. With the Automatic button 619 selected, the loading map editor 601 maintains a count of exposed fuel bundles and assigns the count plus one to the next fuel bundle position 612 assigned the reinsert fuel bundle category. The assigned number is displayed at the bottom of the fuel bundle position 612. The loading map editor 601 maintains a count of the fresh bundle types as well; thus when a fuel bundle position 612 is assigned a fresh bundle category count plus one, referred to above as N, is assigned to that position. “F” and the value of N are displayed at the bottom of the fresh fuel bundle position. When the Manual button 620 is selected, the loading map editor maintains the count of the number of fuel bundle positions 612 assigned the reinsert fuel bundle category, but does not assign numbers to the fuel bundle positions 612. Instead, the user may position a cursor in the fuel bundle position 612 and enter the number manually. The assigned numbers in each bundle position 612 represent assigned priorities indicating an order for loading exposed fuel bundles based on an attribute of the exposed fuel bundles. The attributes include, but are not limited to, K infinity (which is a well-known measure of the energy content of the fuel bundle, exposure of the bundle (which is accumulated mega-watt days per metric ton of uranium in the bundle), residence time of the bundle (which is how long the bundle has been resident in the nuclear reactor core), etc. In one exemplary embodiment, the shade of the color associated with the reinserted fuel bundle positions varies (lighter or darker) in association with the assigned priority. The loading map editor 600 also provides several viewing options via a view menu 610 and a zoom slide button 621. Adjusting the zoom slide button 621 by clicking and dragging the zoom slide button 621 to the left and the right decreases and increases the size of the displayed fuel bundle field 611. Under the view menu 610, the user has the option to view a single quadrant of the template, or a full core view of the template. Additionally, the user can control whether certain template attributes are displayed. Specifically, the view menu 610 includes the options of displaying the following in the loading template: control blades, bundle coordinates, core coordinates, etc. As discussed above, instead of creating a new template, a previously created template may be viewed and, optionally, edited using edit button 609. Using the file menu 608, the user selects an “open” option. The loading map editor 600 then displays the accessible templates stored in fuel loading templates database 256. The user then selects an accessible template, for example, by clicking on one of the accessible templates. The loading map editor 600 will then display the chosen template. The user may then edit the selected template. For example, for a given selected template, the user may select the fresh fuel and exposed fuel bundles desired for the template in order to generate the test reactor core design. The loading map editor 600 thus allows the user the option of reloading both fresh and exposed fuel bundles currently residing in one or more fuel pools. One or more of these exposed fuel bundles residing in the available fuel pool inventory may be accessed by the user from fuel pool database 258, for example, and one or more of the fresh fuel bundles may be accessed from fresh fuel bundle design database 252. After accessing, creating and/or editing a fuel loading template using the loading map editor 600 as discussed above, the user may then create a loading map using the template. From the file menu 608, the user chooses a “load” option. The loading map editor 600 then displays a loading screen that includes a template access window, template information window, reload window and a load fresh window. The template access window provides a user with a drop down menu for selecting a loading template stored in the loading templates database 256. The template information window displays summary information for the selected loading template, such as the number of fresh bundle types, the number of reinserted (exposed) fuel bundle positions and the number of locked bundle positions in the loading template. The summary information may also indicate the number of fresh bundle types and number of reinserted (exposed fuel) bundles currently added in creating the loading map. FIG. 8 is a screen shot to illustrate the reload window displayed by the loading map editor 600. The window is divided into a filtered fuel pool table 622 and a reloading pool table 624. The filtered fuel pool table 622 lists (1) the exposed fuel bundles currently in the fuel template of the nuclear reactor being evaluated, except for those fuel bundles in locked fuel bundle positions 612, and (2) the fuel bundle inventory in one or more fuel pools for this and other nuclear reactors. As is well-known, exposed fuel bundles removed from a nuclear reactor are stored in what is known as a fuel pool. Fuel bundles from two or more nuclear reactor cores located at a same site may be stored in the same fuel pool. The filtered fuel pool table 622 lists each exposed fuel bundle by its serial number and bundle name (BName). Each fuel bundle is assigned a unique serial number, used to assure traceability of the bundle from a quality assurance perspective. The bundle name is a character string identifier used to identify the fuel bundle product line as well as nuclear characteristics, such as uranium and gadolinia loading. The filtered fuel pool table 622 also lists one or more attributes of each exposed fuel bundle listed, such as K infinity, exposure, and the last fuel cycle number for which the bundle was resident in the core. Additional attributes for an exposed fuel bundle may include: 1) bundle product line, 2) initial uranium loading, 3) initial gadolinium loading, 4) number of axial zones, 5) historical fuel cycle numbers previous to the most recent for which the bundle was resident in the core, 6) the corresponding reactor in which the fuel bundle was resident for each of the historical fuel cycles, 7) accumulated residence time, and 8) fuel bundle pedigree, a parameter that reflects the usability of the bundle for continued reactor operation. The fuel bundle pedigree is determined from a number of factors the foremost being an inspection of the fuel, either visually or by some other non-destructive test procedure, which is designed to detect a current failed fuel bundle or the vulnerability of the bundle to future failure. Failure mechanisms include such items as corrosion, debris impact, and mechanical bowing of the fuel bundle. Another factor affecting pedigree is possible reconstitution of a fuel bundle, which is a repair process involving the replacement of damaged fuel rods with replacement rods that may be a uranium containing fuel rod or alternatively, a non-uranium containing rod (e.g. stainless steel), henceforth referred to as a ‘phantom’ rod. A pedigree attribute might be ‘RU’ and ‘RP’ for reconstituted with uranium and phantom rods, respectively, and ‘DC’, ‘DD’ and ‘DB’ for damaged by corrosion, debris, and bow, respectively. A ‘blank’ pedigree attribute designates a bundle that was undamaged and useable. The reloading fuel pool table 624 provides the same information for each fuel bundle as provided by the filtered fuel pool table 622. Additionally, the reloading fuel pool table 624 indicates the priority number 626 for each fuel bundle group as set forth in the loading template. FIG. 8 shows that the highest priority (1) reinserted fuel bundle position(s) are a group of four fuel bundles, and the next highest priority (2) reinserted fuel bundle(s) are a group of eight fuel bundles. The reloading fuel pool table 624 is populated by moving fuel bundles from the filtered fuel pool table 622 into the reloading fuel pool table 624. The reload window further includes a set of tools 630 for aiding the user in selecting and moving fuel bundles from the filtered fuel pool table 622 to the reload fuel pool table 624. The set of tools 630 may include a filter tool 632, a move right tool 634, a move left tool 636 and a delete tool 638. Tools 634, 636 and 638 enable to user to move fuel bundles between the filtered fuel pool table 622 and reload fuel pool table 624, such as by highlighting one or more bundles to and then clicking on the move right tool 634 (or left tool 636), which causes the selected fuel bundles to populate the highest priority unpopulated fuel bundle positions in the reload fuel pool table 624 (or filtered fuel pool table 622). The delete tool 638 enables the user to select one or more fuel bundles in one of the tables 622, 624, and then to click the delete tool 638 to delete the selected fuel bundles from the table. FIG. 9 is a screen shot illustrating the filter tool 632. A user clicks on the filter tool 632 to open a filter window 640 as shown in FIG. 9. The filter window lists the same attributes listed in the filtered fuel pool table 622, and allows the user to indicate to filter based on the attribute by clicking in the selection box 642 associated with the attribute. When an attribute has been selected, a check is displayed in the associated selection box 642. The user may also unselect an attribute by again clicking in the associated selection box. In this case, the check mark will be removed. It is of note that exposed bundles stored in available fuel pools may be accessed from the fuel pool database 258 by the user to populate the appropriate tables in FIG. 8 for eventual filtering by the filter tool 632. For each attribute, the filter window 640 may display one or more filter characteristics associated with the attribute. For the K infinity attribute, the user may select a filter operator 644 of greater than, less than, or equal to and enter in a filter amount 646 associated with the filter operator 644. As shown in FIG. 9, a user has selected to filter based on K infinity, chosen the greater than filter operator, and entered the filter amount of 1.2. As a result, the loading map editor 600 will filter the fuel bundles in the filtered fuel pool table 622 to display only those fuel bundles having a K infinity greater than 1.2. As another example, the exposure attribute in box 642 also has an associated filter operator and filter amount. For the cycle attribute, the filter window provides a drop down menu for selecting the cycle number. FIG. 9 shows cycles 2 and 4 selected from the drop down menu for the cycle attribute. As a result, the loading map editor filters the filtered fuel pool table 622 to display only those fuel bundles whose most recent residence was in cycle 2 or cycle 4. Similarly, the user may elect to filter bundles based on their pedigree, product line, etc. Once the attributes for filtering on have been selected and the filter characteristics have been entered, the user causes the loading map editor 600 to filter the filtered fuel pool table 622 based on this information by clicking on the OK selection box. Alternatively, the user may cancel the filter operation by clicking on the CANCEL selection box. The filtered fuel pool table 622 also provides a filtering mechanism for filtering the fuel bundles listed therein. A user may sort the filtered fuel pool table 622 in ascending or descending order of an attribute by clicking on the attribute heading in the filtered fuel pool table 622. Once the user clicks on the attribute, the loading map editor displays a popup menu with the options “Sort Ascending” and “Sort Descending”. The filtered fuel pool table 622 is then filtered in ascending or descending order of the attribute based on the option clicked on by the user. FIG. 10 is a screen shot to illustrate a load fresh window displayed by the loading map editor 600 for loading of fresh bundles into the fuel template. The window is divided into a fresh bundle types table 650 and a fresh bundle pool table 660. The fresh bundle types table 650 lists the available fresh fuel bundle types, listing each fresh fuel bundle type by its bundle name. The bundle name is a character string identifier used to identify the fuel bundle product line as well as nuclear characteristics, such as uranium and gadolinia loading. The fresh fuel bundle types table 650 also lists one or more attributes of each fresh fuel bundle type listed, such as K infinity, fuel bundle product line, average uranium-235 enrichment, percent (as a function of total fuel weight) of gadolinia burnable poison contained in the fuel bundle, number of gadolinia-containing fuel rods, and number of axial zones (an axial zone is defined by a cross-sectional slice of the bundle that is homogeneous along the axial direction). Other attributes of the fresh bundle may include parameters for predicted thermal behavior, such as R-factors and local peaking, calculated for various bundle exposure values. R-factors are used as inputs to the critical power ratio (CPR) and are determined from a weighted axial integration of fuel rod powers. Local peaking is a measure of the fuel rod peak pellet and clad temperature. The fresh bundle pool table 660 provides the same information for each fuel bundle as provided by the fresh bundle types table 650. Additionally, the fresh bundle pool table 660 indicates the type number 662 for each type of fresh bundle in the loading template and then number of fresh fuel bundles of that type in the loading template. FIG. 10 shows that the first type of fresh fuel bundle position(s) are a group of four fuel bundles, and the next type of fresh fuel bundle(s) are a group of eight fuel bundles. Similar to FIG. 8, the fresh bundle pool table 660 may be continually populated and/or updated by moving fuel bundles from the fresh bundle types table 650 into the fresh bundle pool table 660, and includes the tools 630 described in FIG. 8; the filter tool 632 (which may filter out fresh fuel bundles accessed from fresh fuel bundle design database 252), move right tool 634 and delete tool 638 for aiding the user in selecting and moving fuel bundles from the fresh bundle types table 650 to the fresh bundle pool table 660, as described above. The loading map editor 600 also provides for filtering the fresh bundle types table 650 in ascending or descending order of an attribute in the same manner that the filtered fuel pool table 622 may be sorted. Once the reinserted (exposed fuel) and fresh fuel bundle positions 612 are filled using the tools described in detail above, the user may click on a “populate” button displayed in the loading screen to have the loading map displayed. The user may then save the created loading map by using the “Save” or “Save As” options in the file menu 608. At this point, the user has now generated an initial test core design to be evaluated using a plurality of automated tools in order to determine an uprated core design that satisfies all constraints or limits that have been set in an effort to achieve a core design suitable uprated power operations. To load the test core design for evaluation using the automated tools, the user simply clicks a load button and enters ID prefixes for the fresh fuel types; the test core design is ready for evaluation. Block 700-Automated Tools For the initial test core design, the user may modify one or more of the rod pattern, core loading pattern and/or unique fresh fuel bundle types in the test core design until a reactor core design is determined that satisfies all limits and/or constraints for uprated power. As will be explained below, each automated tool may provide associated procedural recommendations to assist the user or designer with any modifications that need to be made to the test core design in order to achieve a desired core design for uprated power. For each automated tool, a simulation of the test core design with one of a selected or modified rod pattern design (for the modify rod pattern tool option), a selected or modified core loading pattern (core loading design tool), and selected unique subset of fresh fuel bundle types (N-Streaming tool) is to be performed. In other words, each tool will establish a respective desired initial test rod pattern design, core loading pattern design, and/or unique fresh fuel bundle type of placement in the test core design output from the functions of block 600. For each of the automated tools, the plant to be evaluated has already been selected and the limits set (at inputs block 500). As the initial test core design has been loaded at the output of block 600, each of the automated tools may be invoked as to be described in more detail below. For each automated tool, based on the corresponding given test rod pattern, core loading pattern or unique fresh fuel bundle types set for the test core design, the simulation results are compared to the limits and data related to the comparison is provided to the user in the form of one or more graphical or text pages; this data may indicate those limits that were violated during simulation and may also be used with procedural recommendations to provide feedback to the user as to what to modify with regard to the rod pattern, the core loading pattern and or the unique fresh fuel bundle types used in the core design. As previously discussed, these tools may be invoked sequentially (i.e., the modify core pattern tool may be iterated with acceptable rod pattern and fresh fuel bundle designs input thereto from the rod pattern and unique fresh fuel type tools, and vice versa), or the tools in another example may be iterated independently from one another, and the best design (as indicative by the objective function values) may serve as a derivative core design resulting there from may be evaluated by the other two tools, sequentially or again independently, to refine the design for uprate power operations. The Modify Rod Pattern Tool The initial test core design generated above in block 600 may not have the desired rod pattern design (e.g., notch positions and sequences for control blade patterns in BWRs, or group sequences for control rod patterns in PWRs, etc.) for uprated power operations. This may be determined by evaluating the results of simulation of the given test reactor core design that has been loaded with a test rod pattern design (e.g., notch positions and sequences for control blade patterns for BWRs, group sequences for control rod patterns for PWRs, etc.) and then simulated using core simulation software to generate simulation results for comparison against the limits (inclusive of those limits for uprated power operations) set in the inputs block 500. Accordingly, a simulation of the given test reactor core design from block 600, but which includes an initial test rod pattern design, will need to be performed. The implementation of the rod pattern tool and routine for modifying rod patterns in the test core design is described in detail in co-pending and commonly assigned U.S. patent application Ser. No. 10/321,441 to Russell, II et al. In general, a user implements the modify rod pattern tool via the GUI 230 and through processing power (the local computer at user, the application server 200 and/or calculation servers 400, etc.) to virtually modify rod pattern designs (e.g., notch positions and sequences for control blade patterns for BWRs, group sequences for control rod patterns for PWRs, etc.) for review on a suitable display device by the user. The iteration of the modify rod pattern subroutine may also provide feedback and/or procedural recommendations to the user, based on how closely a proposed rod pattern design solution meets the user input limits or constraints that need to be satisfied for the test reactor core design to meet the requirements for uprated power operations. It is of note that many of the steps and figures describing the modify rod pattern tool are similar for the core loading pattern and unique fresh fuel bundle type tools. Accordingly, several of the figures describing the modify rod pattern tool will be referenced in the discussion of the core loading pattern and unique fresh fuel bundle type tools, with only the differences noted for purposes of brevity. FIG. 11 is a flow chart illustrating implementation of the modify rod pattern tool in accordance with an exemplary embodiment of the invention. FIG. 11 is described in terms of determining a rod pattern design for an exemplary boiling water reactor, it being understood that the method may be applicable to PWRs, gas-cooled reactors and heavy-water reactors. Referring to FIG. 11, a reactor plant is selected for evaluation in Step S5 and limits which are used in a simulation for a test rod pattern design of the selected plant are defined (Step S10). These steps have previously been performed by the functions in input block 500, but are reiterated herein for clarity. Based on the limits, a sequence strategy (initial rod pattern design) for control mechanism movement (e.g., control blade notch positions, control rod positions, etc.) is established (Step S20). Reactor operation may be simulated (Step S30) on the entire test core design with the selected rod pattern design, or focused on a subset of the test rod pattern design, which may be a subset of fuel bundles in a reactor core for example, in order to produce a plurality of simulated results. The simulated results are compared to the limits (Step S40), and based on the comparison, data is provided illustrating whether any limits have been violated (Step S50). The data also provides the user with an indication of which location in a simulated core were the largest violators or largest contributors to a limit violation. Each of these steps is now described in further detail below. Once the plant is selected, an initial rod pattern design may be selected by entering a previous test rod pattern (sequence strategy) using GUI 230 to access a plant configuration webpage. For example, the webpage may enable to user to select a “model size”, e.g., quarter core, half core, or full core, for evaluation in a subsequent simulation. Additionally, a user may select a core symmetry option (e.g., octant, quadrant, no symmetry) for the selected model size, by clicking on a suitable drop down menu and the like. In an example, this core symmetry option has already been performed as evident in FIG. 6 for example. As previously shown in FIG. 5, the client-inputted plant specific constraints, which may be configured as limits on input variables to the simulation and limits on the simulation results have already been set for the simulation. A sequence strategy (test rod pattern design for the test reactor core design output from block 600) for positioning one or more subsets of a test rod pattern design is established (Step S20) based on the limits. In an embodiment where the reactor being evaluated is a boiling water reactor, for example, the limits help to establish allowable control blade positions and durations. Control blade themes are defined together with allowable blade symmetry to aid the user in determining an initial sequence strategy. In typical BWR operation, for example, the control blades may be divided into four groups of blades (“A1”, “A2”, “B1”, and “B2”). By apportioning blades into these blade groups, the user may easily identify only the permissible blades within a given sequence. Consequently, undesirable blades are prevented from being used, which would result in an undesirable solution. Because control blade themes are identified for each exposure, satisfactory blade definitions may be forced. By defining control blade themes and blade symmetry, the user need only identify a single blade position within the control blade theme, and the other symmetric control blades will accordingly follow. Thus, the graphical area is not cluttered by redundant control blade location information. Further, automating a sequence strategy prevents illegal control blade position errors from occurring. The user proceeds to identifying all sequences and the initial rod pattern determination from a beginning of cycle (BOC) to end of cycle (EOC). FIG. 12 is a screen shot illustrating how a control blade sequence may be entered. The column entitled blade group at 1217 enables the user to adjust or set the sequence strategy based on what user constraints have already been entered, for example. In FIG. 12, the user has set the exposure steps at 1211, calculation type at 1213, detailed rod pattern at 1215, blade groups at 1217 and any appropriate operation parameters. With the limits having been defined and the sequence strategy (initial test rod pattern for the test reactor core design) having been established, a simulation is initiated (Step S30). The simulation may be executed by calculation servers 400; however, the simulation may be a 3D simulation process that is run external to the arrangement 1000. The user may employ well-known executable 3D simulator programs such as PANACEA, LOGOS, SIMULATE, POLCA, or any other known simulator software where the appropriate simulator drivers have been defined and coded, as is known. The calculation servers 400 may execute these simulator programs based on input by the user via GUI 230. The user may initiate a 3D simulation at any time using GUI 230, and may have a number and different means to initiate a simulation. For example, the user may select a “run simulation” from a window drop down menu, or could click on a “RUN” icon on a webpage task bar, as is known. Additionally, the user may receive graphical updates or status of the simulation. Data related to the simulation may be queued in queue database 253 within relational database server 250. Once the simulation is queued, the user may have an audio and/or visual indication as to when the simulation is complete, as is known. Once the user initiates simulation, many automation steps follow. FIG. 13 is a flow chart illustrating simulation Step S30 in further detail. Initially, all definitions for the test core design with the selected rod pattern design problem are converted into a 3D instruction set (e.g., a computer job) for the 3D reactor core simulator (Step S31). This enables the user to have a choice of several types of simulators, such those described above. Selection of a particular simulator may be dependant on the plant criteria entered by the user (e.g. the limits). The computer job is readied for queuing in the queue database 253 of each relational database server 250 (Step S33). The storing of the data for a particular simulation enables any potential simulation iteration to start from the last or previous iteration. By storing and retrieving this data, future simulation iterations to a rod pattern design take only minutes or seconds to perform. Concurrently, a program running on each of the available calculation servers 400 scans every few seconds to look for available jobs to run (Step S37). If a job is ready to run, one or more of the calculation servers 400 obtains the data from the queue database 253 and runs the appropriate 3D simulator. As described above, one or more status messages may be displayed to the user. Upon completion of the simulation, all results of interest may be stored in one or more subordinate databases within the relational database server 250 (e.g., simulation results database 255). Accordingly, the relational database server 250 may be accessed in order to calculate the objective function values for the test rod pattern design. FIG. 14 is a flow diagram illustrating the comparing step of FIG. 11 in further detail. The objective function may be stored in relational database server 250 for access by calculation servers 400. Objective function calculations, which provide objective functions values, may also be stored in the relational database server 250, such as in a subordinate objective function value database 257. Referring to FIG. 14, inputs to the objective function calculation include the limits from the limits database 251 and the simulator results from the simulator results database 255. Accordingly, one or more calculation servers 400 access this data from relational database server 250 (Step S41). Although the example embodiments of the present invention envision any number of objection function formats that could be utilized, one example includes an objective function having three components: (a) the limit for a particular constraint parameter (e.g., design constraint for reactor plant parameter), represented as “CONS”; the simulation result from the 3D simulator for that particular constraint parameter, represented as “RESULT”, and a multiplier for the constraint parameter, represented by “MULT”. A set of predefined MULTs may be empirically determined from a large collection of BWR plant configurations, for example. These multipliers may be set at values that enable reactor energy, reactivity limits, and thermal limits to be determined in an appropriate order. Accordingly, the method of the present invention utilizes a generic set of empirically-determined multipliers, which may be applied to over thirty different core designs. However, GUI 230 permits manual changing of the multipliers, which is significant in that user preference may desire certain constraints to be “penalized” with greater multipliers than the multipliers identified by the pre-set defaults. An objective function value may be calculated for each individual constraint parameter, and for all constraint parameters as a whole, where all constraint parameters represent the entity of what is being evaluated in a particular test rod pattern of the test reactor core design (or as will be discussed below, a particular core loading pattern or one or more (N) unique fresh fuel bundle types for the test core design). An individual constraint component of the objective function may be calculated as described in Equation (1):OBJpar=MULTpar*(RESULTpar−CONSpar); (1)where “par” may be any of the client-inputted constraints listed in FIG. 5. These parameters are not the only parameters that could be possible candidates for evaluation, but are parameters which are commonly used in order to determine a suitable core configuration for a nuclear reactor. The total objective function may be a summation of all constraint parameters, orOBJTOT=SUM(par=1,31){OBJpar} (2) Referring to Equation 1, if RESULT is less than CONS (e.g. there is no violation of a constraint), the difference is reset to zero and the objective function will be zero. Accordingly, objective function values of zero indicate that a particular constraint has not been violated. Positive values of the objective function represent violations that may require correction. Additionally, the simulation results may be provided in the form of special coordinates (i, j, k) and time coordinates (exposure step) (e.g., particular time in a core-energy cycle). Therefore, the user can see at which time coordinate (e.g., exposure step) the problem is located. Hence, the rod pattern (or core loading pattern or unique fresh fuel bundle type pattern) is modified only at the identified exposure step. In addition, objective function values may be calculated as a function of each exposure step, and totaled for the entire design problem (Step S43). The objective function values calculated for each constraint, and the objective function values per exposure step, may be further examined by normalizing each objective function value to provide a percentage contribution of a given constraint to a total objective function value (Step S45). Each result or value of an objective function calculation is stored (Step S47) in a subordinate objective function value database 257 within relational database server 250. The objective function values may be utilized in the manual determination of the desired rod pattern for the test core design. For example, the values of the objective function calculations may be viewed graphically by the user in order to determine parameters that violate limits. Additionally, any change in objective function values over successful iterations of rod pattern design changes in the test core design provides the user with a gauge to estimate both improvement and detriment in their proposed design. Increases in an objective function value over several iterations indicate that the user's changes are creating a rod pattern design (and hence power uprate core design) that is moving away from a desired solution, while successive iterations of lesser objective functions values (e.g., the objective function value decreasing from a positive value towards zero) may indicate improvements in the iterative design. The objective function values, limits and simulation results over successive iterations may be stored in various subordinate databases within relational database server 250. Therefore, designs from past iterations may be quickly retrieved, should later modifications prove unhelpful. FIG. 15 is a screen shot to illustrate exemplary graphical data available to the user for review after completion of the objective function calculations. Upon completion of the objective function calculation, the user may be provided with data related to the objective function calculations, which may include limits that have been violated during the simulation of the test reactor core design configured with the evaluated test rod pattern. Referring to FIG. 15, there is displayed a list of constraint parameters which may represent the input limits, and the values of each of objective function value calculation on a per constraint basis. FIG. 15 illustrates limits which have been violated with a check in a box, as indicated by checked box 1505 for example. Additionally, for each limit violation, its contribution and percent (%) contribution (based on the calculations and the normalization routines described with respect to FIG. 14), are displayed. Accordingly, based on this data, the user may be provided with a recommendation as to what modifications need to be made to the rod pattern design for a subsequent iteration. Although the individual rod pattern modifications may alternatively be left to the desires of the user, procedural recommendations may be provided in the form of a pull down menu, for example. These recommendations may be divided into four categories: energy beneficial moves, reactivity control, energy detrimental moves, and converting excessive margin (from thermal limit) into additional energy. As discussed above, one technique is to address problems using energy beneficial moves rather than energy detrimental moves. Even if the rod pattern design meets all of the limits (client-inputted plant specific constraints, design limits, thermal limits, etc.) the user may verify that any excessive margin to a particular limit is converted into additional energy. Accordingly, the following logic statements may represent the above procedural recommendations modifying the rod pattern: Energy Beneficial Moves If peaking off top of blade, insert blade deeper If NEXRAT (e.g., a thermal margin constraint) problem at EOC, insert blade deeper earlier in cycle If kW/ft peaking during mid cycle, drive deeper rods deeper earlier in cycleReactivity Control If flow too high during sequence, pull deep rods If flow too low during sequence, drive rods deeperEnergy Detrimental Moves If peaking at bottom of core during sequence, insert shallow blade in local areaConverting Excessive Margin Into Additional Energy If extra MFLCPR margin at EOC, drive blades deeper earlier in cycle If extra kW/ft margin EOC, drive blades deeper earlier in cycle If extra MFLCPR margin in center of core at EOC, drive center rods deeper earlier in cycleBased on the location, and on the time exposure of limit violations, as indicated by the objective function, a user may easily follow one or more of the above recommendations to address and fix constraint violations. The data resulting from the objective function calculations may be interpreted on a suitable display device. For example, this data may be displayed as a list of constraints with denoted violators, as described with respect to FIG. 15. However, the user may access a number of different “result” display screens that may configurable as 2- or 3-dimensional views, for example. The following Table 1 lists some of the exemplary views available to the user. TABLE 1GRAPHICAL VIEWS AVAILABLE TO USERObjective function results - listingGraph of max core value vs. exposureGraph of nodal maximum value vs. exposureGraph of location of max core value vs. exposureGraph of pin value vs. exposureGraph of bundle maximum value vs. exposureView 3D rotational diagramReport performance relative to previous iterationReport improvement rates of various designersDisplay of server statusDisplay of queue statusDisplay system recommendations FIGS. 16, 17A and 17B are screen shots to illustrate graphical data available to the user to review the status of given constraints or limits subsequent to simulation and the objective function analysis, in accordance with the invention. Referring to FIG. 16, a user may pull down a suitable drop down menu from a “view” icon on a task bar in order to display views of certain constraints or parameters. As illustrated in FIG. 16, a user has selected a Maximum Fractional Limiting Power Density (MFLPD) constraint parameter. There are a number of different graphical views available to the user, as indicated by pull-down menu 1610. The user simply selects the desired view and may then access a page such as is illustrated in FIG. 17A or 17B. FIG. 17A illustrates two different 2-dimensional graphs of particular constraints, as seen at 1705 and 1710. For example, the user can determine where violations of Maximum Average Planar Heat Generation Rate (MAPLHGR) occur (in a core maximum vs. exposure graph 1705, and an axial values of MFLPD vs. exposure graph 1710) for a particular exposure in a core cycle. The limits for these constraints are shown by lines 1720 and 1725, with violations shown generally at 1730 and 1735 in FIG. 17A. FIG. 17B illustrates another view, in this case a two dimensional view of an entire cross section of a core, in order to see where the biggest violation contributors for MAPLHGR vs. exposure are located. As can be seen at 1740 and 1750, the encircled squares represent the fuel bundles that are the largest violation contributors to MAPLHGR in the core (e.g., 1740 and 1750 pointing to bundles violating MAPLHGR). This gives the user an indication of which fuel bundles in the rod pattern design may need modification. FIGS. 18A and 18B are flow diagrams describing rod pattern modification and iteration processing steps in accordance with the example modify rod pattern tool, in accordance with the invention. Referring to FIG. 18A, by interpreting the data at Step S60 (such as the data shown and described in FIGS. 15 through 17B, for example), the user may be inclined to initiate a modifying subroutine (Step S70). In all practicality, the test reactor core design with the initial test rod pattern will not be an acceptable design, and the modifying subroutine likely will be required. In one embodiment, the user can manually direct this modifying subroutine, with the help of the graphical user GUI 230. In another embodiment, the subroutine may be performed within the bounds of an optimization algorithm that automatically iterates simulation, calculation of objective function and evaluation of the results or values of the objective function calculations for a number of rod pattern design iterations. The user determines, based on the displayed data, whether any limits are violated (Step S71). If no limits are violated, the user determines if any identifiers indicate that characteristics of maximum power (for uprated power operations) are obtained from the test reactor core design with selected test rod pattern. For example, these identifiers may include an indication of good thermal margin utilization (such as margins on MFLCPR and LHGR) by driving rods deeper into the core to maximize plutonium generation for cycle extension. Power requirements may be shown to be met when the minimum EOC eigenvalue is obtained for the cycle design (eigenvalue search) or the desired cycle length is determined at a fixed EOC eigenvalue. If there is an indication that max power has been obtained from the test reactor core design with selected test rod pattern (the output of Step S72 is YES), an acceptable test rod pattern for the test reactor core design has been determined, and the user may access a report of the results related to the rod pattern design (Step S73) and/or use the selected rod pattern in the modify core and/or unique fresh fuel bundle type design tools (Step S78). If limits are violated (the output of Step S71 is YES) or limits are not violated but there is an indication that max power has not been obtained from the rod pattern design (the output Step S72 is NO) then the user determines whether any characteristics indicate that modification of the sequence strategy is required (Step S74). Characteristics that indicate a need to modify the sequence strategy may include excessive control blade (control rod) history, excessive MFLCPR at EOC in local areas and inability to contain MFLCPR at individual exposures. Additionally, if several iterations of rod pattern design changes have been attempted and there has been no real improvement to the objective function, this is a further indication that an alternative rod pattern sequence might need to be explored. FIG. 19 is a screenshot for defining blade groups in order to modify the rod pattern. Accordingly, if the sequence strategy requires modification (the output of Step S74 is YES), the user creates a derivative rod pattern design by changing the sequence strategy (Step S75). For example, and referring to FIGS. 12 and 19, the user may select an edit option on the operations configuration page to change the blade groupings (see 1217 in FIG. 12). If there are no characteristics indicating that the sequence strategy needs to be modified (the output of Step S74 is NO) the user may modify the test rod pattern design to create a derivative rod pattern (and hence a derivative reactor core design) by changing positions of control blades or control rods. Referring to FIG. 12, the user checks a “set rods” box 1230 for a particular exposure and selects edit icon 1235. As shown in FIG. 19, these operations may bring up another display that enables the user to manually alter the notch positions of the control blades in a particular group. In FIG. 19, there is shown a “Define Blade Groups” screenshot 1940, which illustrates a core cross section with a blade group Interior A1 selected at cell 1941. By selecting options pull down menu 1942, the user may display another window called a “Set Blade Constraints” window 1945. The minimum withdrawal column 1950 identifies how far a blade is allowed into the core. The maximum withdrawal column 1955 identifies how far the blade is allowed out of the core, and the Not Allowed column 1960 identifies blade locations that are not allowed for this particular rod pattern design. The example are not limited to changing control blade notch positions for boiling water reactors, but also to changing rod position of control rods in pressurized water reactors, as well as control rod positions in other types of reactors (e.g., gas cooled reactor, heavy water reactors, etc.). Regardless of whether the test rod pattern was modified by changing rod positions or modified by changing sequence strategy, Steps S30-S50 are repeated to determine if the derivative rod pattern (and hence the derivative reactor core design) meets all limits (Step S77). This may become an iterative process. FIG. 18B illustrates the iterative process in accordance with an example embodiment of the invention. For each derivative rod pattern design that has been simulated, the user determines whether any data that is related to the comparison between simulated results and limits (e.g., the calculated objective function values) still indicates that there are limit violations. If not, the user has developed an acceptable rod pattern design that may be used in a particular reactor, and may access graphical results related to the acceptable rod pattern design (Step S73) and/or use the selected rod pattern in the modify core and/or unique fresh fuel bundle type design tools (Step S78). If an iteration still indicates that limits are violated (the output of Step S160 is YES) then the modifying subroutine in Step S70 is iteratively repeated until all limits are satisfied, or until all limits are satisfied within a margin that is acceptable, as determined by the user (Step S170). The iterative process is beneficial in that it enables the user to fine tune a rod pattern design, and to perhaps extract even more energy out of an acceptable rod pattern design than was previously possible of doing with the conventional, manual iterative process. Further, incorporation of the relational database server 250 and a number of calculation servers 400 expedite calculations. The iterative process as described in FIG. 18B may be done in an extremely short period of time, as compared to a number of weeks using the prior art manual iterative process of changing one parameter at a time, and then running a reactor core simulation. To this point, the modify rod pattern tool been described in terms of a user or designer interpreting data via GUI 230 and modifying the test rod pattern (and hence test reactor core design) iteratively, by hand, based on displayed feedback (the data from the objective finction) in order to get a desired design. However, the aforementioned steps of FIGS. 18A and 18B may also be effectuated by way of an optimization process. Such an optimization process is equally applicable to the modify core loading pattern tool and unique fresh fuel bundle type design tool to be explained hereafter. The optimization process iterates the steps in FIGS. 18A and 18B over a number of different rod pattern designs, constantly improving on violated limits in order to achieve an optimal rod pattern design to be used in a nuclear reactor core. FIG. 20 illustrates a screen shot to initiate such a process. For example, after selecting the plant and the test rod pattern, the user may display an optimization configuration screen 2005. The user may select optimization parameters 2040 of optimize rod patterns, optimize core flow, and optimize sequence intervals, for example. Optimize rod patterns means making an optimal determination of individual rod (control blades in BWRs) positions within a control rod grouping (called a sequence), for the duration of time during the operating cycle when the given sequence is being used to control the reactor. Rod positions affect the local power as well as the nuclear reaction rate. Optimize core flow means making an optimal determination of reactor coolant flow rate through the reactor as a function of time during the operating cycle. Flow rate affects global reactor power as well as the nuclear reaction rate. Optimize sequence intervals means making an optimal determination of the time duration a given sequence (i.e., control rod grouping) is used to control the reactor during the operating cycle. Sequence intervals affect local power as well as the nuclear reaction rate. Using a suitable input device (e.g., keyboard, mouse, touch display, etc.), the user may select, via GUI 230, one or more of the optimization parameters by clicking in the selection box 2042 associated with a given optimization parameter 2040. When selected, a check appears in the selection box 2042 of the selected optimization parameter. Clicking in the selection box 2042 again de-selects the optimization parameter. As shown in FIG. 20, to optimize the iterative modify rod pattern design process, the user would select boxes 2042 for optimize rod patterns, optimize core flow and optimize sequence intervals. Memory (relational database server) 250 may also store constraint parameters associated with the optimization problem. These may be stored in limits database 251 for example. The constraint parameters are parameters of the optimization problem that must or should satisfy a constraint or constraints, where a constraint may be analogous to the limits described above. FIG. 21 illustrates a screen shot of an exemplary optimization constraints page listing optimization constraints associated with an optimization problem of boiler water reactor core design. As shown, each optimization constraint 2150 has a design value 2152 associated therewith. Each optimization constraint must fall below the specified design value. The user has the ability to select optimization parameters for consideration in configuring the objective function. The user selects an optimization constraint by clicking in the selection box 2154 associated with an optimization constraint 2150. When selected, a check appears in the selection box 2154 of the selected optimization constraint 2150. Clicking in the selection box 2154 again de-selects the optimization constraint. Each optimization parameter may have a predetermined credit term and credit weight associated therewith stored in relational database server 250. Similarly, each optimization constraint has a predetermined penalty term and penalty weight associated therewith, which may be stored in relational database server 250, such as in limits database 251 and/or objective function values database 257. As seen in FIG. 21, the penalty term incorporates the design value, and the user can change (i.e., configure) this value as desired. Additionally, the embodiment of FIG. 21 allows the user to set an importance 2156 for each optimization constraint 2150. In the importance field 2158 for an optimization constraint, the user may have pull down options of minute, low, nominal, high and extreme. Each option correlates to an empirically predetermined penalty weight such that the greater the importance, the greater the predetermined penalty weight. In this manner, the user selects from among a set of predetermined penalty weights. Once the above selections have been completed, a calculation server 400 retrieves the selections above from relational database server 250 and configures the objective function according to the generic definition discussed above and the selections made during the selection process. The resulting configured objective function equals the sum of credit components associated with the selected optimization parameters plus the sum of penalty components associated with the selected optimization constraints. Additionally, this embodiment provides for the user to select a method of handling the credit and penalty weights. For example, the user is supplied with the possible methodologies of static, death penalty, dynamic, and adaptive for the penalty weights; is supplied with the possible methodologies of static, dynamic and adaptive for the credit weights; and the methodology of relative adaptive for both the penalty and credit weights. The well-known static methodology maintains the weights at their initially set values. The well-known death methodology sets each penalty weight to infinity. The well-known dynamic methodology adjusts the initial weight value during the course of the objective function's use in an optimization search based on a mathematical expression that determines the amount and/or frequency of the weight change. The well-known adaptive methodology is also applied during the course of an optimization search. In this method, penalty weight values are adjusted periodically for each constraint parameter that violates the design value. The relative adaptive methodology is disclosed in U.S. patent application Ser. No. 10/246,718, entitled METHOD AND APPARATUS FOR ADAPTIVELY DETERMINING WEIGHT FACTORS WITHIN THE CONTEXT OF AN OBJECTIVE FUNCTION, by the inventors of the subject application, filed on Sep. 19, 2002. Optimization Using the Objective Function FIG. 22 illustrates a flow chart of an optimization process employing the objective function in accordance with an exemplary embodiment of the present invention. This optimization process is disclosed in co-pending and commonly assigned U.S. patent application Ser. No. 10/246,716, entitled METHOD AND APPARATUS FOR EVALUATING A PROPOSED SOLUTION TO A CONSTRAINT PROBLEM, filed on Sep. 19, 2002. The relevant portions of the '716 describing the optimization process are hereby incorporated in its entirety by reference herein. For the purposes of explanation only, the optimization process in FIG. 22 is described as being implemented by the architecture illustrated in FIG. 1. As shown, in Step S2210 the objective function is configured as discussed above in the preceding section, then the optimization process begins. In Step S2212, the calculation processors 400 retrieve system inputs from relational database 250, or generate one or more sets of values for input parameters (i.e., system inputs) of the optimization problem based on the optimization algorithm in use. For example, these input parameters may be related to determining rod patterns bundles in the test reactor core so as to meet the constraint or limit criteria for uprated power operations, but may also be related to the other automated tools, i.e., for determining an core loading pattern and/or a unique fresh fuel bundle types for placement in the test reactor core so as to meet the constraint or limit criteria for uprated power operations. However, optimization is not limited to using these parameters, as other input parameters in addition to placement of fresh and exposed fuel bundles within the reactor, selection of the rod groups (sequences) and/or placement of the control rod positions within the groups as a function of time during the cycle can be used, such as parameters directed to core flow as a function of time during a cycle, reactor coolant inlet pressure, etc. Each input parameter set of values is a candidate solution of the optimization problem. The core simulator as described above runs a simulated operation and generates a simulation result for each input parameter set of values. The simulation result includes values (i.e., system outputs) for the optimization parameters and optimization constraints. These values, or a subset of these values, are values of the variables in the mathematical expressions of the objective function. Then, in step S2214, a calculation processor 400 uses the objective function and the system outputs to generate an objective function value for each candidate solution. In step S2216, the calculation processor 400 assesses whether the optimization process has converged upon a solution using the objective function values generated in step S2214. If no convergence is reached, then in step S2218, the input parameter sets are modified, the optimization iteration count is increased and processing returns to step S2212. The generation, convergence assessment and modification operations of steps S2212, S2216 and S2218 are performed according to any well-known optimization algorithm such as Genetic Algorithms, Simulated Annealing, and Tabu Search. When the optimization is utilized to determine an acceptable rod pattern design (and/or an acceptable core loading pattern design or unique fresh fuel bundle types for placement in the core), the optimization is run until convergence (e.g., acceptable results as in steps S73/S173 of FIGS. 18A and 18B) is obtained. The Modify Core Loading Pattern Tool The modify core loading pattern tool is described in detail in co-pending and commonly-assigned U.S. application Ser. No. 10/401,602 to William Earl Russell, II et al., filed Mar. 31, 2003 and entitled “Method and Arrangement for Developing Core Loading Patterns in Nuclear Reactors. The user may implement to core loading pattern tool in conjunction with an accepted rod pattern design determined for the test reactor core, or the tool may be iterated on the existing test reactor core design determined in block 600. As described with the rod pattern tool, the test reactor core design (loaded in accordance with the limits identified in FIG. 5 and supporting disclosure, with the desired inventory of fresh fuel bundles from FIG. 6, with a desired initial test core loading pattern determined as described in FIGS. 8 and 9 and fresh fuel loading pattern as described in FIG. 10, the test reactor design is simulated. As per the simulation, the only difference in FIG. 13 is that in step S31 the desired core loading pattern, instead of (or in addition to) the rod pattern (control blade sequence) for the test core design is converted into a 3D instruction set. Thus, a desired control blade sequence (rod pattern) determined from iteration of the modify rod pattern tool may be set for the desired core loading pattern (and incorporated as well into the 3D instruction set) for the test core design to be simulated. Similarly as described in FIG. 14, inputs to the objective function calculations from the limits database 251 and the simulator results from the simulator results database 255 are accessed by one or more calculation servers 400 (Step S41), and the objective function values are calculated as a function of each exposure step, and totaled for the entire design problem (Step S43). The calculated objective function values calculated for each constraint, and the objective function values per exposure step are normalized to provide a percentage contribution of a given constraint to a total objective function value (Step S45). Each result or value of an objective function calculation is stored (Step S47) in a subordinate objective function value database 251 within relational database server 250. As shown in FIG. 15, the user may be provided with data indicating those limits that may have been violated during the simulation. Accordingly, based on this data, the user may be provided with recommendation(s) as to what modifications may need to be made to the selected core loading pattern of the test core design for a subsequent iteration. As discussed above with the modify rod pattern tool, while individual core loading pattern modifications may alternatively be left to the desires of the user, procedural recommendations as energy beneficial moves, energy detrimental moves, and converting excessive margin (from thermal to limit) into additional energy may be accessible by the user. Even if the test core design with the desired core loading pattern meets all of the limits (client-inputted plant specific constraints, design limits, thermal limits, etc.) the user may verify that any excessive margin to a particular limit is converted into additional energy. Accordingly, the following logic statements may illustrate the above procedural recommendations for modifying the core loading pattern: Energy Beneficial Moves If Critical Power Ratio (CPR) margin too low towards core perimeter, move more reactive (less exposed) fuel toward core center If MFLPD (e.g., a thermal margin constraint) problem at EOC, move more reactive fuel towards problem location If shutdown margin (SDM) problem at core perimeter at BOC, place less reactive fuel toward core perimeterEnergy Detrimental Moves If Minimum Critical Power Ratio (MCPR) margin too low at EOC, move less reactive (more exposed) fuel into problem location(s) If KW/ft margin (MAPLHGR) too low at EOC, move less reactive fuel into problem location(s)Converting Excessive Margin Into Additional Energy If extra MCPR margin in center of core at EOC, move more reactive fuel from core perimeter location to core center The data resulting from the objective function calculations may be displayed as a list of constraints with denoted violators, as shown in FIG. 15 and Table 1. The user may again be presented with graphical displays as shown in FIGS. 16 to 17B to see the effect of the core design on certain constraints or parameters. For example, in FIG. 17B at 1740 and 1750, the encircled squares represent the fuel bundles that are the largest violation contributors to MAPLHGR in the core (e.g., 1740 and 1750 pointing to bundles violating MAPLHGR). This gives the user an indication of fresh and/or exposed fuel locations in the test core design that may need modification. FIGS. 23A and 23B are flow diagrams describing core loading pattern modification and iteration processing steps in accordance with the example modify core loading pattern tool, in accordance with the invention. The functions described in these figures are similar to FIGS. 18A and 18B, and thus similar notation has been used for clarity with the differences discussed in more detail below. As with the modify rod pattern tool, the user may direct each iteration of this modifying subroutine, with the help of the graphical user GUI 230, or the modifying subroutine may be performed within the bounds of an optimization algorithm that automatically iterates simulation, calculation of objective function and evaluation of the results or values of the objective function calculations for a number of core loading pattern design iterations. The user determines whether any limits are violated (Step S71), and if none, determines if any identifiers indicate that characteristics of maximum power are obtained from the selected core loading pattern for the test core design. If there is an indication that maximum power has been obtained from the test core design with the selected core loading pattern (the output of Step S72 is YES), an acceptable core loading pattern design has been determined, and the user may access a report of results and data related to the accepted design (Step S73) and/or use the selected core loading pattern in the modify unique fresh fuel bundle type design tool and/or as an input back to the modify rod pattern tool (Step S78). If limits are violated (the output of Step S71 is YES) or limits are not violated but there is an indication that maximum power has not been obtained from the core loading pattern design (the output Step S72 is NO) the user determines whether any characteristics indicate that modification of fresh fuel bundle numbers is required (Step S74). Characteristics that indicate a need to modify the fresh fuel bundle number may include an energy shortfall, a margin shortfall with acceptable energy, and/or a loss of reactivity due to outage date changes. Additionally, if several iterations of core loading pattern design changes have been attempted and there has been no real improvement to the objective function, this is a further indication that an alternative core loading pattern design might need to be explored. Accordingly, if the output of Step S74 is YES, the user may create a modified or derivative core loading pattern design by re-estimating the number of fresh fuel bundles needed, rounding bundle numbers down as required for core symmetry and loading the core according to the revised or derivative test core loading pattern (Step S75). Step S75 generally corresponds to steps 604 through 607 in FIG. 6. If there are no characteristics indicating a need to modify the fresh fuel bundle number (the output of Step S74 is NO) the user may modify the test core design (Step S76) to create a derivative core design. In making a modification to the test core loading pattern based on the procedural recommendations described above, the user may alter the core loading pattern via the GUI 230. For example, and using a suitable input device (mouse, keyboard, touch screen, voice command, etc) and GUI 230, a designer may identify the core symmetry option for any exposed (or fresh) fuel bundle(s) in the core design that the user desires to move, may select these “target” exposed fuel bundle(s), and may select the “destination” exposed (or fresh) fuel bundles in the core design for replacement by the target exposed (or fresh) fuel bundle(s). The target and destination bundles may be “shuffled” according to the required symmetry (mirror, rotational, etc.). This process may be repeated for any exposed (or fresh) fuel bundle shuffle that is required to re-load modified or derivative test core design in the desired manner. FIG. 24 is a screen shot illustrating the modifying Step S76 of FIG. 23A in further detail in accordance with an exemplary embodiment of the invention. FIG. 24 illustrates the functionality available to the user so as make swift design modifications to the core loading pattern of the test core design. A user may select a fuel shuffling page 2405 and may select a “bundle shuffle” taskbar 2410 in order to display a screen 2415 of a portion of a core design. In FIG. 13, an exposed fuel bundle designated at 2420 is being changed from one fuel bundle type (IAT type 11) to another (IAT type 12). In an example, an exposed fuel bundle may be swapped with a fresh fuel bundle by selecting a fresh fuel bundle in the core design, the exposed fuel bundle, and selecting the “SWAP” button 2430. The portion of the core shown in screen 2425 may be color coded to show the various exposures (GWD/st) of each of the exposed and fresh fuel bundles. A corresponding color coded key may be displayed as indicated at 2427 for example. Selection of items in FIG. 24 may be possible via a suitable input device, such as a mouse, keyboard, touch screen, voice-activated command, etc. These core loading pattern design modifications may be saved in relational database 250, such as in 3D Simulator input parameters database 259, for example. Referring again to FIG. 23A, regardless of whether the test core loading pattern was modified as described in Steps S75 or S76, Steps S30-S50 may be repeated to determine if the derivative core design which meets all limits (Step S77). This may become an iterative process. FIG. 23B illustrates an iterative process in accordance with an exemplary embodiment of the invention. For each derivative core loading pattern design from Step S70 that has been simulated, the user determines whether any data that is related to the comparison between simulated results and limits (e.g., the calculated objective function values) still indicates that there are limit violations (Step S160). If not, (output of Step S160 is NO) the user has developed an acceptable derivative core loading pattern design that may be used in a particular reactor, and may access graphical results related to the acceptable core loading pattern design (Step S73) and/or use the selected derivative core loading pattern in the modify unique fresh fuel bundle type design tool and/or as an input back to the modify rod pattern tool (Step S78). If an iteration still indicates that limits are violated (the output of Step S160 is YES) then the modifying subroutine in Step S70 may be iteratively repeated until all limits are satisfied/maximum power obtained, or until all limits are satisfied/maximum power obtained within a margin thereto that is acceptable, as determined by the user (Step S170). The iterative process may be beneficial in that it enables the user to fine tune a core loading pattern design, and to perhaps extract even more energy out of an acceptable core loading pattern design than was previously possible of doing with the conventional, manual iterative process. Further, incorporation of the relational database server 250 and a number of calculation servers 400 expedite calculations. The iterative process as described in FIG. 23B may be done in an extremely short period of time, as compared to a number of weeks using the prior art manual iterative process of changing one parameter at a time, and then running a reactor core simulation. As described with the modify rod pattern tool in FIGS. 18A and 18B, the functions of FIGS. 23A and 23B may also be iterated over N different core loading pattern designs, in an effort to consistently improve toward a desired core loading pattern design that satisfies all user limits and constraints, for use in a nuclear reactor core. Referring again to FIG. 20, for optimizing the iterative core loading pattern process, the user would check box 2042 for the optimize fuel loading parameter. Optimize fuel loading selection means making an optimal determination of the once and twice burnt fuel. As shown in FIG. 21, the user may select desired optimization constraints (box 2152) associated with a given optimization constraint 2150, and may set the importance 2156 for each optimization constraint 2150 as described above (minute, low, nominal, high and extreme), correlating to empirically predetermined penalty weights such that the greater the importance, the greater the predetermined penalty weight. Once the above selections have been completed, a calculation server 400 retrieves the selections above from relational database server 250 and configures the objective function according to the generic definition discussed above and the selections made during the selection process. The resulting configured objective function equals the sum of credit components associated with the selected optimization parameters plus the sum of penalty components associated with the selected optimization constraints. As previously described, the user may select possible methodologies of static, death penalty, dynamic, and adaptive for the penalty weights; is supplied with the possible methodologies of static, dynamic and adaptive for the credit weights; and the methodology of relative adaptive for both the penalty and credit weights, as disclosed in the '718 application above. Accordingly, the optimization process of FIG. 22 may then proceed to determine an acceptable derivative core loading pattern design. Unique Fresh Fuel Bundle Type Design Tool (N-Streaming) Based on one or both of an accepted core loading pattern from the modify core loading design tool and an accepted rod pattern from the modify rod pattern design tool, the unique fresh fuel bundle type design tool may be iterated to determine the different types and/or locations of fresh fuel bundles (available from the fresh fuel bundle inventory) that may be used for the desired power uprate core design. Alternatively, the user may implement this tool using the original test core design determined in block 600, in order to determine the placement and makeup of fresh fuel bundle types in the design, and output this design to one or both of the modify core loading pattern and/or modify rod pattern tools. The unique fresh fuel bundle type design tool, hereafter “N-Streaming tool” for convenience is described in detail in co-pending and commonly-assigned U.S. application Ser. No. 10/325,831 to David J. Kropaczek et al., filed Dec. 23, 2002 and entitled “Method and Arrangement for Determining Nuclear Reactor Core Designs”. This tool is adapted to determine N unique fresh fuel bundle types for the power uprate core design. Although the sequence of operations to generating the test core design for N-streaming is a slight modification from that described in FIGS. 5-11, the test core design is simulated as per FIG. 13 and objective function values calculated as per FIG. 14. As to the simulation, the difference in FIG. 13 is that in step S31 the test core design incorporates a certain unique subset of fresh fuel bundle types that is converted into a 3D instruction set. Moreover, a desired control blade sequence (rod pattern) determined from iteration of the modify rod pattern tool (and base core loading pattern (the placement and make-up of once and twice burnt fuel bundles, and fresh fuel bundles other than the unique subset being evaluated) may be set for the test reactor core design (and incorporated as well into the 3D instruction set) for the test core design to be simulated. FIG. 25 is a flow chart illustrating the functionality of the N-Streaming tool in accordance with an exemplary embodiment of the invention. As described in FIG. 11, a reactor plant is selected for evaluation in Step S5 and limits which are used in a simulation for a test rod pattern design of the selected plant are defined (Step S10). These steps have previously been performed by the functions in input block 500, but are reiterated herein for clarity. As previously shown in FIG. 5, the client-inputted plant specific constraints, which may be configured as limits on input variables to the simulation and limits on the simulation results are set for the simulation. A test core design with an initial fresh fuel loading pattern is then generated (Step S15) for the selected reactor. For example, historical core loading pattern design database 254 may be accessed to find a historical reactor core design most consistent with the defined limits. A historical core design may be consistent if it is of a similar core size and power output rating, has similar cycle energy, and has similar operational performance characteristics to the core design being developed for the selected reactor plant. Using the similar historical design as a basis, and similar to as described in FIG. 6, the total energy content of the historical core may be calculated and a difference from the required energy content (e.g., the desired energy output from the determined core design, as based on customer requirements for example) defined. The difference in energy between historical core the energy content desired should be supplied by the loading of fresh fuel assemblies. Thus, to generate the reference core design, the user should select (Step S20) fresh fuel bundle type(s) for the reference core design that can best meet the energy requirement(s) (which may be included in the limits) for the reactor core design to be developed. The bundles designs may be selected from the fresh fuel bundle design database 252, which provides a wide variety of fresh fuel bundle designs (or N streams) that have been previously created and modeled. FIG. 26 illustrates a screen shot of a bundle selection web page 2600. Entitled “N-Streaming”, a user may bring up page 2600 via GUI 230 using a suitable input device, such as a modem, keyboard, pointer and the like. A plurality of selectable fresh fuel bundle types 2605 may be displayed; these bundle types 2605 have been previously modeled, so information relating to the performance of these bundle types 2605 is readily available to the user. The user may then select desired bundle types to be used in the loading pattern of the reference core design by checking boxes 2610. With the fresh bundle types selected, core loading symmetries are accounted for as described in step 606 of FIG. 6. Then, one or more current fresh fuel bundles in the test core design may be replaced (Step S25) with one or more of the selectable fresh fuel bundles 2605 during an iterative improvement process. The selection may be performed via GUI 230, which provides the user with a summary of each bundle's performance characteristics. Once the “N-streaming” (selected fresh fuel bundles) have been defined, a looping process described in terms of Steps S25 and S30 is initiated, whereby a systematic process of replacement and analysis for fresh fuel bundles is performed. In an example, at an outermost level (“outer loop”) each fresh fuel location in the current reference core design is examined in sequence. By “examined”, reactor core operation is simulated (Step S30) similar to as described in FIG. 13 for the test core design with each particular unique fresh fuel loading pattern, and performance characteristics of the N-streams (unique fuel bundles) are reviewed to determine whether the test (or derivative) core design can best meet the power uprate energy requirement(s) for the reactor core design to be developed. At the innermost level, each “replacement” fresh fuel bundle 2605 selected from page 2600 is examined in each fuel location (S35). During this process, a current fresh fuel bundle in the reference core design is replaced with each new “N-streaming” fresh fuel bundle 2605. As discussed above, reactor operation is simulated (Step S30) on the test core design containing one or more of the select fresh fuel bundles, in order to produce a plurality of simulated results, or outputs. The iterative steps of replacement and simulation are repeated (output of Step S35 is NO) until all selected fresh fuel bundles have been inserted at each fuel location and each derivative core design has been simulated (e.g., output of Step S35 is YES). Substitution of all selected fresh fuel bundles 2605 into each of the fresh fuel locations is therefore complete upon exiting the inner and outer loops. The iterative improvement process described above may be beneficial in that it enables the user to fine tune the fresh fuel loading pattern for the core design, and to perhaps extract even more energy out of an acceptable core design than was previously possible of doing with the conventional, manual iterative process. Further, incorporation of the relational database server 250 and a number of calculation servers 400 expedite calculations, reducing processing time to hours instead of weeks. The outputs from simulation are ranked based on the limits (Step S40), as described in FIG. 14. A user may display data related to each of the outputs, if desired. This enables a user to make a comparison against the reference core design to determine whether there was any improvement, where improvement may be defined in terms of not exceeding the defined limits, or meeting certain energy requirements, for example. If the top ranked output is an improvement (output of Step S50 is YES) the core design corresponding to that highest ranked output is set (Step S80) as the new core design (and may be input for iteration of the core loading tool or rod pattern tool to refine the design) with the results stored (Step S90) in relational database server 250, such as in simulator results database 255. This completes a single iteration of the iterative improvement process. Steps S25, S30, S40 and S50 (inclusive of steps S30 through S90) are repeated (e.g., N iterations), with each “improvement” becoming the new reference core design for a subsequent iteration. The defined limits are applicable to the reference core design in each of the N iterations. If, for a given iteration, there is no improvement in the top ranked output, the iterative process is complete, and data relating to the reference core design at that point, since it is the top ranked design may be displayed and interpreted (Step S60) by the user. The data may also provide the user with an indication of which location in a simulated core were the largest violators or largest contributors to a limit violation. At Step S60, the user may be inclined to initiate a modify subroutine (Step S70) similar to that described in FIGS. 18A-B and 23A-B. In an example, this could be an optional step, since the rod pattern and core loading pattern for the uprate power core design may already have been determined and set with the exception of fine tuning the location and/or makeup of the fresh fuel bundles to achieve an acceptable power uprate core design. As discussed above with the modify rod pattern and modify core loading tools, procedural recommendations may be accessible by the user. Even if the test core design with the desired N streams of fresh fuel bundles meets all of the limits (client-inputted plant specific constraints, design limits, thermal limits, etc.) the user may verify that any excessive margin to a particular limit is converted into additional energy. Accordingly, the following logic statements may illustrate the above procedural recommendations for modifying the fresh fuel loading pattern for the test or derivative core design: Energy Beneficial Moves If Critical Power Ratio (CPR) margin is too low towards core perimeter, bring more reactive fuel toward core center If NEXRAT (Nodal Exposure Ratio, a thermal margin constraint) problem at end-of-cycle (EOC), move more reactive (e.g., less exposed) fuel to problem location; If ShutDown Margin (SDM) problem at perimeter of core at beginning of cycle (BOC), place less reactive fuel towards perimeterEnergy Detrimental Moves If CPR margin too low at EOC, move less reactive fuel into problem location If kW/ft margin too low at EOC, move less reactive fuel into problem locationConverting Excessive Margin into Additional Energy If extra CPR margin in center of core at EOC, move more reactive fuel from perimeter locations to core center The data resulting from the objective function calculations may be displayed as a list of constraints with denoted violators, as shown in FIG. 15 and Table 1. The user may again be presented with graphical displays as shown in FIGS. 16 to 17B to see the effect of the core design with unique fresh fuel loading pattern on certain constraints or parameters, to give the user an indication of those fresh fuel locations in the test core design that may need modification. FIGS. 27A and 27B are flow diagrams describing the modification of the N-streams (unique fresh fuel types) in accordance with the invention. The functions described in these figures are similar to FIGS. 18A and 18B or FIGS. 23A and 23B, and thus similar notation has been used for clarity with the differences discussed in more detail below. As previously discussed, the user may direct each iteration of this modifying subroutine, with the help of the graphical user GUI 230, or the modifying subroutine may be performed within the bounds of an optimization algorithm that automatically iterates simulation, calculation of objective function and evaluation of the results or values of the objective function calculations for a number of core loading pattern design iterations. Referring to FIG. 27A, by interpreting the data at Step S60, the user may be inclined to initiate a modify subroutine (Step S70). In such a case practicality, the original test core design will not be an acceptable design, and the modify subroutine may be required if the iterative improvement process fails to provide a core design that is acceptable to the user, such as may be the case where certain limits which shall not be violated are still violated with each iteration. As with the modify rod pattern and modify core loading tool, the user may direct each iteration of this modifying subroutine, with the help of the graphical user GUI 230, or the modifying subroutine may be performed within the bounds of an optimization algorithm that automatically iterates simulation, calculation of objective function and evaluation of the results or values of the objective function calculations for a number of core loading pattern design iterations. The user determines, based on the displayed data, whether any limits are violated (Step S71). If no limits are violated, the user determines if any identifiers indicate that characteristics of maximum energy are obtained from the core design (i.e., indication of good thermal margin utilization (such as margins on MFLCPR and LHGR) by moving fuel so as to maximize plutonium generation for cycle extension, or where the minimum EOC eigenvalue is obtained for the core design to be used for the fuel cycle (eigenvalue search) or the desired cycle length is determined at a fixed EOC eigenvalue. If there is an indication that maximum energy has been obtained from a core design (the output of Step S72 is YES), the acceptable core design with desired fresh fuel loading pattern has been determined, and the user may access a report of the results related to the core design (Step S73) and/or use the selected fresh fuel loading pattern in the modify rod pattern design tool and/or as an input back to the modify core loading pattern tool (Step S78). If limits are violated (the output of Step S71 is YES) or limits are not violated but there is an indication that maximum energy has not been obtained from the core design (the output Step S72 is NO) then the user determines a fresh fuel loading pattern modification to be made to the current reference core design (Step S74). In making a modification to the fresh fuel loading pattern, and based on the recommendations from above, the user may alter the fresh bundle loading via the GUI as described in FIG. 24, such by identifying the bundle symmetry option of any potential fresh bundle(s) in the test core design to be moved, and selecting the “target” fresh fuel bundle(s), the destination(s) where the target bundle(s) is/are to be moved. The identified target bundles are then shuffled according to the required symmetry (mirror, rotational, etc.). This process may be repeated for any fresh bundle shuffle that is required to re-load the core reference pattern in the desired manner. Further, the test core design modifications may be saved in simulator input parameters database 259, for example. The user may repeat steps S30 to S50 (Step S75) incorporating the design modifications. The resultant highest ranked output establishes a new reference core design from which the iterative improvement process of FIG. 25 may be repeated. In other words, Steps S30-S50 may be repeated to determine if the derivative core design meets all limits (Step S75). This may become an iterative process, as described above. Referring to FIG. 27B, the modify subroutine in Step S70 is iteratively repeated until all limits are satisfied, or until all limits are satisfied within a margin that is acceptable, as determined by the user (Step S170). The functions of FIGS. 27A and 27B may be iterated over N different fresh fuel loading pattern designs, in an effort to consistently improve toward a desired fresh fuel pattern design that satisfies all user limits and constraints, for use in a nuclear reactor core. Assuming the core loading pattern for exposed fuel and rod patterns have been set to satisfy the limits for power uprate operations, this fine tuning determines the desired fresh fuel loading pattern, that could with exposed fuel and rod pattern designs, generates the power uprate core design reported to the customer at block 1100 in FIG. 4 Referring again to FIG. 20, for optimizing the iterative N-streaming process, the user would check box 2042 for the optimize fuel loading and optimize bundle selection parameters, and select desired optimization constraints 2150 and set the importance for each optimization constraint 2150 as described in FIG. 21. Once the above selections have been completed, a calculation server 400 retrieves the selections above from relational database server 250 and configures the objective function according to the generic definition discussed above and the selections made during the selection process, and the optimization process in accordance with FIG. 22 may then proceed to determine an acceptable derivative core design with the desired fresh fuel loading pattern. Accordingly, the N-streaming tool permits any number or combinations of fresh fuel bundle loading pattern designs (e.g., “N streams”) to be utilized in order to determine an accepted core design that satisfies the plurality of limits or constraints that have been input by a user for uprated power operations. The thus determined power uprate core design may include N new fresh fuel bundle solutions therein, for example. In contrast to current reactor core designs, which typically utilize at most one or two fresh fuel bundle types (i.e., a one or two stream solution), any number or combinations of fresh fuel bundle loading pattern designs for location and type of fresh fuel bundle (e.g., “N streams”) may be utilized in order to determine the desired fresh fuel bundles for placement in the core design. In an example, the N-streaming methodology may be used to determine a core design satisfying the limits and/or constraints for uprated power operations with N unique fresh fuel bundle types (N streams), where N equals or exceeds at least 2 unique fresh fuel bundle types (N≧2). Accordingly, as described here above, each of the rod pattern, core loading pattern and unique fresh fuel bundle type design tools may be iterated sequentially and/or together by the user and can provide feedback to one or both of the other tools, until all rod (control blade), exposed fuel and/or fresh fuel type changes have been exhausted in the test core design and/or a given “candidate” modified test core design satisfies each of the limits or constraints set for uprated power operations. For each automated tool, operation or iteration of the selected automated tool to evaluate the test core design against the input limits or constraints includes performing a simulation based on the constraints (inclusive of limits or targets for power uprate operations), in order to produce a plurality of simulation results for comparison against the constraints. Data indicating the constraints that were violated by the test core design during the simulation may be accompanied by procedural recommendations to modify one or more of the rod pattern core loading pattern and or unique fresh fuel bundle type makeup in the core design. One, some or all of the automated tools are implemented and iteratively repeated until a test core design is determined which meets all constraints for uprated power operations. This design thus represents an acceptable power uprate core design. As discussed above, the comparison includes configuring an objective function to generate a corresponding objective function value for each output using the objective function. The objective function values are evaluated based on the constraints to determine which of the outputs violate a limit. Based on procedural recommendation and/or by the user's own choosing, the user may then modify the test core design via GUI 230 to create a derivative reactor core design. Modifications may be made to control blade placement (modify rod pattern design tool), exposed fuel placement (modify core loading pattern design tool), fresh fuel bundle type make-up and placement (N-streaming design tool), and repeating simulation to evaluate if there is any performance improvement in a given derivative test core design. Referring back to FIG. 4, if after all tools have been iterated and all rod (control blade), exposed fuel and fresh fuel bundle changes have been exhausted, one or more limits from block 500 are still not satisfied, the user may be required to make global changes to exposed fuel to reload in block 600 and to reload the test core in accordance with the new exposed fuel loading pattern, as described in FIGS. 8 and 9. As previously discussed, if limit violations still exist after all modifications in block 700 have been exhausted, the user may return to block 600 and evaluate whether all exposed fuel bundles from the fuel pool inventory have been used. The user may change the exposed fuel bundles that are to be reloaded in the core design as described in FIGS. 8 and 9, accessing a different combination of exposed fuel bundles from the fuel pool with the filter window of FIG. 9, for example. If one or more limits from block 500 are still not satisfied, the user may be directed back to the processes in core loading block 600 and modify the original fuel loading template, inclusive of changing the exposed and fresh fuel placements of bundles, in order to change the locations in the template for insertion of different exposed and/or fresh fuel bundles selected from existing inventory. Once these changes have been made, a modified or derivative core design may be re-evaluated using one, some or all of the automated tools in block 700 until all limits have been satisfied and/or are within an acceptable margin to the limits as determined by the user. If the example methodology implemented by the user has produced an accepted power uprate core design that satisfies the limits and/or is within acceptable margins thereto, and the user has received the data and design parameters (e.g., via suitable reports and graphical representations) corresponding to the acceptable power uprate core design at block 1100, the plant operators and/or staff may modify the actual reactor core being evaluated in accordance with the accepted design at a next planned outage to achieve the desired increased thermal output which satisfies NRC licensing requirements. However, the user or designer may desire to fine tune the design using the latest actual core operating conditions, and test each and every cycle of the core in implementation block 1200. The functions in implementation block may be understood as performing an exposure accounting process (referred to as a core tracking routine) to get actual conditions and limits, performing a revised margin analysis using the accepted power uprate core design to determine revised margins to limits for power uprate. The revised margins and actual operating conditions and limits can be incorporated into an “online operations optimization routine”. This latter function incorporates the modify rod pattern tool and associated simulation, evaluation using the objective function and iteration of modifying the rod pattern using the optimization routine of FIG. 22. As these functions have already been described, attention is directed to an overview of the core tracking and revised margin determination functions in implementation block 1200. Core Tracking FIG. 28 is a flow diagram for illustrating the core tracking function in more detail. Typically, a process computer at the reactor plant being evaluated has a process computer that records data (operating and limit data) from the core at a particular time, such as daily. This recorded data is written to an exposure accounting file (a text file) in a particular format depending on the core monitoring program used. Known core monitoring programs use two primary formats, a 3DM format and a PowerPlex format. In accordance with this core tracking function, these exposure accounting files may be sent from the plant's process computer through a firewall to a drop box or computer (such as host processor 210) within system 1000, to be stored in memory (i.e., at a desired location in the relation database server 250, for example. Exposure accounting files for each cycle of the plant (historical up to current) may be stored for access in server 250, such as in a global exposure accounting database within server 250. Referring now to FIG. 28, the user, using a suitable input device (mouse, keyboard, touch screen, voice command, etc) and GUI 230, may initiate a core tracking operation by the user (2810) and selects a new exposure accounting for the plant of interest (2820) by selecting a given icon on a webpage for example. The user is queried to create a new exposure accounting case (2830). In 2830, a new dialogue may appear on the display and the user is prompted to enter inputs such as cycle number, cycle start date, and to initiate a wrapup operation in which a wrapup file including all cycles from the plant is copied from the exposure accounting databases in server 250 to a temporary directory for access by the user. A new cycle ID strategy ID may be created for the new case, with the information from the inputs written to the database. The new exposure accounting case is ready to accept exposure data. The user then selects the exposure accounting files for the case in the cycle of interest (2840). Each file may be processed in turn. In 2840, the user may be prompted to add exposure accounting files and browse the temporary directory to select and upload the desired files. Each file is then parsed (2850), with the parsed or extracted data sent to the exposure accounting database. All files are to be parsed without error, thus there is an error checking routine which informs the user as to which files suffered an error in the parsing operation, and the reasons for the error. For example, in the parsing operation, parsed data from each exposure is held in memory to allow for sorting and error checking. The following example accounting data is read, including but not limited to: core name, date and time, power, flow, sub-cooling, core dome pressure, measured data for the local power range monitors (LPRMs), such as MFLCPR, MFLPD and MAPRAT(MAPLHGR), positions of control blades (rods), axial relative power (core average radial power distribution) and feedwater flow. The parsed exposure data is sorted by exposure and checked for the following: (i) all exposures from correct plant, (ii) all exposures from correct cycle, (iii) date and exposure always increase, (iv) exposure not later than EOC date, (v) exposure filename does not already exist, (vi) additional error checks (checksum). When no errors are found, the parsed data is written to the database, but are copied on server 250 so they can be viewed via GUI 230. A simulation is then run using the accepted power uprate core design ((2860), in this example, it may be a PANAC simulator) using the parsed data of a single exposure accounting file, and stored (2870) in the exposure accounting database in server 250. The data output from the simulation is read by the user via GUI (2880), and then the next file is uploaded (2890), with the routines in 2830 to 2880. Accordingly, the core tracking function allows accounting data from all exposures in a given cycle of interest to be stored in the server 250. Revising Margins to Limits for the Uprated Core Design Conventionally, differences between given on-line margin (actual plant) and given off-line margin (virtual or modeled core being simulated for the actual plant of interest) determinations to operating limits exist. As used hereafter, operating limits refer to thermal, reactivity and/or power-related operating limits. These differences force plant operators to require additional margin to the operating limits, so as to insure trouble free operation. Additional margin typically is obtained by making changes to the operational parameters, and/or by selection and positioning of different rod patterns. However, the cost of such changes typically is a loss of power or fuel cycle efficiency. Moreover, a “larger than needed” margin requirement has an adverse economic impact on the plant. To protect against these differences, engineers have developed standard design margins or historical design margins that are to be used to account for or “cover” these differences. However, these standard design margins are crude at best. Sometimes, the historical required design margin is inadequate, resulting in manipulation of control rods during operation in order to regain lost margin. If rod pattern changes do not alleviate or correct the problem, plants have been even known to have to de-rate (lower power production). Either solution is extremely costly to the fuel cycle efficiency and can cost millions of dollars in lost revenue. Additionally, the historical design margin is occasionally inappropriately conservative, thereby resulting in a reduction in possible fuel cycle efficiency. For the example power uprate core design, revised margins are even more critical to proper plant operation. Accordingly, the core tracking data described in FIG. 28 may be used in conjunction with the accepted power uprate core design to simulate reactor operation and determine revised margins to these limits, inclusive of the limits for operating the core in excess of rated thermal output. FIG. 29 is a process flow diagram to explain the method of determining a revised margin to a given operating limit. The example methodology is described in co-pending and commonly assigned U.S. application Ser. No. 11/320,919 to William Earl Russell, II, et al., filed Dec. 30, 2005 and entitled “Method of Determining Margins to Operating Limits for Nuclear Reactor Core Operation”. The relevant portions of the '919 application describing the margin determination are hereby incorporated in their entirety by reference herein. In general, data in the exposure accounting files stored in server 250 may be used to model the on-line reactor plant being evaluated for the particular cycle, matching the operating parameters at the current exposure in cycle, so as to execute an off-line simulation (e.g., an executable 3D simulator program such as PANACEA, LOGOS, SIMULATE, POLCA) of the accepted power uprate core design. The simulation provides results including predicted margins to given operating limits for the uprated core design in the cycle of interest, which hereafter may be referred to as “predicted dependent variable data”. The predicted dependent variable data may be stored in server 250, and is also provided to a calculation processor 400, which is to be used for determining a revised operating margin to a given operating limit. Once the margin value has been calculated, this data may be used by processor 400 to determine revised operating parameters for the on-line plant being evaluated, and may be communicated to plant operators at plant so as to change operating parameters (i.e., control rod sequence, core flow, power level, etc.) at the current exposure (time in operating or energy cycle) or at a future point in the current operating cycle of the plant. These margin calculations may be performed continuously at any desired frequency or periodicity, in an effort to maximize plant efficiency, for example. Referring now to FIG. 29, in the example method 2900, operating conditions and monitored parameters are accessed (2910) during a current operating cycle from the operating plant being evaluated by the process computer and saved to database 250. This data is transferred to system 1000 as described in FIG. 28 and all exposure accounting files for the cycle of interest in the plant being evaluated are created and stored in the exposure accounting database within relational server 250. For the exposure accounting files, example independent variables (i.e., rod pattern, operating conditions such as reactor power and core flow, plant configuration, mechanical conditions, core conditions, enrichment and gadolinium properties, cycle exposure, etc.) are thus saved to the exposure accounting database in order to correlate any potential trends between simulation biases and core configuration. Similarly, all monitored results or dependent variable data such as Maximum Fraction of Limiting CPR (MFLCPR), MFLPD, MAPLHGR, cold shutdown margin, reactivity-related parameters (such as Hot Eigenvalue, etc.), and predicted margins to these operating limits are also saved to database server 250. The plant operating conditions retrieved from the process computer by the core tracking routine in FIG. 28 may thus be understood as independent variables, and monitored or measured operating limit data (thermal and power-related limits and margins thereto) retrieved by the process computer is actual dependent variable data. These independent and dependent variables from one or more exposure points in the current operating cycle may thus be saved or stored in database server 250. With the above information saved to the database server 250, a reactor simulation input file can be created or prepared. The simulation input file uses identical independent variables as described above and may be stored in an electronic file format (i.e., ASCII) that is recognized by the identified core simulation software program (off-line simulator). Once the input file is prepared, the off-line simulator executes its program (2920) to simulate reactor operation of the plant off-line and to generate simulator outputs. The simulator outputs are a prediction of the dependent variables, referred to as predicted dependent variable data. The predicted dependent variable data may be understood as a nominal estimate of future results, and therefore may be used to calculate a nominal estimate of operating margins, but does not take into account any uncertainty for the predictions. Ideally, the off-line simulated dependent variables (predicted margins to limits such as MFLCPR, MFLPD, MAPLHGR, etc.) and the measured or actual dependent variable data (actual margins to MFLCPR, MFLPD, MAPLHGR, etc.) from plant operation would be identical. However, due to several (or more) of the factors identified above, these typically are not. At this time, the predicted dependent variable data is normalized (2930) with respect to time (exposure) relative to anticipated EOC (End of Cycle). In other words, the data is normalized by calculation processor 400 on a BOC (Beginning of Cycle) to EOC time range of 0.0 (BOC) to 1.0 (EOC). In doing so, the normalized predicted dependent variable data can be evaluated with results (such as normalized historical dependent variable data) from many reactor simulations of other plant cycles, with the normalized data being stored in database server 250. Relational database server 250 contains a substantial collection of reactor simulations (in subordinate databases 255, 257, and 259), and hence, includes a substantial amount of historical dependent variable data from reactor simulations of operating cycles in other reactor plants. For example, because the assignee has provided fuel and engineering services for approximately 30 BWR's over approximately 20 years, almost 400 complete exposure depletion cycles are available (given an approximate 1½ year average cycle length). A collection of data for 400 operating cycles is a significant collection of information for evaluating operation of nuclear reactors. This information can be utilized by the example methodology of the present invention and for the resulting predictions there from. For example, as part of step 230, the processor 400 retrieves historical simulation data from plants having a similar plant configuration to plant. This historical dependent variable data is also normalized with respect to time on the 0.0 to 1.0 scale for evaluation, although any other normalized scale could be employed, as would be evident to one of skill in the art. While all of this data has been normalized with respect to time (exposure) so that all data ranges from 0 to 1 (0.0=BOC, 1.0=EOC), some of the operating strategies for the various stored operating cycles are dissimilar. Consequently, it may be desirable to filter the larger collection of cycle data in database server 250 to collect a sub-set of data that is most similar in plant operation style to the specific plant being evaluated. Filtering parameters may include, but are not limited to: cycle length, power density, average gadolinia concentration, flow strategies, loading strategies, etc. Thus, the filtered historical data incorporates data from similar plant operation styles. As a result of the above filtering process, predicted uncertainties may become smaller, and may be used to improve fuel cycle efficiency. Similarly, it may also be desirable to provide correlation to the above continuous variables by way of least squares, neural networks, or any other trend capturing mathematics. In doing so, a larger set of data can be incorporated and global trends may be included, possibly resulting in a further reduction in predicted uncertainties, and may be used to improve fuel cycle efficiency. FIG. 30 is a graph of calibrated time-dependent bias for a given operating limit versus normalized time to explain the calculation of a time-dependent average bias value in accordance with the example method. FIG. 31 is a graph of time-dependent uncertainty versus normalized time to explain the calculation of the time-dependent uncertainty value in accordance with the example revised margin determination method. In general, the normalized historical dependent variable data will be used by processor 400 in order to calculate a time-dependent average bias value that will provide a predicted expected bias at all future times in the cycle for the predicted dependent variable data (such as a given margin to a given operating limit) calculated as a result of the off-line simulation of plant. In FIG. 30, there are shown time-dependent bias curves for 30 identified operating cycles of plants having a similar plant configuration to plant. This information is retrieved from database server 250 by processor 400. For each historical cycle being evaluated, a bias value for the historical dependent variable data is known and has been calculated in advance (and stored in database sever 250). The known bias value at a given exposure point, for a given stored historical operating cycle, represents a difference between the measured and the predicted operating limit at that exposure point for the given historical cycle. Once the selected data has been collected for all 30 cycles between 0.0 to 1.0 (the data here being the known bias values along all exposure points for each of the historical dependent variable data of each historical operating cycle), the data is calibrated relative to the current time in operation of the operating cycle of plant being evaluated, e.g., the point in cycle time being evaluated. For example, and referring to FIG. 30, if cycle operation of plant 2910 is approximately 10% complete (t=0.10), all data (all bias values) at all time intervals should be calibrated upwards or downwards until the value at t=0.10 is set to zero on the y-axis (ratio of measured minus predicted dependent variable). Calibrations to bias values after t=0.10 would be adjusted to correct for the calibration. FIG. 30 illustrates how the multiple curves (30 curves) are calibrated at t=0.10. Whereas the above example identifies calibration by way of addition-subtraction to set values to zero at t=0.10, calibration can also be performed using multiplication-division to set values to one at t=0.10. Calibration by way of addition-subtraction or multiplication-division may be selected by the mathematical processes which provide the smallest prediction of future uncertainty. As can be seen from FIG. 30, all lines go through zero at t=0.10. This is because at any given current time (t=0.10 in this example) there is a known exact bias between the off-line simulation results (predicted margin) and operating plant-measured result (actual margin to the given operating limit). From the calibrated curves, two time-dependent curves are determined. First, the time (exposure) dependent bias value is determined (2940) by averaging all of the future data (t>0.10). In the above example, the data is somewhat random and the time dependent bias for all future times is approximately 0.0. In FIG. 30 this time dependent bias value is shown as curve 300, which is the average of the bias values of the 30 curves at each evaluated exposure point between t=0.1 (current time) and t=1.0 (future time). Accordingly, to calculate the time dependent bias value (curve 300), the normalized historical dependent variable data is calibrated to force the known bias values to the current exposure point in the operating cycle of plant. The time-dependent average bias value is this determined by averaging all the bias values of the normalized historical dependent variable data at each of the exposure points, as calibrated from the current exposure point in plant being evaluated. Next, and as shown in FIG. 31, a time (exposure) dependent uncertainty is determined (2950). This is determined by calculating the standard deviation at all times greater than the present time (in this example, all times greater than t=0.10). An example of a time dependent uncertainty curve is shown in FIG. 31. The curve in FIG. 31 represents the standard deviation at each exposure point of the time-dependent bias curve 300 in FIG. 30. In FIG. 31, it can be seen that the generally parabolic shape of the curve indicates that the uncertainty in the bias value goes up over time. Thus, if the designer knows where he is at any point in the cycle (past or present), such as at t=0.2, the curve can be used to predict the uncertainty in the bias value for the predicted dependent variable data at any other future time in the cycle. An observation can be made by studying the curves in FIGS. 30 and 31 in greater detail. There is an exact and simple correlation that relates all of the uncertainties, represented as “σ”, of a random system to time. If the uncertainty σ of the system is known at any point (example t=ref), the uncertainty σ of any other point can be calculated by the following set of equations in (1):σtarget=σref[ttarget/tref]1/2 or rewritten as,σ2targettref=σ2refttarget or rewritten as,σ2target/σ2ref=ttarget/tref (1) The last equation of (1) illustrates the relation used to determine required future dependent variable uncertainties for a modeled independent variable measured-to-predicted system, In (1), ttarget=desired time of desired uncertainty, tref=reference time where uncertainty of system is known, σref=reference uncertainty at reference time (tref) and σtarget=desired uncertainty at desired time (ttarget). As shown by the last equation of (1), relative time therefore equals relative uncertainty and relative uncertainty equals relative time. Therefore utilization of this relation can provide a determination of future uncertainties. Consequently, given the amount of future time that is required (i.e. the next control blade sequence interval) and data from a reference time, a good estimate of the required future uncertainty can be determined. The combination of this information can provide maximum fuel cycle efficiency while simultaneously providing event-free operation. Extensive computer experiments have been performed to confirm that expression (1) is exact as the number of uncertainty curves for nuclear reactors increase to infinity. FIG. 32 is a graph to assist explanation of how a margin for a given operating limit is calculated based on the time-dependent bias value and time-dependent uncertainty value so as to satisfy a risk-tolerance level set for the plant being evaluated. Now that a time dependent bias (2940) and a time dependent uncertainty (2950) have been determined, this information can be used to determine a required or revised margin to the given operating limit (2970). This calculation is based on obtaining a risk-tolerance level 2960 for the plant being evaluated. The risk-tolerance level may be understood as a desired predictability of meeting the operating limits of a customer's reactor plant, or in other word, a probability of an event not occurring in plant during a given period in the current operating cycle. For example if the number of historical data points is large (such as greater than 30) and the customer wanted a given probability (i.e. 90%, 95%, 99%, 99.9%, etc) of operating their nuclear reactor with fixed rod patterns for the first sequence of cycle operation, the following margin in FIG. 32 would be required. Where a smaller set of historical data points is used or specific confidence levels are required, multiplier constants known as probability K values should use appropriate confidence corrections. In FIG. 32 (t=0), curve A represents the actual operating limit of any required thermal or power-related result (MFLCPR, MFLPD, MAPLHGR, etc). This is a line that should not be exceeded during the operation of the plant. The curve B represents the needed design target to make sure the operating limit is not violated at any future time. If the customer wanted to make sure that a first sequence of operation (t=0 to t=0.1) would not require any rod pattern modification, they would utilize the required margin prediction at t=0.1. In FIG. 32, a design target operating limit of 0.971 would provide a sufficient margin to ensure, with 99.0% probability, that rod adjustments will not be needed (see curve C at t=0.1). Similarly, in FIG. 32 a design target of 0.953 would be required to ensure, with 99.9% probability, that rod adjustments will not be required. The 99.9% represents the risk tolerance level of the customer for this “non-event”. Accordingly, the probability value or risk-tolerance level is used to determine the multiplier constant K that is to be multiplied by the time-dependent uncertainty σ, or σtarget=Kσref, where σref is the known reference uncertainty at a given point in time, which provides a prediction of the uncertainty at any point in the cycle. In either case, a customer specific or plant specific solution may be easily determined. In most cases, determining an operating margin based on a desired predictability of meeting the operating limit may provide additional margin for greater operating flexibility and superior fuel cycle efficiency (higher than the historic design target limit of curve D). In any case, the example methodology may reflect a more knowledgeable plan for reactor operation. Based on the revised margin calculation at 2970 by processor 400, the designer can then revise plant operating parameters (2980) using processor 400, either by hand (manual calculations) or using an optimization routine to determine the desired rod pattern, core flow, power level, etc. Any suggested changes may be forwarded to operators of the plant being evaluated to change the operating conditions during the current cycle, if necessary or desired. Referring now back to FIG. 4, once the core tracking routine of FIG. 28 and the revised margin determination routines of FIGS. 29-32 have been completed, the modify rod pattern tool as described above may be iterated in the cycle of interest, for the accepted power uprate core design, using the revised margins and current operating conditions determined from the core tracking and revised margin routines. Accordingly, the limits in block 500 are updated and follow on simulation, evaluation and possible rod pattern modification sub-processing is performed until all limits have been satisfied and/or are within an acceptable margin thereto as determined by the user or designer. Moreover, the processing steps 12a-c in block 1200 may be repeated (block 1300) for each energy cycle until all energy cycles have been evaluated for the accepted power uprate core design. Therefore, in accordance with the example embodiments, there is provides a method and system for designing a nuclear reactor core for increased power operations. The system includes a GUI and processing to simulate a test initial reactor core design in an uprated power environment. Constraints or limits related to requirements for operating a reactor core above 100% of its currently-licensed power level are input by the user via the GUI, and an initial test reactor core design is generated based on the limits. The user then may select one or more automated design tools to modify the rod pattern (control blade pattern) core loading pattern and desired unique fresh fuel bundle types (fresh fuel loading pattern) to meet power uprate requirements. For each tool, the user initiates a reactor simulation and utilizes an objective function to determine how closely a simulated modified core design meets the constraints. Via the GUI, the user may then modify a given core design with a change in bundle design, control rod placement, exposed or fresh fuel placement, etc., and repeat simulation to gauge performance improvement in the derivate reactor core design. The modifying, simulating and evaluating functions may be iteratively repeated until a core design satisfies all limits (inclusive of limits for uprated power operations) so as to operate a reactor loaded based on the acceptable power uprate core design at increased power, or within a margin to given limits(s) that are acceptable to the user and/or customer. As desired, the user may implement a core tracking function and a revise margin determination function to revise the accepted power uprate core design based on the current plant conditions and the power uprate requirement for operation therein. Upon determination of the revised margins to operating limits (thermal, power, etc.), the modify rod pattern tool may be iterated to confirm that the power uprate core design with the revised margins still satisfies all operating limits and constraints for operation in excess of its rated thermal output. Based on the accepted power uprate core design, the plant operators and/or staff may then modify their reactor core at a next planned outage to achieve the desired increased thermal output while satisfying NRC licensing requirements. The example embodiments of the present invention being thus described, it will be obvious that the same may be varied in many ways. For example, the functional blocks of FIGS. 1-4, 6, 11, 13, 14, 18A-B, 22, 23A-B, 25, and 27A-B through 29 describing the exemplary methodologies and system may be implemented in hardware and/or software. The hardware/software implementations may include a combination of processor(s) as shown and/or article(s) of manufacture. The article(s) of manufacture may further include storage media and executable computer program(s). The executable computer program(s) may include the instructions to perform the described processes or functions to determine a power uprate core design. The computer executable program(s) may also be provided as part of externally supplied propagated signal(s). The technical effect of the invention may be a system and/or method invoking the processing capabilities of multiple processors(s) and/or computer program logic of the programs implemented by the one or more processors to provide a way to efficiently develop a power uprate core design a nuclear reactor that satisfies all input limits and constraints for reactor operation in a given plant being evaluated. Additionally, the technical effect of the example embodiments provide a computer/processing-driven system for providing internal and external users the ability to quickly develop, simulate, modify and perfect a power uprate core design with a specified rod pattern, core loading pattern for exposed fuel and fresh fuel assemblies which satisfies all input limits including those limits related to reactor power operations in excess of the plant's rated thermal output. The accepted design may thus be loaded in the core of the evaluated reactor plant at a next scheduled outage, with the core operating in a next and subsequent cycles in accordance with the thus determined power uprate core design. Such variations are not to be regarded as departure from the spirit and scope of the exemplary embodiments of the present invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims. |
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description | The present invention relates to a novel process for reprocessing a spent nuclear fuel, based on uranium oxide or on mixed uranium-plutonium oxide, which makes it possible for the uranium and plutonium to be very effectively decontaminated from other chemical elements contained in this fuel without leaving, at any moment during this process, plutonium without uranium, so as to minimize the risk of misappropriating the plutonium for military purposes. The process of the invention also makes it possible to obtain, at the end of this decontamination, a mixed uranium-plutonium oxide powder that can be used directly in processes for manufacturing MOX (Mixed OXide Fuel) nuclear fuels, such as the MIMAS (MIcronized MASter Blend) process. At the present time, all plants for reprocessing spent nuclear fuels use the PUREX (Plutonium Uranium Refining by EXtraction) process for recovering the uranium and plutonium that are present in these fuels. This is obtained by carrying out several purification cycles using the liquid-liquid extraction technique (that is to say by mixing an aqueous phase and a solvent phase that are mutually immiscible, followed by separation of these two phases by settling) which is carried out in multistaged units of the mixer/settler type, pulsed columns or centrifuge extractors, which are connected in series so as to allow these cycles and the various operations that they comprise to be carried out continuously. The PUREX process, as implemented in modern reprocessing plants such as UP3 and UP2-800 plants on the Areva NC site at La Hague in France, or Rokkasho plant in Japan, schematically comprises three purification cycles: a first cycle, the purpose of which is essentially to decontaminate both the uranium and the plutonium from the fission products and from two minor actinides, namely americium and curium, and also to partition these two elements into two separate streams; and two complementary cycles called the “second plutonium cycle” and “second uranium cycle”, respectively, the purpose of which is to purify the plutonium and the uranium after their partition. The first cycle starts with an operation which consists in extracting both the uranium and the plutonium, the first being in oxidation state (VI), and the second being in oxidation state (IV), from the aqueous phase in which they are found. This aqueous phase is obtained by dissolving a spent fuel in nitric acid and clarifying the mixture thus obtained. This phase is commonly called the “dissolution liquor”. The coextraction of the uranium and plutonium is carried out by means of a water-immiscible solvent phase that contains an extractant having a high affinity for uranium(VI) and for plutonium(IV), in this case tri-n-butyl phosphate (or TBP) used with a concentration of 30% (v/v) in an organic diluent, in this case a dodecane. The uranium and the plutonium thus pass into the solvent phase, while most of the fission products, americium and curium remain in the aqueous phase. It is followed by one or more scrubbing operations in which the solvent phase is scrubbed by one or more aqueous nitric phases of different acidities, so as to remove the fission products from said solvent phase that were extracted with the uranium and the plutonium. The aqueous phase or phases resulting from these coextraction and scrubbing operations (or raffinates), which are laden with fission products, are removed from the cycle whereas the solvent phase, which is itself laden with uranium(VI) and with plutonium(IV), is directed to a zone in which the partition of these two elements is carried out. This partition comprises: an operation for the purpose of back-extracting the plutonium from the solvent phase by means of an aqueous nitric phase of low acidity, containing a reducing agent capable of reducing plutonium(IV), which is highly extractable by TBP, to plutonium(III) which is itself only barely extractable, and to do so without reducing the uranium, and also a nitrous acid scavenger, the role of which is to stabilize both the uranous nitrate and the plutonium(III) by destroying the nitrous acid that tends to form in the aqueous nitric phase; in this case the reducing agent is uranous nitrate, while the nitrous acid scavenger is hydrazinium nitrate, also called hydrazine. an operation whose purpose is to complete the back-extraction of plutonium from the solvent phase by means of an aqueous nitric phase, also of low acidity and containing uranous nitrate and hydrazine; and an operation whose purpose is to back-extract the uranium(VI) from said solvent phase by means of a very dilute aqueous nitric solution. Since the back-extraction of the plutonium from the solvent phase is accompanied by partial back-extraction of the uranium, the partition further includes a step whose purpose is to remove the uranium from the aqueous nitric phase resulting from the operation of back-extracting the plutonium by means of a solvent phase, of the same composition as that used for coextracting the uranium and plutonium. Thus, what are obtained after the first cycle are: a first aqueous stream that contains more than 99% of the plutonium initially present in the dissolution liquor and no longer contains any uranium; and a second aqueous stream that contains more than 99% of the uranium initially present in the dissolution liquor and no longer contains any plutonium. The first aqueous stream resulting from the first cycle is then subjected to the “second plutonium cycle”, the purpose of which is to complete the decontamination of the plutonium from the fission products liable to be still present in trace amounts in this stream. Thereafter, this stream which contains plutonium at a purity level of greater than 99.9%, is directed to a zone where the plutonium is converted into the oxide (PuO2) and then stored in this form, for the purpose of its subsequent use in the manufacture of MOX nuclear fuel pellets. In parallel, the second aqueous stream resulting from the first cycle is subjected to the “second uranium cycle”, the purpose of which is to complete the decontamination of the uranium from the fission products, but especially to separate it from the neptunium. This is because, in the first cycle, most of the neptunium present in the dissolution liquor is extracted, mainly in the form of neptunium(VI) at the same time as the uranium and the plutonium. During the reducing back-extraction of the plutonium in the first cycle, the neptunium(VI) is reduced by uranous nitrate to neptunium(IV), in which state it can be extracted by TBP, although less than in oxidation state (VI). The neptunium therefore almost quantitatively follows the uranium during all the operations of the first cycle, hence the need to subject the aqueous uranium-laden stream, resulting from the partition, to a complementary cycle suitable for stripping it of the neptunium before it is converted to uranium oxide. Like the plutonium resulting from the “second plutonium cycle”, the uranium has, after the “second uranium cycle”, a purity level of greater than 99.9%. It is also converted to the oxide and stored in this form. Moreover, a method has been proposed in U.S. Pat. No. 4,278,559 for recycling spent nuclear fuels with the aim of limiting, in all the stages of this recycling, the risk of plutonium being diverted for military purposes. This method is designed to obtain, after the step of coextracting, the uranium and plutonium in a solvent phase containing, apart from these elements, 0.1 to 10% of the fission products initially present in the dissolution liquor, then to obtain, during the partition step, a plutonium production stream diluted with uranium and containing most of the radioactive fission products present in said solvent phase. This plutonium production stream is then processed by a sol-gel method in order to obtain plutonium-uranium-fission products mixed oxide, which are subsequently used to manufacture fresh nuclear fuels. Although the presence of radioactive fission products in non-negligible quantities in fresh nuclear fuels is not an obstacle to the use of these fuels in fast neutron reactors, this is not the case as regards their use in the current light-water reactors. This is because such a use would make it necessary to develop a new type of fuel having a substantially increased content of fissile material because of the neutron-absorbing character of certain fission products and to carry out lengthy and expensive studies in order to obtain homologation of this fuel. Moreover, the presence of radioactive fission products at all stages in the spent fuel reprocessing and fresh fuel production chains, as provided in the method described in US-A-4 278 559, means having installations provided with radiation protection systems suitable for processing radioactive streams, and this being the case for each of the plants involved in this chain. Implementation of this method on an industrial scale would therefore necessarily involve either substantial reinforcement of the radiological protection with which existing nuclear fuel reprocessing and production installations are provided, or the production of new installations especially designed for operating a highly radioactive processing chain, which in both cases would result in a very considerable overcost. With a view to developing new plants for reprocessing spent nuclear fuel, the inventors were set the objective of providing a process which, like the PUREX process previously described, allows uranium and plutonium to be effectively decontaminated from the other chemical elements present in a spent nuclear fuel, and in particular from the fission products, but which, unlike the PUREX process, at no time leaves plutonium by itself, whether in the solid or liquid state. The inventors were also set the objective that this process should make it possible to obtain a mixed uranium-plutonium oxide that can be directly used for the manufacture of MOX nuclear fuels, whatever the purpose of these fuels: namely fast neutron reactors or light-water reactors. They were furthermore set the objective that this process should use, at least in part, the knowledge and know-how acquired in the PUREX process, in terms of both procedures and installations, so that it can be industrially exploited in the short or medium term. These objectives, and yet others, are achieved by a process for reprocessing a spent nuclear fuel and of preparing a mixed uranium-plutonium oxide, which comprises at least: a) a step of separating the uranium and plutonium from the fission products, the americium and the curium that are present in an aqueous nitric solution that results from dissolving the spent nuclear fuel in nitric acid, this step comprising at least one operation of coextracting uranium, in oxidation state (VI), and plutonium, in oxidation state (IV), from said aqueous solution by bringing this solution into contact with a water-immiscible solvent phase containing at least one extractant; b) a step of partitioning the uranium and plutonium coextracted in step a) into two aqueous phases, namely a first aqueous phase containing plutonium and uranium, and a second aqueous phase containing uranium but not containing plutonium; c) a step of purifying the plutonium and uranium present in the first aqueous phase obtained after step b) from the fission products also liable to be found in this phase; and d) a step of coconverting the plutonium and uranium that are present in the aqueous phase obtained after step c) into a mixed uranium-plutonium oxide. Thus, unlike the PUREX process described above, which provides, once the uranium and the plutonium have been coextracted from the dissolution liquor, for the complete separation of these two elements from each other, and subsequently processes them independently of each other, the process of the invention itself proposes to only partially separate the plutonium from the uranium and to keep it, throughout all the consecutive steps of this separation, in the presence of uranium until a mixed uranium-plutonium oxide is obtained. Preferably, step a) of the process of the invention includes, in addition to a coextraction operation, at least one scrubbing operation carried out on the solvent phase obtained after this coextraction in order to remove from this phase the fission products that were extracted together with the uranium(VI) and the plutonium(IV), this scrubbing operation being performed by bringing said solvent phase into contact with an aqueous nitric phase. Also preferably, step b) of the process of the invention comprises at least: b1) a step of back-extracting the plutonium, in oxidation state (III), and a fraction of the uranium, in oxidation state (VI), which are present in the solvent phase obtained after step a), by bringing this phase into contact with an aqueous nitric phase containing a reducing agent capable of reducing plutonium(IV) to plutonium(III) without reducing the uranium, for example uranous nitrate (uranium(IV) nitrate) or hydroxylammonium nitrate; and b2) an operation of back-extracting the uranium that has not been back-extracted from the solvent phase during operation b1), by bringing this phase into contact with an aqueous nitric phase. Two aqueous phases are thus obtained, one of which contains plutonium and uranium, while the other contains uranium but does not contain plutonium. Step c) of the process of the invention then preferably comprises, at least: c1) an operation of coextracting the plutonium, in oxidation state (IV), and the uranium, in oxidation state (VI), from the aqueous phase obtained after step b1), by bringing this phase into contact with a water-immiscible solvent phase containing at least one extractant in an organic diluent; c2) a scrubbing operation in which the solvent phase obtained after operation c1) is scrubbed in order to remove the fission products from this phase that were extracted together with the plutonium(IV) and the uranium(VI) during operation c1), this scrubbing operation being performed by bringing this phase into contact with an aqueous nitric phase; and c3) an operation of back-extracting the plutonium, in oxidation state (III), and a fraction of the uranium, in oxidation state (VI), from the solvent phase obtained after operation c2), by bringing this phase into contact with an aqueous nitric phase containing a reducing agent capable of reducing the plutonium(IV) to plutonium(III) without reducing the uranium, for example uranous nitrate or hydroxylammonium nitrate. It goes without saying that, if steps b) and c) are carried out in the manner that has just been described, then the process of the invention further includes, between these steps, an oxidation operation in order to reoxidize the plutonium(III) present in the aqueous phase obtained after operation b1) to plutonium(IV). This oxidation operation also allows the uranium(IV) also liable to be present in this phase to be reoxidized to uranium(VI) especially if the reducing agent used during operation b1) is uranous nitrate. Moreover, if step b) is carried out in the manner as just described, the aqueous phase obtained after operation b1) inevitably contains neptunium. It is therefore necessary to remove the latter if it is desired to obtain, at step d), a mixed uranium-plutonium oxide containing no neptunium. Therefore, in one embodiment mode of the process according to the invention it includes the removal of the neptunium present in the aqueous phase obtained after operation b1), either during step b) or during step c). This removal of the neptunium may, firstly, be carried out by adding to step b) an operation b3) of re-extracting the neptunium, in oxidation state (IV), from the aqueous phase obtained after operation b1), by bringing this phase into contact with a water-immiscible solvent phase containing at least one extractant in an organic diluent. Since uranium(VI) and neptunium(IV) behave in a relatively similar manner, a fraction of the uranium(VI) present in the aqueous phase obtained after operation b1) is back-extracted with the neptunium. Therefore, the invention provides the possibility of adding the uranium either to the aqueous nitric phase subjected to operation b3), or to the aqueous nitric phase subjected to operation c3), or to both, if it is deemed necessary for these phases to be recharged with uranium. In all cases, the uranium added may be uranium(VI) or uranium(IV). The removal of the neptunium may also be carried out during operation c2) by adding, to the aqueous phase used during this operation, a reducing agent capable of selectively reducing neptunium(VI) to neptunium(V) that is to say without reducing the plutonium or the uranium, and to do so in order to allow the neptunium to pass into the aqueous phase while leaving the plutonium and the uranium in the solvent phase. As a variant of the above it is also possible not to remove the neptunium present in the aqueous phase obtained after operation b1) but to let it follow the plutonium present in this phase as far as step d), so as to obtain a mixed uranium-plutonium-neptunium oxide. Therefore, depending on whether or not the neptunium present in the aqueous phase obtained after operation b1) is removed, depending on the way in which this removal is carried out, depending on the type of reducing agent used during operation c3) and depending on whether or not the uranium(IV) is added during this operation, there is obtained, after step c), an aqueous phase that contains plutonium(III), uranium(VI) and possibly also contains uranium(IV) and/or neptunium(IV) or (V). However, in all cases, the aqueous phase obtained after step c) preferably does not contain more than one μCi of fission products per gram of plutonium so as to meet the NF ISO 13463 standard of June 2000 relating to the manufacture of MOX nuclear fuels for light-water reactors. Moreover, this aqueous phase advantageously has a U/Pu mass ratio ranging from about 20/80 to 50/50. Thus, the function of step c) is twofold: namely to purify the plutonium and uranium that are present in the first aqueous phase obtained after step b) with respect to the fission products, on the one hand, and to allow the uranium/plutonium mass ratio to be adjusted on the other hand. Step d) itself is preferably carried out as described in French Patent Application No. 2 870 841, that is to say: by stabilizing the plutonium, in oxidation state (III), the uranium, in oxidation state (IV), and, where appropriate, the neptunium, in oxidation state (IV), by a singly-charged cation consisting only of atoms chosen from oxygen, carbon, nitrogen and hydrogen atoms, such as the hydrazinium cation; by coprecipitating the thus stabilized plutonium, uranium and, where appropriate, neptunium by oxalic acid or by one of its salts or of its derivatives; and then by calcining the resulting coprecipitate, preferably in an inert or very slightly oxidizing gas, for example a gas containing predominantly argon, in order to remove the carbon and prevent the formation of U3O8. In accordance with the invention, the process advantageously also includes a storage step, which consists in storing either the aqueous phase obtained after operation c3) before step d) is carried out, or the mixed uranium-plutonium oxide obtained after step d). This storage step, which advantageously corresponds to several months of reprocessing spent nuclear fuels by the process of the invention, for example about 4 to 6 months, makes it possible, on the one hand, to ensure that the workshops responsible for reprocessing spent nuclear fuel are decoupled from those responsible for manufacturing fresh nuclear fuel from the mixed uranium-plutonium oxide obtained after this reprocessing, and on the other hand, to adjust the isotopy of the plutonium to that required by the workshops for manufacturing fresh nuclear fuel. By means of the fact that plutonium, uranium and neptunium are more stable in solution in the oxidized state than in the reduced state, but that their coconversion to a mixed oxide requires them to be in the reduced state, the process of the invention provides, in the case of the aqueous phase obtained after step c3) being stored, for this phase to be subjected: between step c) and the storage step, to an oxidation operation in order to reoxidize the plutonium(III) and, where appropriate, the uranium(IV) and/or the neptunium(IV) or (V) to plutonium(IV), to uranium(VI) and to neptunium(VI), respectively, and then to a concentration operation in order to reduce the volume of material stored; and between the storage step and step d): to a reduction operation, for example of the electrolytic type, in order to reduce the plutonium(IV), the uranium(VI), and, where appropriate, the neptunium(VI) to plutonium(III), to uranium(IV) and to neptunium(IV), respectively; or else to a reduction operation, for example by U(IV) or NHA, in which the plutonium(IV), and, where appropriate, the neptunium(VI), are reduced to plutonium(III) and neptunium(IV), respectively, in which case this reduction operation is supplemented with an operation of extracting the uranium(VI), by bringing said aqueous phase into contact with a water-immiscible immiscible solvent phase containing at least one extractant in an organic diluent. In all cases, the process of the invention makes it possible to obtain a mixed uranium-plutonium oxide which, depending on whether or not the neptunium present in the aqueous phase after operation b1) has been removed, contains no neptunium or, on the contrary, also contains neptunium. In either case, this mixed oxide, which is in the form of a powder, can then be used directly for the manufacture of pellets of a mixed nuclear fuel. For this manufacture, this mixed oxide preferably has a U/Pu mass ratio of around 50/50 when it does not contain neptunium and a U/Pu/Np mass ratio of around 49/49/2 when it does contain neptunium. The parameters used during the various operations of the process of the invention such as the volume ratios of the solvent phases to the aqueous phases, the number and the duration of the contacting operations between these phases, the acidity of the aqueous phases, etc., and also the amounts of U(VI) or (IV) that can be added during operations b1) and c3), are therefore adjusted accordingly. As a person skilled in the art will have understood on reading the foregoing text, the extractant for the solvent phases which is used in steps a) and c) and also during the uranium(VI) extraction operation prior to step d), is preferably chosen from extractants that complex the metallic species in oxidation states (IV) and (VI) more strongly than the metallic species in oxidation states (I), (II), (III) and (V), so that uranium(IV), uranium(VI), plutonium(IV), neptunium(IV) and neptunium(VI) are considerably more extractable than plutonium(III) and neptunium(V). This extractant may in particular be a trialkyl phosphate, such as tri-n-butyl phosphate (or TBP), triisobutyl phosphate (TiBP) or a triisoamyl phosphate. The organic diluent for this extractant may itself be chosen from various hydrocarbons proposed for liquid-liquid extractions, such as toluene, xylene, t-butylbenzene, triisopropylbenzene, kerosene and linear or branched dodecanes, such as n-dodecane or hydrogenated tetrapropylene (HPT). However, it is preferred to use, as in the PUREX process, tri-n-butyl phosphate in a dodecane, and to do so in a volume ratio of around 30/70. As mentioned above, the reducing agent capable of reducing plutonium(IV) to plutonium(III), which is used during operations b1) and c3), may especially be uranous nitrate or hydroxylammonium nitrate. Preferably either one is used in conjunction with a nitrous acid scavenger, preferably hydrazine. As regards the reducing agent capable of reducing neptunium(VI) to neptunium(V) without reducing either the uranium or the plutonium, which is used during operation c2) for removing the neptunium, this may especially be a compound of the family of butyraldehydes or hydrazine. In one particularly preferred embodiment mode of the process of the invention, when the uranium and plutonium are coextracted in step a) by means of a solvent phase containing about 30% (v/v) tri-n-butyl phosphate in a dodecane, this step comprises: a first scrubbing operation carried out on the solvent phase obtained after coextracting the uranium and plutonium, in order to remove most of the fission products, and in particular ruthenium and zirconium, which were extracted during this coextraction, by bringing said solvent phase into contact with an aqueous nitric phase containing around 1 to 3 mol/L of HNO3; a second scrubbing operation carried out on the solvent phase in order to remove, from this phase, the technetium which was extracted during the coextraction operation, by bringing said solvent phase into contact with an aqueous nitric phase containing around 3 to 5 mol/L of HNO3; and a complementary operation of coextracting uranium and plutonium from the aqueous phase obtained after the second scrubbing operation, by bringing this phase into contact with a solvent phase containing about 30% (v/v) tri-n-butyl phosphate in a dodecane. Moreover, in this particular preferred embodiment mode: the aqueous nitric phases used during operations b1) and c3) contain around 0.05 to 2 mol/L of HNO3; the aqueous nitric phase used during operation b2) contains around 0 to 0.05 mol/L of HNO3; whereas the aqueous nitric phase used during operation c2) contains around 1 to 3 mol/L of HNO3. If necessary, the process of the invention may also include operations of purifying the uranium present in the second aqueous phase obtained after step b), in order to complete its decontamination from the fission products and/or to separate it from the neptunium liable to have followed it in the aqueous phase during operation b2). These operations may be carried out as in any conventional PUREX process (see, for example, the article BN 3 650 (07-2000) of the treatise “Génie Nucléaire”—“Techniques de l'Ingénieur”). The process according to the invention has many advantages. While being just as effective as the PUREX process in terms of decontamination, unlike the latter, it never allows plutonium to be left without uranium, and thus it minimizes the risk of plutonium being misappropriated for military purposes. It also makes it possible to obtain a mixed uranium-plutonium oxide powder that can be used directly for the manufacture of MOX nuclear fuels for fast neutron reactors or light-water reactors of the second or third generation. Moreover, it is equally applicable to the reprocessing of a spent uranium oxide nuclear fuel as to the reprocessing of a spent mixed uranium-plutonium oxide nuclear fuel. Other advantages and features of the process of the invention will become apparent upon reading the rest of the description that follows, which refers to examples of processes for the industrial-scale implementation of this process. Of course, these examples are given merely to illustrate the invention and in no way constitute a limitation thereof. In the embodiments modes shown in FIGS. 1 to 5, all the extraction, back-extraction and re-extraction operations are carried out in multi-staged units of the mixer/settler, pulsed column or centrifuge-extractor type. The directions of flow of the solvent phase entering or leaving these units are shown symbolically by a continuous double-line arrow, while the directions of flow of the aqueous phase entering or leaving the said units are shown symbolically by a continuous single-line arrow. The description firstly refers to FIG. 1, which shows a block diagram of a first embodiment mode of the process of the invention, designed to obtain a mixed uranium-plutonium oxide powder containing no neptunium and able to be used directly in the manufacture of an MOX nuclear fuel, from a dissolution liquor of a spent UO2 nuclear fuel that has been conventionally prepared, that is to say by dissolving this fuel in nitric acid and clarifying the resulting mixture. Such a dissolution liquor typically contains 200 to 300 g/L of uranium per 2 to 3 g/L of plutonium, i.e. a U/Pu ratio of about 100/1, and has a content of fission products of around 50 to 70 Ci per gram of plutonium. As mentioned previously, the process of the invention firstly comprises a step designed to separate the uranium and the plutonium from the fission products, the americium and the curium. As may be seen in FIG. 1, this separation step comprises: an operation, labelled “U/Pu coextraction”, which consists in extracting both the uranium and the plutonium, the first in oxidation state (VI), the second in oxidation state (IV), from the dissolution liquor by bringing this liquor into contact with a solvent phase containing about 30% (v/v) tri-n-butyl phosphate (TBP) in an organic diluent, for example, a dodecane; an operation, labelled “FP scrubbing”, which consists in removing from the solvent phase the fission products, particularly ruthenium and zirconium, that have been extracted during the “U/Pu coextraction” by bringing the solvent phase resulting from this coextraction into contact with an aqueous nitric phase of moderate acidity, for example a 1 to 3M nitric acid solution; an operation, labelled “Tc scrubbing”, which consists in removing from the solvent phase the technetium that has been extracted during the “U/Pu coextraction”, by bringing the solvent phase resulting from the “FP scrubbing” into contact with an aqueous nitric phase of moderate acidity, but higher than that of the aqueous nitric phase used for the “FP scrubbing”, for example a 3M to 5M nitric acid solution; and an operation, labelled “U/Pu complementary coextraction”, which consists in recovering the U(VI) and Pu(IV) fractions that have followed the technetium in the aqueous phase during the “Tc scrubbing”, by bringing this aqueous phase into contact with a solvent phase, again consisting of about 30% (v/v) TBP in a dodecane. Four phases are thus obtained: the two aqueous phases (or raffinates) resulting from the “U/Pu coextraction” and from the “U/Pu complementary coextraction”, which are laden with fission products and, in the case of the first of said aqueous phases, with americium and with curium, and which are removed from the process; the solvent phase resulting from the “U/Pu complementary coextraction” which is sent to the unit where the “U/Pu coextraction” takes place, to be added to the solvent phase flowing through this unit; and the solvent phase resulting from the “Tc scrubbing” which is laden with U(VI), with Pu(IV) but also with neptunium(VI) (since most of the neptunium present in the dissolution liquor is extracted by the TBP) and is sent to a zone where the uranium/plutonium partition step takes place, in which two aqueous phases are formed, the first containing plutonium and uranium, the second phase containing uranium but no plutonium. This partition step comprises: an operation, labelled “Pu/U back-extraction”, which consists in back-extracting, from the solvent phase resulting from the “Tc scrubbing”, the plutonium(IV) and a fraction of the uranium(VI) that are present in this phase, by bringing said solvent phase into contact with an aqueous phase of low acidity, for example a 0.05 to 2M nitric acid solution, containing a reducing agent that reduces the Pu(IV) to Pu(III) and neptunium(VI) to neptunium(IV), respectively, and to do so without reducing the uranium, and also a nitrous acid scavenger suitable for destroying the nitrous acid that tends to form in the aqueous phase, and thus to stabilize the reducing agent and the Pu(III). This reducing agent is, for example, uranous nitrate (or U(IV)), whereas the nitrous acid scavenger is, for example, hydrazine (or NH); an operation, labelled “Pu barrage”, which consists in completing the back-extraction of the plutonium(IV) by bringing the solvent phase resulting from the “Pu/U back-extraction” into contact with an aqueous nitric phase of low acidity, for example a 0.05 to 2M nitric acid solution, containing the same reducing agent and the same nitrous acid scavenger as those used for the “Pu/U back-extraction”; and an operation, labelled “U back-extraction” which consists in back-extracting the uranium(VI) from the solvent phase resulting from the “Pu barrage”, by bringing this solvent phase into contact with a very dilute aqueous nitric phase, for example a 0 to 0.05M nitric acid solution. Since neptunium(Iv) is less extractable by TBP than neptunium(VI), it is partially back-extracted during the “Pu/U back-extraction”. The aqueous phase resulting from this operation therefore contains neptunium in addition to plutonium and uranium. The partition step therefore also includes an operation, labelled “Np scrubbing”, which consists in back-extracting the neptunium(IV) present in the aqueous phase resulting from the “Pu/U back-extraction”, by bringing this phase into contact with a solvent phase consisting of about 30% (v/v) TBP in a dodecane, in order to remove from this aqueous phase the neptunium fraction that it contains. Moreover, since uranium(VI) and neptunium(IV) behave in a relatively similar manner, a fraction of the uranium(VI) present in the aqueous phase resulting from the “U/Pu back-extraction” is re-extracted with the neptunium, which fraction may be relatively large depending on the parameters used to carry out the “Np scrubbing”. Thus, and as may be seen in FIG. 1, it is possible, according to the invention, to add uranium to the aqueous phase subjected to the “Np scrubbing”, just before said phase leaves the unit where this operation takes place, in order to recharge this aqueous phase with uranium should this be deemed necessary. This uranium, which may without distinction be uranium(VI) or uranium(IV), may be added in the form of an aqueous nitric solution, it being understood that, if it is uranium(IV), the latter is then stabilized by a nitrous acid scavenger of the hydrazine type. The aqueous phase resulting from the “Np scrubbing” is then subjected to an oxidation operation for bringing the Pu(III) back to oxidation state (IV) and, where appropriate, the U(IV) to oxidation state (VI), before the step of purifying the plutonium and the uranium that it contains is carried out. This oxidation operation may especially be carried out in a conventional manner, that is to say by making said aqueous phase flow, after possibly being diluted with an aqueous nitric phase of high acidity, for example a 12M nitric acid solution, in a stream of nitrogen oxides NOX so as to destroy the nitrous acid scavenger that it contains, thereby making it possible for the nitrous acid to reform and reoxidize the Pu(III) to Pu(IV), the excess nitrous acid then being removed by decomposition to NO and NO2 and venting of the nitrogen oxides thus formed. The step of purifying the plutonium and the uranium which follows the partition step and is for the purpose of completing the decontamination of these two elements from the fission products, i.e. so as to obtain in practice an aqueous plutonium-uranium stream preferably having a content of fission products of at most 1 μCi per gram of plutonium, comprises: an operation, labelled “Pu/U coextraction”, which consists in jointly extracting the plutonium(IV) and the uranium(VI) from the aqueous phase resulting from the oxidation operation, by bringing this phase into contact with a solvent phase consisting of about 30% (v/v) TBP in a dodecane, as previously; an operation, labelled “FP scrubbing”, which consists in removing from the solvent phase resulting from the “Pu/U coextraction” the fission products that have been extracted during this coextraction, by bringing this solvent phase into contact with an aqueous nitric phase of moderate acidity, for example a 1 to 3M nitric acid solution; and an operation, labelled “U/Pu back-extraction”, which consists in back-extracting from the solvent phase resulting from the “FP scrubbing” the plutonium(IV) and a fraction of the uranium(VI) that are present in this phase, by bringing the latter into contact with an aqueous nitric phase of low acidity, for example a 0.05 to 2M nitric acid solution, containing a reducing agent capable of reducing the Pu(IV) to Pu(III), without touching the uranium, for example, hydroxylammonium nitrate (or HAN), stabilized by a nitrous acid scavenger of the hydrazine type. As may be seen in FIG. 1, it is possible, here again, to add uranium to the aqueous phase subjected to the “U/Pu back-extraction” just before the aqueous phase leaves the unit where this operation takes place, in order to recharge this aqueous phase with uranium should this be deemed necessary. As previously, the uranium thus added may be uranium(VI) or uranium(IV) as an aqueous nitric solution containing, in addition, a nitrous acid scavenger if the uranium is uranium(IV). The raffinate resulting from the “Pu/U coextraction” is removed from the process. The solvent phase resulting from the “Pu/U back-extraction”, which contains uranium but no longer contains plutonium, rejoins the solvent phase resulting from the “Np scrubbing”. The aqueous phase resulting from the “Pu/U back-extraction”, which is laden with purified plutonium(III) and uranium(IV) or (VI) is itself sent to a unit where, after an oxidation operation for reoxidizing the Pu(III) to Pu(IV) and which is preferably carried out in the same way as the oxidation operation following the “Np scrubbing”, it is subjected to the concentration step in order to increase its plutonium content and its uranium content. The aqueous phase thus concentrated is then stored, for example in tanks with a system of tubes, for a period advantageously corresponding to several months of implementation of the reprocessing process, for example 4 to 6 months, so as to have a stock of purified plutonium and uranium sufficient so that the workshops responsible for manufacturing MOX nuclear fuel are able to work independently of the workshops responsible for reprocessing the spent fuel. This storage also makes it possible to adjust the isotopy of the plutonium to that required for the workshops for manufacturing MOX nuclear fuel. After the storage, the process of the invention further includes: an operation to reduce the plutonium(IV) present in the concentrated aqueous phase to Pu(III) by the addition of an aqueous nitric solution containing U(IV) stabilized by a nitrous acid scavenger; and an operation, labelled “U scrubbing”, which consists in removing from the concentrated aqueous phase the uranium(VI) that it contains by bringing this aqueous phase into contact with a solvent phase, again consisting of approximately 30% (v/v) TBP in a dodecane. The solvent phase resulting from the “U scrubbing” rejoins the solvent phase resulting from the “Pu/U back-extraction” of the purification step and, with it, the solvent phase resulting from the “Np scrubbing”. The aqueous phase resulting from the “U scrubbing” is itself directed to a unit where, after a possible adjustment of its U(IV) content suitable for in this phase a U(Pu) mass ratio consistent with that which the mixed uranium-plutonium oxide that it is desired to prepare must have, the step of coconverting the plutonium and uranium to a mixed oxide is carried out. As previously mentioned, this coconversion step is preferably carried out according to the process described in FR-A-2 870 841, that is to say by coprecipitation, by oxalic acid or one of its salts or one of its derivatives, of uranium(IV) and plutonium(III) that had been prestabilized by a singly charged cation consisting only of atoms chosen from oxygen, carbon, nitrogen and hydrogen atoms, such as the hydrazinium cation, or by a compound such as a salt, capable of forming such a cation, followed by calcination of the resulting coprecipitate, preferably in an inert or very slightly oxidizing gas, for example a gas containing predominantly argon. The mixed uranium-plutonium oxide powder thus obtained can then be used to manufacture MOX nuclear fuel pellets, for example by a MIMAS process, in which case this powder is screened, mixed with uranium oxide and possibly with scrap from the manufacture of pellets in the form of chamotte, and then the resulting mixture undergoes a pelletizing operation followed by a sintering operation. As indicated above, it is preferred to use, for the manufacture of MOX nuclear fuels, a mixed oxide powder having a U/Pu mass ratio of approximately 50/50. Consequently, the amount of uranium that is introduced into the unit where the “Np scrubbing” takes place is preferably such that the aqueous phase resulting from this operation has a U/Pu mass ratio of around 20/80 to 50/50, and the parameters used to carry out the “Pu/U back-extraction” located downstream of this “Np scrubbing” are preferably adjusted so as to obtain, after this back-extraction, an aqueous phase having a U/Pu mass ratio of around 20/80 to 50/50 and, ideally, of around 20/80 to 30/70. It is desirable to obtain such a ratio in order to minimize the volume of material stored in tanks. In any case, by adjusting the U(IV) content of the aqueous phase resulting from the “U scrubbing” it is possible, should it be necessary, to give the aqueous phase subjected to the co-conversion operation a U/Pu ratio equal or approximately equal to 50/50. In the embodiment mode of the process of the invention illustrated in FIG. 1, this includes, in addition, operations of purifying the uranium present in the aqueous phase resulting from the “U back-extraction”, which operations are intended to complete its decontamination from the fission products and most particularly to separate it from the neptunium fraction that had followed it during the “Pu/U back-extraction” and the “U back-extraction” of the partition step. These purification operations may be carried out as in any conventional PUREX process and have consequently not been shown in FIG. 1 for the sake of simplifying this figure, nor have they been shown in the following figures. The process of the invention may also include ancillary operations, in particular operations of scrubbing, with pure diluent, the aqueous phases intended to be sent to the vitrification unit and operations of scrubbing and regenerating the spent solvent phases. Here again, these operations, which are well known in the prior art, have not been shown in FIGS. 1 to 5 for the sake of simplifying these figures. A block diagram of a first variant of the embodiment mode illustrated in FIG. 1 will now be described with reference to FIG. 2, in which: the partition step does not include “Np scrubbing”; and the neptunium present in the aqueous phase resulting from the “Pu/U back-extraction” of this step is removed during the purification step. To do this, after a “Pu/U coextraction” operation similar to that of the embodiment mode described above, the “FP scrubbing” of the purification step is carried out by bringing the solvent phase resulting from this coextraction into contact with an aqueous phase of moderate acidity, for example a 1 to 3M aqueous nitric acid solution, to which has been added a reducing agent capable of reducing neptunium(VI) which is extractable by TBP, to neptunium(V), which is not extractable by TBP, and to do so without reducing either the plutonium or the uranium. This reducing agent is, for example, a butyraldehyde (ButAl). The neptunium thus passes into the aqueous phase, whereas the plutonium and the uranium remain in the solvent phase. The “Pu/U back-extraction” is then carried out in the mode of implementation described above, but by suitably adjusting the parameters of this operation so as to obtain a U/Pu mass ratio of around 20/80 to 50/50 and, ideally, of around 20/80 to 30/70 in the aqueous phase, bearing in mind that the U/Pu mass ratio of the solvent phase resulting from the “FP scrubbing” is likely to be almost the reverse because of the absence of “Np scrubbing”. FIG. 3 shows a block diagram of a second variant of the embodiment mode illustrated in FIG. 1. This second variant differs from the variant that has just been described in that the reducing agent present in the aqueous phase used during the “Pu/U back-extraction” and “Pu barrage” operations of the partition step is an agent that is capable of reducing plutonium(IV) to plutonium(III) and neptunium(VI) to neptunium(V), respectively, this being the case, for example, of hydroxylammonium nitrate. Since neptunium(V) is not extractable by TBP, it is therefore completely back-extracted during the “Pu/U back-extraction” and the “Pu barrage” and what are obtained after the partition step are two aqueous phases, one of which, resulting from the “Pu/U back-extraction”contains plutonium, uranium, and neptunium, whereas the other, resulting from the “U back-extraction” contains uranium but does not contain either plutonium or neptunium. The removal of the neptunium present in the aqueous phase resulting from the “U/Pu back-extraction” of the partition step is then carried out during the “FP scrubbing” of the purification step, in exactly the same way as in the first variant. This variant makes it possible, in the case of reprocessing spent nuclear fuels that have cooled for about 10 years or more, to dispense with the operations intended to purify the uranium. FIG. 4 shows schematically yet another variant of the embodiment mode of the process of the invention illustrated in FIG. 1, which differs from this embodiment mode in that the aqueous phase resulting from the “Pu/U back-extraction” of the purification step is sent directly to a unit where the step of coconverting the plutonium and uranium to a mixed oxide is carried out. In this variant, the “Pu/U back-extraction” operation of the purification step therefore necessarily includes the addition of a suitable amount of uranium(IV) so as to give the aqueous phase resulting from this operation a U/Pu mass ratio consistent with that which the mixed uranium-plutonium oxide that it is desired to manufacture must have. Moreover, the decoupling between the workshops responsible for manufacturing MOX nuclear fuel and the workshops responsible for reprocessing spent nuclear fuel is ensured by storing the mixed uranium-plutonium oxide powder obtained after the coconversion step. FIG. 5 illustrates schematically a second embodiment mode of the process of the invention which, unlike the previous ones, is designed to obtain a mixed uranium-plutonium oxide powder, which also contains neptunium, from a dissolution liquor of a spent UO2 nuclear fuel. In this embodiment mode, the process takes place as in the mode of implementation illustrated in FIG. 3 except that it does not include the elimination of the neptunium during the “FP scrubbing” of the purification step. The neptunium therefore accompanies the plutonium with which it was back-extracted during the “Pu/U back-extraction” throughout all the subsequent steps of the process until a mixed uranium-plutonium-neptunium oxide is obtained. To give an example, a simulation was carried out using the PAREX Code of the Commissariat à L'Energie Atomique for the first embodiment mode of the process of the invention illustrated in FIG. 1. The data of this simulation were the following: Dissolution liquor: [U]=250 g/L [Pu]=2.55 g/L 4.5M HNO3 Feed flow rate in the process=637 L/h; Solvent phases: 30% (v/v) TBP in TPH; Step of separating the uranium and plutonium from the fission products, americium and curium: Solvent phase entering the “U/Pu coextraction” unit: flow rate=1272 L/h Aqueous phase entering the “FP scrubbing” unit: 2M HNO3, flow rate=273 L/h Aqueous phase entering the “Tc scrubbing” unit: 1.5M HNO3, flow rate=38 L/h; then 12M HNO3, flow rate=200 L/h Solvent phase entering the “U/Pu complementary coextraction” unit: flow rate=545 L/h; Partition step, separating uranium and plutonium into two aqueous phases: Aqueous phase entering the “Pu barrage” unit: 0.2M HNO3+0.15M hydrazine, flow rate=236 L/h; then 0.2M HNO3+150 g/L U(IV), flow rate=9.4 L/h Addition of U(IV) to the “Pu/U back-extraction” unit: 150 g/L, flow rate=21.7 L/h Solvent phase entering the “Np scrubbing” unit: flow rate=215 L/h Aqueous phase leaving the “Np scrubbing” unit: [Pu]=6 g/L [U]=1.83 g/L Flow rate=272.5 L/h; Oxidation operation: Aqueous phase used to dilute the aqueous phase resulting from the “Np scrubbing”: 12M HNO3, flow rate=110 L/h Aqueous phase resulting from the oxidation operation: [Pu]=4.2 g/L [U]=1.3 g/L Flow rate=385.4 L/h; Plutonium/uranium purification step: Solvent phase entering the “Pu/U coextraction” unit: flow rate=95 L/h Aqueous phase entering the “FP scrubbing” unit: 1.5M HNO3, flow rate=24 L/h; then 12M HNO3, flow rate=6 L/h Aqueous phase entering the “Pu/U back-extraction” unit: 0.2M HNO3+0.2M hydrazine+0.4M hydroxylammonium nitrate, flow rate=47 L/h Addition of U(IV) to the “Pu/U back-extraction” unit: 150 g/L, flow rate=11 L/h Aqueous phase leaving the “Pu/U back-extraction” unit: [Pu]=27.5 g/L [U]=6.7 g/L Flow rate=59.1 L/h; Concentration step: Aqueous phase resulting from the concentration operation: [Pu]=200 g/L [U]=50 g/L; Reduction operation: Addition of U(IV) to the reduction unit: 200 g/L, flow rate=8.1 L/h Aqueous phase resulting from the reduction operation: [Pu]=100 g/L [U]=125 g/L; and Adjustment operation Aqueous phase adjustment: HNO3+200 g/L U(IV). What is thus obtained is an aqueous phase containing 50 g/L of Pu, in oxidation state (III), and 50 g/L of U, in oxidation state (IV), which have a purity level of greater than 99% and are suitable for being coconverted to a mixed uranium-plutonium oxide. US-A-4 278 559 FR-A-2 870 841 Article BN 3650 (07-2000): The treatise “Génie Nucléaire”—“Techniques de l'Ingénieur”. |
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048287822 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Referring first to FIG. 1, a typical BWR fuel assembly is shown, designated generally by numeral 1. The essential components of a BWR fuel assembly are: an array of fuel rods 2, an upper tie plate (not shown), a lower tie plate 4 and a fuel channel slip 6, and a number of fuel rod spacers 8. The fuel rods 2 are supported in a square array by the lower and upper tie plates. The fuel rod array ("the fuel bundle") typically comprises either a 7.times.7, 8.times.8, or a 9.times.9 square. For purposes of example, the fuel bundle will hereafter be assumed to comprise an 8.times.8 array. The fuel channel slip 6 fits around the fuel bundle to form the fuel assembly. The entire assembly typically weighs about 700 pounds and has a 5.33 inch outside dimension. The upper tie plate is a stainless steel, machined gridwork casting. The casting maintains the regular arrangement of fuel rods within a fuel rod assembly. The casting has welded to it a lifting bail 10 used for movement of the assembly. The lower tie plate 4 is also a stainless steel casting that provides grid holes for the fuel rod end plugs. Coolant flow is directed through the holes in the nosepiece into the lower tie plate grid 4, which distributes the flow to the fuel bundle. The fuel rod spacers 8 maintain even lateral spacing of the fuel rods 2, and suppresses fuel rod vibration. Each spacer 8 is a lattice with finger springs that press laterally against the walls of the fuel rods. Referring now to FIG. 2, a bottom view of the BWR fuel assembly is shown. A nosepiece 12 with an inverted tripod 14 extends downward from the fuel assembly 1. Looking up into the fuel assembly, the lower tie plate 4 can be seen, comprising a grid having a plurality of first apertures 16 for housing the end plugs 18 of each fuel rod 2, and a plurality of slightly larger second apertures 20 which allow coolant water to flow between the fuel rods during operation. In the exemplary 8.times.8 fuel assembly, there are 64 first apertures 16 and 49 second apertures 20. In accordance with the present invention, an ultrasonic probe is inserted up through the tripod 14 and successively inserted into any of the second apertures 20. While the probe could be inserted into all 49 second apertures it is actually only necessary to access 16 of the second apertures (depicted in FIG. 3 with small circles and identified by reference numeral 20A) since the probe can be rotated in each aperture 20A to examine each of the fuel rods in the surrounding first apertures 16. As is evident from FIG. 2, many of the apertures are obstructed by the nosepiece cover and the presence of the inverted tripod. Accordingly, in the present invention, a flexible probe passes through a single pivot point in or slightly below the tripod to facilitate access to each of the apertures 20. Referring now to FIG. 3, a simplified representation of the lower tie plate 4 is portrayed, showing the location of each of the 64 fuel rods in the exemplary 8.times.8 array of the BWR fuel assembly. As can be seen from FIG. 3, only 16 of the fuel rods are substantially visible through the nosepiece from a straight bottom view of the fuel assembly. The inverted tripod 14 forms three tridents I, II and III with respect to the exposed fuel rods. As is evident from FIG. 3, tridents I and II are symmetrical with respect to the fuel rods, while trident III is non-symmetric. The present inventor has found that if the probe passes through a single pivot point 22 in or slightly below trident III, the non-symmetric trident, each of the fuel rods can be readily accessed by passing the probe into any of the 49 second apertures 20, or preferably only the 16 apertures 20A (if the probe is rotated). Moreover, the present inventor has found that if this single pivot point 22 is located substantially centrally in or slightly below trident III, the access to each of the apertures 20 is made most accessible. Multiple single pivot points, one in each trident, can also be provided to allow multiple inspections to occur in parallel. In order to rotate the probe about the pivot point, many different embodiments of the present invention are possible, three of which are described herein. In a first embodiment, shown in FIG. 4, a fixed ball joint 24 is located at the pivot point 22. The ball joint 24 has an aperture 26 through which the probe 28 passes, and the probe is pivoted at its lower end by means of an x-y scanning bridge (not shown in FIG. 4). Alternatively, in a second possible embodiment of the invention, the probe is seated at its lower end in a two-stage goniometric cradle which, when rotated, pivots the probe about the single pivot point. In a still further embodiment of the invention, the probe is seated in a goniometric cradle disposed on a rotational table, which, when rotated, causes the probe to pivot about the single pivot point. In either of these last two embodiments, the center of rotation of the cradle must be located external to the body of the cradle, preferably about 9 inches up. In order to access the apertures 20 from a single pivot point, the probe 28 of the present invention has a flexible midsection 30 which allows it to bend. The probe is provided with a bullet-shaped nose 32 to guide the probe into each opening, and to protect the probe as it contacts the edge of each aperture 20. In the preferred embodiment, a Krautkrammer transducer serves as the active UT portion of the probe 28 in the preferred embodiment, although any appropriate commercially available ultrasonic device may be used. Probe 28 may utilize one transducer which both sends and receives the ultrasonic signal, or separate transmit and receive transducers may be provided. The probe operates by sending out ultrasonic signals which vibrate the outer shell of the fuel rod; the presence of water inside the fuel rod will affect the vibration of the outer shell of the fuel rod, but more significantly, will dampen the amplitude of the reflected ultrasonic signal. To test the fuel rods, the probe is rotated until a maximum signal is received. In the preferred embodiment of the invention, the return ultrasonic signal is integrated, and a threshold detector is used to determine whether the signal is of appropriate strength, indicating no water in the particular fuel rod. The amplitude of the return signal must be greater than a prescribed threshold value for a fuel rod to pass. The probe is then rotated 90.degree. to test the next fuel rod, and so on around the entire 360.degree.. Alternately, the probe may be rotated continuously through the entire 360.degree., and the rods' return signal captured on the fly. Next, the probe is brought down beneath the aperture 20, pivoted about the single pivot point, and reinserted into the next aperture, and so on until all the fuel rods of the assembly have been tested. The novel arrangement of the present invention allows each fuel rod to be examined individually and therefore isolates a leakage problem to a particular fuel rod. The arrangement of the present invention is also extremely accurate, requiring a reinspection rate of only 0.05%, as compared to a typical reinspection rate of 1.5% for the previously-described vacuum sipping techniques. Although the present invention has been described in connection with a plurality of preferred embodiments thereof, many other variations and modifications will now become apparent to those skilled in the art. For example, the novel positioning design of the invention can be used to position other tools in the nosepiece, at the bottom of the fuel rods, or in the flow paths between the fuel rods to perform other inspection or maintenance tasks (e.g., visual inspection, machining and polishing operations, welding operation, debris retrieval tasks, rod to rod gap measurements, crud sampling and removal, fuel rod dimensioning, eddy current inspection transducers, Electro Magnetic Acoustic Transducer inspection, rod reactivity measurements, serial number verification, inspection of lower fuel spacers, rod vibration, oxide measurements, flow channel blockage analysis and temperature measurements). It is preferred, therefore, that the present invention be limited not by the specific disclosure herein, but only by the appended claims. |
041349415 | abstract | There are provided pressed spherical fuel elements for high temperature reactors made of a graphite matrix with separate embedded coated fuel and fertile material particles wherein the fuel elements comprise 3 concentric layers including a graphite nucleus or core (1) which only contains fertile material particles (4), this graphite nucleus (1) is surrounded by a graphite zone (2) which only contains the fuel particles (5) and this is encased in a shell (3) of pure graphite, the same graphite material being present in all three layers. |
052873919 | summary | BACKGROUND OF THE INVENTION This invention relates to a metering system, and more particularly, for a metering system for controlling the feed of portions of nuclear fuel assemblies. After reprocessing in a chemical plant where the nuclear material (e.g. UO.sub.2) in the nuclear fuel assemblies is dissolved in a solvent stream, residues of the assemblies need to be encapsulated in a cementitious matrix in standard storage containers or drums. These residues chiefly comprise appendages of the nuclear fuel assemblies and hulls which are small sheared tubular portions of cladding tubes which contained the nuclear material. In order to provide a high utilisation of the space within the drum, it is advisable to regulate the ratio of appendages to hulls fed to the drum. SUMMARY OF THE INVENTION According to the present invention there is provided a system for metering the feed of portions of nuclear fuel assemblies, the system comprising a declivitous screen for receiving the portions and for separating the portions into larger and smaller components thereof, a first receptacle for collecting the larger components falling from the screen and for discharging the larger components into a collecting receptacle, the collecting receptacle being controllable so as to discharge the larger components therein on to means for directing the larger components into a container, means for sensing the level of the portions in the container, a buffer chamber for receiving the smaller components passing through the screen, a discharge port of the buffer chamber for discharging the smaller components from the buffer chamber into an oscillatable trough adapted to oscillate around the longitudinal axis thereof to discharge a metered quantity of the smaller components therefrom, and means for receiving the smaller components discharged from the trough and for feeding said smaller components to the container. The invention will now be further described by way of example only with reference to the accompanying drawings, in which: |
059845791 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention generally relates to synthetic soil and, more particularly to a blend of materials which can be used as a plant-growth medium, as a cover material for solid waste landfills or disturbed soil areas, and as a medium for treating contaminated water. 2. Background Art Currently, solid waste disposal sites are typically covered with clay or another impermeable membrane such as strong plastic to keep rain water from infiltrating the waste and leaching out into the ground water. Such coverings are very expensive to install and maintain. The desirability of using synthetic soil materials in covering closed landfills and reclamation of disturbed soil areas such as mined, filled, and regraded land has recently been recognized. An article in the Post-Standard newspaper on Apr. 19, 1991 reported on a student proposal "to use processed bottom ash from an incinerator and some form of compost in a 1-to-30 ratio" (subsequently corrected to 1-to-3 ratio) as a waste bed cover. Compost, for a long time has been used as a soil enriching material. Compost contains the nutrients used in relatively large quantities for plant growth. The earlier proposed waste bed cover accordingly employed compost as the primary ingredient. Landfills and the disturbed soil areas often cover many acres and thus would require vast quantities of compost to implement the earlier proposal. Unfortunately, compost is not as readily available as municipal solid waste incinerator ash which today is produced in many locations and is typically disposed of in landfills. Accordingly, a need still persists for a readily available and economically feasible alternative to the use of impermeable membranes in covering and vegetating solid waste landfills. SUMMARY OF THE INVENTION This need is satisfied and other benefits realized, in accordance with the principles of the present invention, by the provision of a residual biotech soil composition comprising a blend or mixture of one to nine parts of incinerator ash and one part of compostable organic material or compost, on a dry weight basis. Preferably, the blend contains a majority of incinerator ash, and most preferably, two to six parts of incinerator ash for every part of organic material. The incinerator ash may be unprocessed or processed and preferably is derived from incineration of municipal solid waste at a solid waste combustor or other waste-to-energy facility. The incinerator ash may include bottom ash and an admixture of fly ash. The compost is normally derived from composting organic solid waste. Combining incinerator ash and compost into such a synthetic soil mix to use as a barrier-protection layer and topsoil layer enclosing a landfill helps reduce waste and promotes a beneficial use of these two materials for supporting vegetative growth, provision of landfill covers and protection of disturbed soil areas. The mixture is also suitable as a medium for treating biologically and/or chemically contaminated water. |
claims | 1. A fusion energy device comprising:(a) a vacuum tank having an interior tank surface, an interior vacuum pressure, an exterior tank surface and an outside tank diameter, said vacuum pressure being generally predetermined from a range of 1 micro Torr to 0.1 Torr, and said outside tank diameter being generally predetermined from a range of 0.5 meters to 100 meters;(b) a plurality of magnetic coils disposed to lie inside said interior tank surface, said magnetic coils having cross-sectional outlines selected from the group consisting of circle, ellipse, and polygon, said magnetic coils having plan-view outlines selected from the group consisting of circle, ellipse, and polygon, said magnetic coils having open bores passing respectively through said plan-view outlines, said plan-view outlines disposed to lie respectively parallel to the faces of a polyhedron, said polyhedron having a polyhedral center and plural surrounding polygonal faces, said polygonal faces having polygonal centers, said polygonal centers having a plurality of surrounding polygonal edges, said magnetic coils carrying electrical currents such that a plurality of magnetic field vectors are disposed to lie respectively at said polygonal centers, said magnetic field vectors having polarities and magnitudes, said polarities being oriented to point from said polygonal centers towards said polyhedral center, said magnitudes being generally predetermined from a range of 100 gauss to 100 kilogauss;(c) a plurality of airtight containers respectively surrounding said magnetic coils, said airtight containers being approximately conformal to said cross-sectional outlines and approximately conformal to said plan-view outlines, said airtight containers being disposed to lie respectively on said polygonal faces, said airtight containers being spaced apart from said surrounding edges of said polygonal faces by spacing distances, said spacing distances being generally predetermined from a range of 1 millimeter to 1 meter;(d) a plurality of hollow airtight legs, said legs having interiors, said legs having ends and opposite ends, said ends sealed to said interior tank surface, said opposite ends sealed to said metal containers, said legs urging said plurality of airtight containers respectively onto said plurality of polygonal-faces, said urging being against the force of gravity and against the force of magnetic pressure;(e) at least one high-voltage-power-supply positioned outside said exterior tank surface and electrically connected to said airtight containers by a plurality of electrical wires, said wires passing through said interiors of said legs, said wires carrying voltages generally predetermined from a range of 1 kilovolt to 1 megavolt;(f) one or more electron emitters, each of said electron emitters disposed in position at a predetermined emitter-inset-distance from said interior tank surface and disposed in orientation so as to emit electrons in a direction generally toward said polyhedron center, said emitter-inset-distance being generally in the range 1 millimeter to 1 meter;(g) a plurality of recirculating electron beams, said beams being respectively disposed inside said open bores of said plurality of coils, said electron beams being comprised of electrons having energies generally in a range of 1 kilo-electron-volt to 1 mega-electron-volt;(h) one or more gas-cells, each of said cells having a cell-inside and a cell-outside, each of said cells sealed and bounded by a pair of aperture plates disposed to lie perpendicular to a selected one of said electron beams and disposed to lie inside a corresponding one of said open bores, said pair of aperture plates being separated from each other by a distance predetermined and selected from the range 1 millimeter to 1 meter, said pair of aperture plates penetrated respectively by a pair of holes of predetermined diameters from the range 1 millimeter to 1 meter, said pair of holes disposed so that said electron beam passes freely through said pair of holes and also through said cell-inside, each of said cells containing a fuel gas, said gas having a gas-pressure, said gas-pressure being regulated by gas supplied through a gas delivery tube, said tube being connected from said cell-inside to said exterior tank surface;(i) one or more vacuum pumps located outside said exterior tank surface and connected to said vacuum tank respectively though one or more pumping ports, so that said interior vacuum pressure is smaller than said gas-pressure, the ratio of said gas-pressure to said vacuum-pressure being generally in the range 2 to 10,000 and;(j) one or more concentrically arranged pumped volumes surrounding at least one of said gas-cells, said pumped volumes being of progressively increasing size from an innermost volume to an outermost volume and having progressively decreasing internal gas pressures from said innermost volume to said outermost volume, and being respectively connected to a plurality of auxiliary vacuum pumps, and being respectively sealed and bounded by a plurality of pairs of auxiliary aperture plates aligned so as to pass said selected one of said electron beams through said one of said gas-cells;whereby at least one internal fuel supply is provided such that a fusion reactor is fueled by ions created by electron beams' ionizing gas confined inside one or more gas-cells by multiple pairs of aperture plates containing small holes to let the electron beams pass freely through the gas while hindering the flow of neutral gas out through the holes. 2. The fusion energy device of claim 1 further including:one or more electron-extractor electrodes, each of said electrodes including a thin plate of material chosen from the group metal and ceramic, said plate having a thickness generally in the range 1 mm to 1 cm, said late having a diameter generally, in the range 1 cm to 100 cm, said plate being rigidly mounted inside a selected one of said open bores, said plate having an inner face and an outer face, said inner face being disposed to lie facing said polyhedral center, said outer face being connected electrically and mechanically to said interior tank surface;whereby a means is provided for selectively removing up-scattered electrons, thereby reducing the overall power lost to electrons, thereby increasing the power balance, and thereby reducing the projected break-even reactor size. |
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047175311 | abstract | A fuel transfer system for a fuel container (26) within a nuclear reactor facility includes a transport car (10) for transporting the fuel container (26) through a transfer tube between a reactor containment handling pool and a spent fuel storage pool. The system includes upending mechanisms for automatically pivoting the fuel container (26) from its horizontal transport mode to its vertical, fuel loading-unloading mode when the fuel container (26) enters one of the pools in response to the translational drive of the transport car (10). The upending mechanisms include slotted brackets (46, 146) mounted upon the fuel container (26), and pivotable pick-up bars (44, 144) for engaging the brackets (46, 146) of the fuel container (26). As the transport car (10) translationally moves past the pick-up bars (44, 144), the brackets (46, 146) of the fuel container (26) engage the bars (44, 144) whereby the latter pivot so as to in turn cause pivotal movement of the fuel container (26) upon the transport car ( 10) through means of trunnions (28). Reverse translational movement of the transport car (10) causes reverse pivotal movement of the container (26) from the vertical to the horizontal mode and ultimate disengagement of the brackets (46, 146) from the pick-up bars (44, 144). |
description | Although the present invention is sometimes described herein in the context of performing diagnostic tests on an outboard engine, the present invention can be utilized in connection with diagnostic tests for many different types of engines, including automobile engines. For example, and although the present invention is described herein in connection with an engine that includes fuel injection, the invention can be utilized in connection with engines that include any one of a wide variety of fuel injection mechanisms, electrically powered engines and in engines that include other fuel or energy supply systems. In addition, the methods and apparatus for enabling operator selection of a language as described herein can be utilized in many different applications in addition to diagnostic testing. The present invention is not limited to use in connection with only outboard engine diagnostic test operations. FIG. 1 is a block diagram of an engine 10 and an external computer 12 coupled to engine 10. Engine 10 is, in one embodiment, an internal combustion engine including fuel injection, such as an Evinrude outboard engine commercially available from Bombardier Motor Corporation of America, Waukegan, Ill. As explained above, engine 10 may be an automobile engine, or any other type of engine which can benefit from the use of diagnostic software. Engine 10 includes an electronic control unit (ECU) 14 coupled to engine components such as engine ignition 16 and fuel injectors 18, which can be conventional or non-conventional. Generally, ECU 14 controls operation of engine 10 and provides information to the operator by controlling various indicator lights in gauges. ECU 14 controls operation of engine 10 by, for example, controlling firing of spark plugs via ignition 16 and controlling supply of fuel to the engine cylinders via injectors 18. Computer 12 is a diagnostic computer that is coupled by a command/data link 25 to ECU 14 for use by a diagnostic operator or repairman for servicing engine 10. Computer 12 includes a microprocessor 20 coupled to a random access memory (RAM) 22 and a read only memory (ROM) 24. Computer 12 also has a monitor 21 with a display screen 23. In one embodiment, computer 12 is a battery-powered laptop computer, and processor 20 is a 286 MHz MSDOS processor. Of course, other operating systems such as MacOS, Linux or Windows NT could be used. Likewise, higher speed processors such as 386 MHz, 486 MHz, Pentium, PentiumII, Cyrix, AMD, Celeron or other more modern processor may be used. Also, ROM 24 includes, in one embodiment, at least 400 kilobytes of available memory. However, the required amount of memory depends upon the program in which the present invention is associated. Computer 12 also includes, for example, a monochromatic or color display 23 and a keyboard 25 for entry of user commands and data Many other types and models of computers can be used, and the present invention is not limited to practice in connection with any one particular type of computer. Commands, programs and data are transmitted from microprocessor 20 to ECU 14 via link 25. Commands, programs and data are transmitted from ECU 14 to processor 20 via link 25. For some operations, only commands will be sent and for other operations only data will be sent. Programs may be sent only on start-up or may be sent periodically as various functions all are selected and activated. As described hereinafter in more detail, ROM 24 includes text files 26 and database files 28. Files 26 and 28 are loaded into ROM from a CD, a high-capacity disk such as a xe2x80x9cZIP diskxe2x80x9d, xe2x80x9cJAZZ diskxe2x80x9d or xe2x80x9cSuper Diskxe2x80x9d, or just a standard floppy disk, e.g., a 1.44 Meg 3xc2xd inch floppy disk, depending primarily upon the size and complexity of the files 26 and 28. Text files 26 contain the text for various screen displays to be shown on the display of computer 12. Although the term xe2x80x9ctext filesxe2x80x9d is used, it will be understood that this term means any file containing data which can be read by computer 12 to cause the display of textual material, or screen displays, on display screen 23. Database files 28 contain the fault codes and other information to be utilized in connection with diagnostic operations. By having text files 26 external separate from the program, no display text internal to the program is required, and there is no need to recompile the program just for a language change. That is, if an operator wants the screens displayed in a different language from the current language (e.g., the operator wants the screen displays in French rather than English), the operator simply follows the steps described below and the program then utilizes the text files corresponding to the selected language to generate screen displays. Prior to operation, the diagnostic software program is loaded into computer 12 from, for example, a floppy disk. An install program is run to cause the diagnostic program, including text files and data bases, to be loaded into ROM 24. In one specific exemplary embodiment, sixteen text files are used for storing text associated with eight different languages such as, for example, English, Francais (French), Espanol (Spanish), Portuguese, Deutsch (German), Italiano (Italian), Swedish, and Finnish. Any multiple number of languages could be used. Upon initialization of the diagnostic program now stored in computer 12, microprocessor 20 uploads from ROM 24 and into RAM 22 text files 26 having the screen display information in English, and causes a screen display 30 (shown in FIG. 2) to be shown on the display screen 23. Although the default language in the embodiment described herein is English, the default language could be any one of the languages for which text files exist. Screen display 30 provides three options 1) (FFI DIAGONOSTICS), 2) (SERVICE UTILITIES) OR X (EXIT PROGRAM). Upon receiving an input indicating that an operator has pressed a 1 key on keyboard 25 in response to the display 30, processor 20 causes a screen display 50 as shown in FIG. 3 to be shown on the computer display. If the operator desires at this stage to configure the computer for a non-English language, the operator may enter a setup program by pushing the F3 key on keyboard 25. Upon receiving an input indicating that an operator has pressed the F3 key on keyboard 25, processor 20 then causes a screen display 52 as shown in FIG. 4 to be shown on the computer display. An operator can then select 1 (SELECT LANGUAGE), 2 (CHANGE COMM PORT), or x (EXIT SETUP) on keyboard 25. If the operator selects a 1, then processor 20 causes a language selection screen display 54 as shown in FIG. 5 to be shown on display screen 23. When screen display 54 is being displayed on-screen 23, an operator can then select 1 (ENGLISH), 2 (FRANCAIS), 3 (ESPANOL), 4 (PORTUGUESE), 5 (GERMAN), 6 (ITALIAN), 7 (SWEDISH), 8 (FINNISH) or x (EXIT) on keyboard 25. For example, if the operator selects 2, i.e., French, then microprocessor 20 uploads from ROM 24 the text file(s) corresponding to French text into RAM 22. That is, the text files from the selected language are copied over the text files currently stored in RAM 22. Processor 20 when generating screen displays then subsequently uses the most recently uploaded text files and returns the operator to screen display 50 (shown in FIG. 3). Upon pressing any key on keyboard 25 while screen 50 is displaying, computer 12 displays the main menu screen display 56 (shown in FIG. 6). Generally, and in some diagnostic testing operations, microprocessor 20, in response to operator commands, instructs ECU 14 to energize certain selected components to verify operation, e.g., to verify operation of ignition 16 and fuel injectors 18. In verifying operation of ignition 16, for example, a command from processor 20 causes ECU 14 to fire a respective ignition circuit, and the operator can observe whether the circuit does, in fact, fire. In one embodiment the portion of the ignition preceding the ignition coil is all in ECU 14, so that ECU 14 may electrically determine if the ignition coil, is working properly. This allows the ECU to determine whether it is spark plugs or ignition coil that is causing any observed failure to fire. In other diagnostic testing operations, microprocessor 20 instructs ECU 14 to display various operating conditions of engine 10. FIGS. 6-10 illustrate screen displays associated with performing an ignition coil test. After performing the program set-up operations as described above, processor 20 causes main menu screen display 56 as shown in FIG. 6 to be displayed on the computer display. If, for example, the operator desires to perform a static test, the operator presses the 2 key on keyboard 25. Upon receipt of data indicating the user selection, processor 20 executes the routine associated with the selected test. Particularly, upon receiving an input corresponding to the user selection of 2, processor 20 causes a static testing menu screen display 58 as shown in FIG. 7 to be displayed. The operator can then select one of four tests by selecting 1-4, or can exit by selecting x. If the operator desires to test ignition firing, the operator selects 2. It will be understood that any number of static tests could be performed, and that the four tests shown in FIG. 7 are only exemplary. Upon receipt of the user selection of 2, processor 20 causes a warning screen display 60 as shown in FIG. 8 to be displayed. The operator can then continue by either canceling the test by selecting less than ESC greater than , or proceeding with the test by selecting any other key of keyboard 25. If processor 20 receives an input indicating that the operator desires to continue with testing, then processor 20 causes a message screen display 62 as shown in FIG. 9 to be displayed. The message instructs the operator to remove all spark plug leads and to connect the leads to a spark tester and to set the air gap. Once the operator has carried out the instructions, the operator continues by pressing any key of keyboard 25. Upon receipt of an input indicating that the operator has performed the instructed tasks, processor 20 then causes an ignition selection screen display 64 as shown in FIG. 10 to be displayed. The operator can then cause any of the ignition circuits to fire by selecting 1-6, or can exit by selecting x. The numbers 1-6 correspond to respective cylinders of a six cylinder engine. For example, if the operator selects 2, processor 20 communicates to ECU 14 that the ignition circuit, or coil, corresponding to cylinder #2 should fire. The operator can then observe whether a spark is generated by such ignition circuit by observing the spark tester after making the selection. In addition to static tests, tests of the engine under operating conditions can be performed. In prior versions of diagnostic software for such applications, it was necessary to reduce the engine speed to idle before running the diagnostic software test for the engine under operating conditions. Accordingly, in those versions a screen warning was presented warning the operator to reduce the speed to title before proceeding with the remaining diagnostic tests. However, the software has now been upgraded to allow the running of the diagnostic tests at full speed, so that warnings no longer appear. The diagnostic software in one embodiment also provides a screen display 66 (shown in FIG. 11) of the operation the engine 10 divided into a sensor section 68, a switches section 70, a voltage section 72 and the results section 74. Sensor section 68 displays air temperature, water temperature, ECU temperature, ignition duration, and TPS (throttle position sensor) actual and calculated values when connected to a water cooled ECU. With a water-cooled ECU, such indications are very important to determine whether or not the ECU is being properly cooled by the water coolant system. ECU malfunctions caused by coolant system malfunctions can now be identified and corrected without destroying a replacement ECU in order to determine that the ECU is not the problem. In one embodiment section 68 displays not only the temperature in degrees Fahrenheit, but also the temperature in degrees Centigrade and the voltage output of the temperature sensor involved. This allows the diagnostic operator to determine if the sensor is working properly by checking to see if proper voltage signals are being provided by the sensors. This section 68 also allows the diagnostic operator to determine if the temperature of air, water and ECU are high or low. Section 70 provides indication of various switch conditions such as overheat, oil pressure within limits, shift interrupt or activated, water found in fuel, and whether or not the engine is in S.L.O.W. Section 72 provides the current voltage is for the alternator, both in the 12 volt and the 26/46 volt portions. Section 74 includes engine speed, ignition timing, spark duration, fuel injection timing, fuel pulse width and barometric pressure. The presence of all these conditions in a single screen display is of particular value to the diagnostic operator or service repairman. If, instead of selecting key 1 in response to screen display 30 (shown in FIG. 2), the operator selects key 2 to select service utilities programs, screen 74 (shown in FIG. 12) appears. This screen display 74 is the service utilities menu and allows the operator to select from four subprograms 1) SERVICE REPORTS, 2) INJECTOR INFORMATION, 3) SERVICE ECU, and 4) RETURN TO PROGRAM SELECTION MENU. If the operator presses key 1 of keyboard 25 to call up the service reports sub program, the service report menu screen display 76 (shown in FIG. 13) will be displayed. This menu gives the operator six choices 1) VIEW CURRENT ECU REPORT, 2) SAVE CURRENT ECU REPORT, 3) PRINT CURRENT ECU REPORT, 4) VIEW SAVED REPORT, 5) PRINT SAVED REPORT, 6) RETURN TO SERVICE UTILITIES MENU. If the operator then presses key 1 of keyboard 25, a very complete two-page ECU printout, as shown in FIG. 14 and FIG. 15, is provided. If the operator presses key 2 of keyboard 25 in response to screen 74 (shown in FIG. 12), injectors information screen display 82 (shown in FIG. 16) is displayed on screen 23. Screen display 82 provides the serial number of each injector for each cylinder of engine 10. This allows the operator to verify that the proper type of injector is installed in engine 10. If an incorrect injector, for example injector 1, is detected, a replacement can be installed and the screen 82 will automatically indicate by a second screen display 84 (shown in FIG. 17) that such injector 1 has been replaced. If the operator determines that replacement of the ECU 14 is required, the operator obtains screen display 74 (shown in FIG. 12) as described above, and presses key 3 on keyboard 25 to call screen display 86 (shown in FIG. 19) on display screen 23. Screen display 86 is a warning screen that advises the operator of the operations that must be performed before the ECU can be replaced using the diagnostic software properly. The operator now confirms that the required operations have been performed and ECU 14 is ready to be replaced. Upon such confirmation, the operator presses any key of keyboard 25 to cause microprocessor 20 to transfer all of the specific ECU information from the replaced ECU into microprocessor 20 and, upon completion of such information transfer, to call up screen display 88 (shown in FIG. 20) indicating successful transfer. Screen display 88 further instructs the operator to now turn the ignition key off, replace the existing ECU with the replacement ECU, and turn the ignition key back on and then press any key the keyboard 25. Upon sensing that a key has been pressed on keyboard 25, microprocessor 20 now transfers the stored information from microprocessor 20 relative to the existing ECU into the replacement ECU and calls up screen display 90 (shown in FIG. 21) to indicate the successful completion of the information transfer from the existing ECU into the new replacement ECU. 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. |
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abstract | The present invention is directed to a mathematical approach to detect faults by reconciling known data driven techniques with a physical understanding of the HVAC system and providing a direct linkage between model parameters and physical system quantities to arrive at classification rules that are easy to interpret, calibrate and implement. The fault modes of interest are low system refrigerant charge and air filter plugging. System data from standard sensors is analyzed under no-fault and full-fault conditions. The data is screened to uncover patterns though which the faults of interest manifest in sensor data and the patterns are analyzed and combined with available physical system information to develop an underlying principle that links failures to measured sensor responses. These principles are then translated into online algorithms for failure detection. |
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claims | 1. A method of chemical decontamination which comprises, in a boiling water reactor nuclear power plant provided with a first structural member having a surface which contacts with a coolant and is made of stainless steel and a second structural member having a surface which contacts with a coolant and is made of carbon steel or an iron-based alloy containing chromium and being inferior in corrosion resistance to the stainless steel, pretreating the second structural member with an oxidation decontaminating solution containing an oxidation decontaminating agent applied to both the first structural member and the second structural member, thereby increasing corrosion resistance of the second structural member, and thereafter decontaminating the first structural member and the second structural member with a reduction decontaminating solution containing a reduction decontaminating agent applied to the first structural member and the second structural member, to remove radionuclides from both the first structural member and the second structural member. 2. A method of chemical decontamination which comprises, in a boiling water reactor nuclear power plant provided with a first structural member having a surface which contacts with a coolant and is made of stainless steel and a second structural member having a surface which contacts with a coolant and is made of carbon steel or an iron-based alloy containing chromium and being inferior in corrosion resistance to the stainless steel, pretreating the second structural member with an oxidation decontaminating solution containing an oxidation decontaminating agent applied to both the first structural member and the second structural member at a state that the first structural member and the second structural member are communicated to each other, thereby increasing corrosion resistance of the second structural member, and thereafter decontaminating the first structural member and the second structural member with a reduction decontaminating solution containing a reduction decontaminating agent applied to the first structural member and the second structural member at a state that the first structural member and the second structural member are communicated to each other, to remove radionuclides from both the first structural member and the second structural member. 3. The method of chemical decontamination according to claim 1 wherein the nuclear power plant is a nuclear power plant which has experienced the HWC (hydrogen water chemistry) operation. claim 1 4. The method of chemical decontamination according to claim 1 wherein the reduction decontaminating solution contains hydrazine. claim 1 5. The method of chemical decontamination according to claim 1 wherein a temperature of the oxidation decontaminating solution is in a range higher than 70xc2x0 C. and lower than 100xc2x0 C. claim 1 6. The method of chemical decontamination according to claim 5 wherein the temperature of the oxidation decontaminating solution is in a range not lower than 75xc2x0 C. and lower than 100xc2x0 C. claim 5 7. The method of chemical decontamination according to claim 1 further including, after a completion of a reduction decontamination with the reduction decontaminating solution, the step of subjecting the reduction decontaminating agent contained in the reduction decontaminating solution to a decomposition treatment. claim 1 8. The method of chemical decontamination according to claim 4 further including, after a completion of a reduction decontamination with the reduction decontaminating solution, the step of subjecting the reduction decontaminating agent and the hydrazine contained in the reduction decontaminating solution to a decomposition treatment. claim 4 9. The method of chemical decontamination according to claim 7 wherein the decomposition treatment is conducted with a catalyst in the presence of an oxidizing agent. claim 7 10. The method of chemical decontamination according to claim 7 wherein the decomposition treatment is conducted by an ultraviolet irradiation in the presence of an oxidizing agent. claim 7 11. The method of chemical decontamination according to claim 1 wherein the iron-based alloy containing chromium and being inferior in corrosion resistance to the stainless steel is an iron-based alloy containing less than 13% by weight of chromium. claim 1 12. The method of chemical decontamination according to claim 1 wherein the reduction decontaminating agent contains at least oxalic acid. claim 1 13. The method of chemical decontamination according to claim 1 wherein the oxidation decontaminating agent contains at least one kind of chemical species selected from the group consisting of chemical species having a reduction potential higher than that of from Fe(3+) to Fe(2+). claim 1 14. The method of chemical decontamination according to claim 13 wherein the chemical species having a reduction potential higher than that of from Fe(3+) to Fe(2+) is at least one kind of chemical species selected from the group consisting of MnO 4 (xe2x88x92), Ce(4+), Cr 2 O 7 (2xe2x88x92), HCrO 4 (xe2x88x92), BrO 3 (xe2x88x92), ClO 3 (xe2x88x92), IO 3 (xe2x88x92), Co(3+), H 2 O 2 and O 3 . claim 13 15. The method of chemical decontamination according to claim 4 wherein the reduction decontaminating solution contains hydrazine to control the pH of an oxalic acid solution used as a reduction decontaminating agent to about 2.5. claim 4 16. The method of chemical decontamination according to claim 1 , wherein the pretreating step is performed so that magnetite in an oxide film on the second structural member changes into hematite, thereby increasing the corrosion resistance of the second structural member by resistance of hematite to dissolution by the reductive decontaminating solution. claim 1 17. The method of chemical decontamination according to claim 1 , wherein the pretreating step is performed so that a corrosion resistant substance forms on the second structural member, thereby increasing the corrosion resistance of the second structural member by resistance of the substance to dissolution by the reductive decontaminating solution. claim 1 |
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abstract | A system includes a containment vessel configured to prohibit a release of a coolant, and a reactor vessel mounted inside the containment vessel. An outer surface of the reactor vessel is exposed to below atmospheric pressure, wherein substantially all gases are evacuated from within the containment vessel. |
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description | The present invention relates to an operational support device and an operational support method for a nuclear power plant. During operation and maintenance of a nuclear power plant, it is necessary to monitor and control the operating status and the safety status of the whole nuclear power plant and respective equipment constituting the nuclear power plant. Therefore, various monitoring devices and control devices are employed in nuclear power plants. A known monitoring device includes, for example, a large display device that is provided in a control room of the nuclear power plant so as to be visible from any place in the control room and personal display devices for operators, and each of the display devices provides a required display in accordance with the operating state, the level of emergency, and the level of importance. For example, the control device enables various types of control operations by applying a display device, such as a CRT display or liquid crystal display, to an operating console, and by performing a mouse operation or a touch operation on this display device. An example of the monitoring device of such a nuclear power plant is the technology disclosed in PTL 1, in which information about the nuclear power plant is layered, where information in the upper layers is always displayed, and information in the middle or lower layers is displayed on a plurality of displays, and the displayed images are arranged, thereby allowing the state of the whole plant to be understood. In a nuclear power plant, for example, in the event of an abnormality, such as tripping of a nuclear reactor, activation of an emergency core cooling system (ECCS), and so forth, the operator is required to perform many tasks immediately, such as collecting information related to safety-related systems, checking the state of the power plant, and so forth. {PTL 1} Japanese Unexamined Patent Application, Publication No. 2000-249782 However, the information indicating the safety status, required by the operator, may not necessarily be displayed at all times on the display devices and monitoring devices that are employed in a highly computerized nuclear power plant. Therefore, there is a problem that, in the event of an abnormality in the nuclear power plant, it may take a long time to ascertain the safety status of the nuclear power plant, i.e., in which piece of equipment the abnormality occurs, what kind of abnormality occurs, and so forth. In addition, there are other problems; for example, even if operations performed by the operator are restricted due to maintenance of facilities critical for safety, the operator is unaware of that situation. The present invention has been conceived to solve the problems described above, and an object thereof is to provide an operational support device and an operational support method in which it is possible to accurately ascertain the state of a nuclear power plant and which can assist an operator in performing prescribed operations immediately after the occurrence of an abnormality. In order to solve the aforementioned problems, the present invention employs the following solutions. A first aspect of the present invention is an operational support device of a nuclear power plant comprising: an operating-status confirming display unit that displays whether a system or equipment required for operating the nuclear power plant is functioning in accordance with a command signal for the system or the equipment; and a bypassed-and-inoperable state display unit that displays whether the system or the equipment is operable. According to the first aspect of the present invention, prescribed command signals from the control device etc. of the nuclear power plant are compared and verified with predetermined design conditions during the operation of the nuclear power plant regularly or continuously. As a result, the operating-status confirming display unit display whether the system and the equipment required for operating the nuclear power plant are properly functioning in accordance with a prescribed command signal. By doing so, for example, in the event of an abnormality that risks the safety in any of the nuclear power plant, it is possible for an operator to understand immediately which system or equipment has caused the abnormality etc., and to take countermeasures against the abnormality at an early stage. In addition, because the bypassed-and-inoperable state display unit also displays whether, at that time, these systems and equipment are in a state where they are operable by the operator, it is possible for the operator to understand immediately the operability of the systems and equipment in question. As described above, even in the event of an abnormality, it is possible to support the operator such that they can perform a prescribed operation immediately. On the operating-status confirming display unit or the bypassed-and-inoperable state display unit, for example, it is possible to set, as the displayed items, items such as names of the system and the equipment required for safely operating the whole nuclear power plant and names of the command signals to the system and the equipment, and to display characters and symbols indicating their states. In the above-mentioned first aspect of the present invention, the operating-status confirming display unit displays whether the system or the equipment is functioning in accordance with a trip signal for at least one of a nuclear reactor, a turbine, and a generator. By doing so, for the safety of the nuclear power plant, in the event of an accident in the nuclear reactor, the turbine, or the generator, which are especially critical systems and equipment, it is possible to allow the operator to immediately understand the fact that the abnormality occurs, and to help taking countermeasures against the abnormality at an early stage. In the above-mentioned first aspect of the present invention, the bypassed-and-inoperable state display unit may display whether the nuclear reactor or the turbine is operable. By doing so, for the safety of the nuclear power plant, even in the event of an abnormality in the nuclear reactor or the turbine, which is especially critical system and equipment, it is possible to support the operator such that they can perform a prescribed operation immediately. In the above-mentioned first aspect of the present invention, the bypassed-and-inoperable state display unit may display a maintenance or an inspection state of an emergency generator or safety injection pump during the operation of the nuclear power plant. By doing so, it is possible to allow the operator to understand the state of maintenance or inspection, such as a state where maintenance or inspection of the emergency generator or the safety injection pump is under progress during the operation of the nuclear power plant etc. Therefore, it is possible to ensure the operational support for the operator, such that it is possible to avoid an inadvertent operation etc., and at the same time, to secure the safety of the worker performing the maintenance or inspection. In the above-mentioned first aspect of the present invention, the operating-status confirming display unit may displays all operating states of the system or the equipment that are displayed on the operating-status confirming display unit in the form of a list. By doing so, it is possible to visually check all of the operating states of the system or the equipment displayed on the operating-status confirming display unit easily without the need for an operation such as displaying the required information and the desired information; therefore, it is possible to provide required information to the operator immediately, and to support him or her such that the required operation can be performed. In the above-mentioned first aspect of the present invention, the operating-status confirming display unit and the bypassed-and-inoperable state display unit may be displayed side-by-side. By doing so, in combination with or in comparison with the operating state of the system or the equipment displayed on the operating-status confirming display unit, it is possible to understand easily and immediately if the system or the equipment is in a state operable by the operator, and therefore it is possible to provide required information to the operator even more immediately, and to support him or her such that required operation can be performed. A second aspect of the present invention is an operational support method of a nuclear power plant, comprising; displaying whether a system or equipment required for operating the nuclear power plant is functioning in accordance with a command signal for the system or the equipment, and displaying whether the system or the equipment is operable. According to the present invention, it is possible to accurately ascertain the state of a nuclear power plant, and to assist an operator in performing prescribed operations immediately after the occurrence of an abnormality. An embodiment of an operational support device according to the present invention will be described below with reference to the drawings. In this embodiment, an example in which an operational support device is applied to a nuclear power plant employing a pressurized water reactor (PWR: Pressurized Water Reactor) will be described. FIG. 1 is a block diagram showing, in outline, a configuration in which an operational support device 1 according to this embodiment is applied to a nuclear power plant, and FIG. 2 is a reference diagram showing, in outline, the configuration of a control room of a nuclear power plant 2. As shown in FIGS. 1 and 2, the operational support device 1 is disposed in a control room 11 of the nuclear power plant and is connected to the nuclear power plant 2 via a control device 3 for running and controlling the nuclear power plant 2. The operational support device 1 is connected to an operating console 4, an operational command console 5, and a voice notification device 12. The nuclear power plant 2 is mainly constructed of a reactor vessel 25, a steam generator 6, a turbine 7 (high-pressure turbine and low-pressure turbine), a condenser 8, and a generator 9. The reactor vessel 25 is a vessel accommodating a reactor core, and decay heat is extracted by circulating coolant around the nuclear fuel in the reactor vessel 25. In the reactor vessel, the reactor core, a core barrel that surrounds the reactor core, a core baffle, and a supporting structure of a fuel assembly are provided. The steam generator 6 generates steam using the heat extracted from the reactor vessel 25 and supplies the steam to the high-pressure turbine of the turbine 7. The high-pressure turbine rotates a shaft directly connected to the generator 9 by the high-pressure steam supplied from the steam generator 6 and supplies the discharged air via a moisture separator reheater to the low-pressure turbine of the turbine 7. The low-pressure turbine is driven by the steam that has been discharged from the high-pressure turbine and dried and reheated by the moisture separator reheater and rotates a shaft of a generator. The condenser 8 condenses the steam used in the high-pressure turbine and the low-pressure turbine into water to reduce its volume, thereby achieving a high vacuum state to increase the efficiency of the turbines. The generator 9 converts the rotating force of the shaft that is driven by the turbine 7 into electrical power, and outputs the power. The turbine 7 may have a multi-stage configuration including a high-pressure configuration and a low-pressure configuration or a medium-pressure configuration. The control device 3 operates or controls the respective systems and equipment constituting the nuclear power plant 2 according to predetermined design conditions, and the control device 3 sends, to the operational support device 1, digital signals indicating process parameters of the nuclear reactor, such as the water level, pressure, flow rate, and so forth, and the states of the respective systems and equipment, and various signals, such as command signals, operation signals, and alarm signals, to the respective systems and equipment. General-purpose or specialized computers can be employed as the control device 3. Such computers (not shown) are provided with a CPU (central processing unit), ROM (Read Only Memory), RAM (Random Access Memory), and so forth. The control device 3 stores, for example, programs, the design conditions, and so forth for operating or controlling the respective systems and equipment of the nuclear power plant 2 in ROM etc. in advance; reads out these programs and the design conditions with the CPU etc. in accordance with the signals from the operating console 4, which will be described below, and the signals from the various sensors etc. provided on the respective systems and equipment; loads them in the RAM; and executes information processing/computational processing, thereby operating or controlling the respective systems and equipment. The operational support device 1 is provided with four display units 15, 16, 17, and 18 for enabling the operator to easily understand the state of the nuclear power plant 2 and for supporting the plant operation. The display units 15, 16, 17, and 18 are disposed adjacent to each other so as to form, as a whole, a large display device that can be viewed from any place in the control room. In the event of, for example, normal operation, abnormal operation, or an accident in the nuclear power plant 2, in order to understand the status of the whole nuclear power plant 2, the respective display units 15, 16, 17, and 18 help operators to immediately understand the state of the nuclear power plant 2 by displaying the state of the main system and equipment summarized in a diagram of the schematic configuration of the nuclear power plant 2 and by displaying typical alarms indicating an abnormality in accordance with the various signals sent from the control device 3. In the operational support device 1, as shown in FIG. 3, any of the respective display units 15, 16, 17, and 18 displays an OK monitor 20 as an operating-status confirming display unit that displays whether the system and equipment required for safely operating the whole nuclear power plant are functioning in accordance with the command signals for the system and the equipment; and displays a BISI (Bypassed and Inoperable State Indication) monitor 21 as a bypassed-and-inoperable state display unit that displays whether the system and the equipment are operable. The displayed items on the OK monitor 20 and the BISI monitor 21 are substantially the same, and so, in order to allow easy visual checking of these monitors by the operator, they are provided side by side. In the example shown in FIG. 3, as the displayed items on the OK monitor 20 and the BISI monitor 21, the signals related to the system and the equipment, which are required for safely operating the entire nuclear power plant, include RT (reactor trip signal), TT (turbine trip signal), GT (generator trip signal), SI Sequence (safety injection sequence signal), BO Sequence (blackout sequence signal), φA (containment vessel isolation signal (T signal)), CVI (containment vessel ventilation system isolation signal), CR Isol (central control room ventilation system isolation), φB (containment vessel isolation signal (P signal)), CS (containment vessel spray activation signal), FW Isol (main feedwater isolation signal), MS Isol (main steam isolation signal), and A F/W Pmp Act (auxiliary feedwater initiation signal). The OK monitor 20 displays “OK” when the respective systems and equipment are functioning properly in accordance with these signals, and displays “NG (No Good)” when they are not functioning properly. Each of the above-mentioned displayed items will be described below. RT (reactor trip signal) is a request signal for shutting down the nuclear reactor in the event of an abnormality or accident, and the operation based on this signal requires control rods to be completely inserted by opening a reactor trip breaker. TT (turbine trip signal) is a request signal for shutting down the main turbine in the event of an abnormality or accident, and the operation based on this signal requires a turbine cutoff valve to be closed by reducing turbine emergency shutdown hydraulic pressure by opening a valve. GT (generator trip signal) is a request signal for shutting down the main generator in the event of an abnormality or accident, and the operation based on the signal requires a generator load-break switch to be opened by opening a field circuit breaker of the main generator. SI Sequence (safety injection sequence signal) is a request signal for injecting cooling water containing boric acid into the reactor core in the event of an accident, and the operation based thereon requires a valve to be opened/closed for ensuring water for activating a high-pressure/low-pressure injection pump and for water injection and then to perform sequential activation (activation at certain time intervals to avoid overload of the emergency generator) of the related auxiliary equipment (a nuclear-reactor auxiliary cooling-water pump etc.). BO Sequence (blackout sequence signal) is a request signal for maintaining cooling of the reactor core when a power source external to the power plant is used (in the event of an accident in power transmission line etc.), and the operation based thereon requires a valve to be opened/closed for ensuring water for activating an auxiliary feedwater pump etc. and for water injection and then to further perform sequential activation (activation at certain time intervals to avoid overload of the emergency generator) of the related auxiliary equipment (the nuclear-reactor-auxiliary cooling-water pump etc.). φA (containment vessel isolation signal (T signal)) is a request signal for closing a plurality of isolation valves in order to prevent the release of radiation from the containment vessel in the event of an accident, and the operation based thereon requires a valve to be closed for isolating the flow path (system) that may be a release path from the containment vessel. CVI (containment vessel ventilation system isolation signal) is a request signal for closing an isolation valve and a damper of a plurality of ventilation and air conditioning systems in order to prevent the release of radiation from the containment vessel in the event of an accident, and the operation based thereon requires a valve and a damper to be closed for isolating the flow path (system) that may be a release path from the containment vessel. CR Isol (central control room ventilation system isolation) is a request signal for activating/stopping a fan and for opening/closing a valve and a damper of a plurality of ventilation and air conditioning systems for preventing the radiation from entering a central control room in the event of an accident and establishing a safe air-circulation system, and the operation based thereon requires a fan to be activated/stopped and a valve and a damper to be set in the opened/closed state in a plurality of ventilation and air conditioning systems for preventing the radiation from entering the central control room and establishing a safe air-circulation system. φB (containment vessel isolation signal (P signal)) is a request signal for closing a plurality of isolation valves in order to prevent, in the event of an accident, the release of radiation from the containment vessel while the containment vessel is performed the spraying operation, and the operation based thereon requires a valve to be closed for isolating the flow path (system) that may become a release path from the containment vessel. CS (containment vessel spray activation signal) is a request signal for closing a plurality of isolation valves in order to cool the nuclear reactor/the containment vessel, in the event of an accident, by spraying water (spray) containing chemicals and boric acid from the top of the containment vessel to prevent the release of radiation, and the operation based theron requires a valve to be opened/closed for ensuring water for activating the containment vessel spray pump and for the spraying operation, and sequential activation of the related auxiliary equipment (the nuclear-reactor-auxiliary cooling-water pump etc.). FW Isol (main feedwater isolation signal) is a signal for isolating a main feedwater system in order to prevent, in the event of an accident, excessive supply of water and over-cooling by the steam generator, and the operation based thereon requires the main feedwater pump to be forcedly stopped and a valve to be set in the closed state for isolating the main feedwater system. MS Isol (main steam isolation signal) is a signal for isolating a main steam system in order to prevent, in the event of an accident, over-cooling by the steam generator, and the operation based thereon requires a main steam isolation valve, a bypass valve, and so forth to be set in the closed state. A F/W Pmp Act (auxiliary feedwater initiation signal) is a signal for activating an auxiliary feedwater system for ensuring the removal of heat with the steam generator in response to the isolation of the main feedwater in the event of an accident, and the operation based thereon requires an electric/turbine-actuated auxiliary feedwater pump to be activated, and the lineup state of the valve to be checked in order to isolate the auxiliary feedwater system. In accordance with the above-mentioned signals, the OK monitor 20 displays “OK” when the required operation in response to the signals has been performed properly, and displays “NG” when the operation has not been performed properly. In other words, these signals are sent to the control device 3, and the control device 3 compares and verifies them with predetermined design conditions etc. As a result, the control device 3 sends, to the operational support device 1, a control signal that asks “OK” to be displayed if the required operation has been performed properly and a control signal that asks “NG” to be displayed if the required operation has not been performed properly. In the OK monitor 20, the OK monitor 20 displays “OK” or “NG” in accordance with this control signal. When the above-mentioned signals are not sent, the OK monitor 20 displays “−”. In the BISI monitor 21, regarding the status indicating whether a system or equipment is operable, “−” is always displayed in an operable state, and “NG” is displayed in an inoperable state. A state in which the operability is unknown includes a bypass operation due to maintenance or inspection or a malfunction, a forced stoppage state due to a malfunction etc. of equipment, cutoff of control power, switching the control to a local panel etc., and so forth: these states constitute the information used by an operator for constantly understanding the status in the power plant, and the information needs to be handed over and checked when the operators are changed. In other words, the control device 3 determines, regularly or continuously, the state of the respective systems and equipment by receiving the signals from various sensors etc. provided on the respective systems and equipment. As a result, regarding the status indicating whether each system or equipment is inoperable, the control device 3 sends, to the operational support device 1, the control signal that asks “−” to be displayed if the respective systems and equipment are operable, and the control signal that asks “NG” to be displayed if the respective systems and equipment are inoperable. In accordance with this control signal, the BISI monitor 21 displays “−” or “NG”. When “NG” is displayed on the BISI monitor 21, the system and the equipment showing “NG” are in the inoperable state, and this includes, for example, a state where the inoperable is caused by a malfunction, in addition to a state where the inoperability is due to maintenance or inspection. In particular, if maintenance or inspection of the emergency generator or the safety injection pump is to be performed during the operation of the nuclear power plant 2, in order to ensure the safety of workers performing maintenance or inspection, “NG” is displayed in the BO Sequence or SI Sequence field, and information notifying the inoperable state is provided to the operator. A diesel generator or a gas turbine generator is employed as the emergency generator. In addition, because maintenance or inspection work during the operation of the nuclear power plant 2 can include, for example, maintenance or inspection work on control panels, protection circuits, and so forth, these items can be set as the displayed items on the BISI monitor 21 as required. The indication “NG” can be displayed together with more details of the state by, for example, changing the background color of the indication depending on the degree of inoperability. In other words, because the nuclear power plant is designed to ensure safety with redundant multiple series of devices and systems that operate in the event of an abnormality or accident, yellow is used if only one series thereof is inoperable, and red is used if two or more series are inoperable. By doing so, it is possible to expect an advantage in that, when red is used, a notification indicating that the inoperable state causes an adverse effect on the actual safety of the plant is given, and in addition, when yellow is used, a notification indicating that an even more inoperable state causes an adverse effect on the safety status is given. Flashing signs may be employed for the display of the notification and an abnormality warning signal. The operating console 4 enables complete monitoring and control of the nuclear power plant 2 by the operator, and it is a terminal capable of checking all information of the nuclear power plant 2. Details of the operations performed by the operator are sent from the operating console 4 to the control device 3, and a prescribed control signal is sent from the control device 3 to the nuclear power plant 2 in accordance with the details of the operations, thereby achieving the desired control. Although the operational command console 5 has some restrictions on the operations regarding running of the nuclear power plant 2, unlike the operating console 4, in order to achieve smooth communication with the operator of the operating console 4 it is possible to, for example, display pictures that are in common with the operational command console 5 on the operating console 4. Therefore, it is possible to control and supervise the operator who is operating the operating console 4, by using the operational command console 5. In the event of an abnormality in the nuclear power plant 2, in addition to the beeping sound or flashing sign that notifies the occurrence of abnormality, the voice notification device 12 uses voice to tell the operator the details of the abnormality. As described above, according to the present invention, during the operation of the nuclear power plant 2, the OK monitor 20 displays “OK” or “NG” in the form of a list indicating whether the systems and the equipment required for safely operating the whole nuclear power plant 2 are functioning properly in accordance with the command signals for these systems and the equipment. Therefore, in the event of an abnormality that risks the safety in the nuclear power plant 2, it is possible for the operator to understand immediately and easily which system or equipment has caused the abnormality etc., and to take countermeasures against the abnormality at an early stage. In addition, because the BISI monitor 21 also displays “−” or “NG” to indicate whether, at that time, these systems and equipment are in a state where they are operable by the operator, it is possible for the operator to understand immediately the operability of the systems and equipment in question. Since the BISI monitor 21 also displays “NG” for an inoperable state due to maintenance and inspection work during the operation of the nuclear power plant 2 to provide the operator with information indicating that the systems and equipment in question are inoperable, it is possible to ensure the safety of the worker performing maintenance or inspection. As described above, even in the event of an abnormality, it is possible to support the operator such that they can perform a prescribed operation immediately. In this embodiment, although an example in which “OK”, “NG”, and “−” are displayed on the OK monitor 20 and the BISI monitor 21 has been described, it is possible to display them using other characters or colors. Additionally, it is possible to notify an operator with a signal that is recognizable to the operator, such as an acoustic signal etc. In this embodiment, although the operational support device has been described as being applied to a PWR, it should be naturally appreciated that application to a boiling water reactor (BWR: Boiling Water Reactor, BWR) is possible. The displayed items on the OK monitor 20 and the BISI monitor 21 are not limited to the examples in this embodiment, and appropriate selection is permissible on the basis of the system and the equipment provided in the nuclear power plant to which the operational support device according to the present invention is to be applied. The present invention is not limited to the above-described embodiment. Suitable design modifications are permissible as necessary. The respective elements in the above embodiment include those easily conceivable by a person of ordinary skill in the art and those that are substantially equivalent thereto. 1 operational support device 2 nuclear power plant 3 control device 11 control room 15, 16, 17, 18 display unit 20 OK monitor 21 BISI monitor |
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abstract | A control rod for a nuclear boiling water reactor is described. The control rod has a longitudinal centre axis and control rod blades, each control rod blade having a first and a second side and being substantially parallel to the longitudinal center axis. Each control rod blade comprises an absorber material which extends from a first absorber end to a second absorber end, the distance between the first absorber end and the second absorber end defining an active length. The control rod blades are provided with distance means on the first and second sides of the control rod blades, the distance means extending a distance of at least a third of the active length of the control rod blade. |
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abstract | A charged particle beam exposure system comprising: a charged particle beam emitting device which generates charged particle beams with which a substrate is irradiated, the charged particle beam emitting device generating the charged particle beams at an accelerating voltage which is lower than that at which an influence of a proximity effect occurs; an illumination optical system which adjusts a beam diameter of the charged particle beams so that density of the charged particle beams is uniform; an character aperture in which an aperture hole is formed in a shape corresponding to a desired pattern to be written; a first deflector which deflects the charged particle beams by an electrostatic field that the charged particle beams have a desired sectional shape and travel towards a desired aperture hole and which returns the charged particle beams passing through the aperture hole to an optical axis thereof; a reducing projecting optical system which forms a multi-pole lens field so that the charged particle beams passing through the character aperture substantially reduce at the same demagnification both in X and Y directions when the optical axis extends in Z directions and form an image on the substrate without forming any crossover between the character aperture and the substrate; and a second deflector which deflects the charged particle beams passing through the character aperture by means of an electrostatic field to scan the substrate with the charged particle beams. |
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044774109 | claims | 1. A device for cooling the main vessel (3) of an integrated-type fast fission nuclear reactor comprising a vessel with a symmetry of revolution about a vertical aixs, referred to as the main vessel, which contains a cooling liquid and inside which are arranged the reactor core (7), a plurality of pumps for circulating the cooling liquid, and a plurality of heat exchangers using the cooling liquid as the primary fluid, a partition (10), with a symmetry of revolution about the axis of the vessel, arranged at a predetermined height in said vessel (3) and at the periphery of said core (7), creating in said vessel two separate zones, a first of said zones (14) containing said core (7) and the sodium heated in contact therewith, which leaves said first zone through an inlet for primary fluid in said heat exchangers, in which the outlet for primary fluid is located in the second of said zones (15), containing the liquid cooled by said heat exchangers, which is carried away by said pumps and injected into the lower part of said core, and at least two shells comprising outer and inner shells (16, 17) coaxial with said main vessel (3) and arranged inside the latter over a predetermined height above said partition (10), creating between them, and between the shell (16) of larger diameter, or external shell, and said vessel (3), at least two annular spaces (19, 20) for the passage of said cooling liquid, one of which emerges in the lower part of said core (7) and the other in said second zone (15), said two annular spaces furthermore being in communication, in their upper part, with a zone (24) located in the upper part of said vessel and filled with inert blanketing gas, in which cooling device the annular space (19) delimited by said main vessel (3) and said external shell (16), or external space, is in communication with said second zone (15) containing cooled liquid, and the internal space (20) delimited between the two shells (16, 17) is in communication, via at least one tube (31, 35), with the lower part of said core (7), into which said cold liquid is injected, part of which is thus caused to circulate from bottom to top in the second space (20) before flowing into the external space (19), at the level of the upper zone (24) of said vessel occupied by said blanket of gas, and moving down again, by gravity, to the lower part of said core, remaining in contact with the internal surface of said main vessel (3) and thus cooling the latter. 2. A cooling device as claimed in claim 1, wherein the tube or tubes (31) bringing said internal space (20), delimited by said two shells (16, 17), into communication with the lower part of said core (7) pass through the second zone (15) of said vessel, containing the cold liquid sodium. 3. A cooling device as claimed in claim 1, in the case where said partition (10), or step, separating the interior volume of said main vessel (3) into two parts consists of two walls (11, 12) having a free space between them, said at least one tube (35) bringing the internal annular space (20) between said two shells (16, 17) into communication with the lower part of said core (7) being arranged inside the free space created between the two parts of said step (10). |
claims | 1. An apparatus comprising:a nuclear reactor including a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel;a reactor cavity inside of which is disposed the lower portion of the reactor pressure vessel, the reactor cavity and a wall of the pressure vessel defining an annular gap therebetween; andan annular neutron stop located at an elevation above the uppermost elevation of the nuclear reactor core, the annular neutron stop including an outer neutron stop ring including an outermost neutron stop ring attached to the wall of the reactor cavity and a middle neutron stop ring attached to the outermost neutron stop ring, an inner neutron stop ring attached to the reactor pressure vessel and comprising a high temperature neutron absorbing material that is stable at an exterior temperature of the reactor pressure vessel during nuclear reactor operation, and comprising neutron absorbing material filling, the annular neutron stop being disposed in the annular gap between the reactor pressure vessel and a wall of the reactor cavity,wherein the outer neutron stop ring defines an annular cutout portion in which the inner stop ring is disposed, the middle neutron stop ring comprises the high temperature neutron absorbing material, and the outermost neutron stop ring comprises a neutron absorbing material that is different from the high temperature neutron absorbing material. 2. The apparatus of claim 1 wherein an interface between the outer neutron stop ring and the inner neutron stop ring is stair-stepped or staggered to block neutrons from streaming through the interface. 3. The apparatus of claim 1 wherein the high temperature neutron absorbing material comprises a composition including boron carbide (B4C). 4. The apparatus of claim 1 wherein the neutron absorbing material that is different from the high temperature neutron absorbing material comprises borated concrete. 5. The apparatus of claim 1 wherein the annular neutron stop comprises a boron containing neutron absorber component in a thermally insulating matrix material. 6. The apparatus of claim 5 wherein the thermally insulating matrix material comprises vermiculite. 7. An apparatus comprising:a nuclear reactor including a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel;a reactor cavity inside of which is disposed the lower portion of the reactor pressure vessel, the reactor cavity and a wall of the pressure vessel defining an annular gap therebetween;an annular neutron stop located at an elevation above the uppermost elevation of the nuclear reactor core, the annular neutron stop including an outer neutron stop ring defining an annular cutout portion, an inner neutron stop ring disposed in the annular cutout portion, and comprising neutron absorbing material filling, the annular neutron stop being disposed in the annular gap between the reactor pressure vessel and a wall of the reactor cavity;a tube penetrating through the annular neutron stop; anda neutron plug disposed in the tube at the penetration of the tube through the neutron stop, the neutron plug comprising neutron absorbing material. 8. The apparatus of claim 7 wherein the neutron plug comprises a boron containing neutron absorbing material. 9. The apparatus of claim 7 wherein the tube comprises an excore instrument guide tube and the apparatus further comprises:an excore instrument suspended from the neutron plug by a cable. 10. The apparatus of claim 9 wherein the neutron plug includes a tortuous path and the apparatus further includes:one or more wires passing through the tortuous path and operatively connecting with the excore instrument. 11. An apparatus comprising:a nuclear reactor including a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel;a reactor cavity inside of which is disposed the lower portion of the reactor pressure vessel;an annular neutron stop located at an elevation above the uppermost elevation of the nuclear reactor core, the annular neutron stop comprising neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity;an excore instrument guide tube penetrating through the annular neutron stop;a neutron plug including a tortuous path disposed in the excore instrument guide tube at the penetration of the excore instrument guide tube through the neutron stop, the neutron plug comprising neutron absorbing material;an excore instrument suspended from the neutron plug by a cable; andone or more wires passing through the tortuous path and operatively connecting with the excore instrument. 12. The apparatus of claim 11 wherein the annular neutron stop comprises:an outer neutron stop ring attached to the wall of the reactor cavity; andan inner neutron stop ring attached to the reactor pressure vessel. 13. The apparatus of claim 12 wherein an interface between the outer neutron stop ring and the inner neutron stop ring is stair-stepped or staggered to block neutrons from streaming through the interface. 14. The apparatus of claim 12 wherein the inner neutron stop ring comprises a high temperature neutron absorbing material that is stable at an exterior temperature of the reactor pressure vessel during nuclear reactor operation and the outer neutron stop ring comprises a neutron absorbing material that is different from the high temperature neutron absorbing material. 15. The apparatus of claim 11 wherein the annular neutron stop comprises a boron containing neutron absorber component in a thermally insulating matrix material. |
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summary | ||
050376039 | claims | 1. A tool for removing a hollow locking tube from a top nozzle of a nuclear fuel assembly, said tool comprising: (a) an elongated hollow tubular assembly having upper and lower opposite end portions with said lower end portion insertable in a hollow locking tube, said lower end portion including an outer tubular element having a circumferential guide wall with a plurality of circumferentially spaced apertures, said lower end portion also including a plurality of expandable and contractible lifting members disposed within said tubular element and having catch elements extendable through and retractable from said apertures of said guide wall of said tubular element for engagement with and disengagement from a lower edge of the locking tube; and (b) an actuator assembly mounted through said tubular assembly for axial movement therealong and having upper and lower end portions, said lower end portion for actuating said lifting members of said tubular assembly lower end portion between expanded and contracted conditions to extend and retract their catch elements through and from said apertures in said guide wall of said tubular assembly lower end portion for engaging with and disengaging from the lower edge of the locking tube; (c) said lower end portion of said actuator assembly including an elongated shaft member extending between and past said lifting members of said tubular assembly lower end portion, said shaft member having upper and lower tandemly-arranged segments, said upper segment being larger in outside diameter than said lower segment such that downward movement of said shaft member removes said lower segment from between said lifting members and inserts said upper segment therebetween causing engagement therewith and expansion of said lifting members from contracted to expanded condition, whereas upward movement of said shaft member removes said upper segment from between said lifting members and inserts said lower segment therebetween permitting contraction of said lifting members from the expanded to contracted condition. said lifting members of said tubular assembly lower end portion have tapered tips; and said shaft member extending between and past said lifting members includes a retractor member mounted at a lower end of said shaft member, said retractor member having a tapered portion for engaging said tapered tips of said lifting members and ensuring that said lifting members move from said expanded to contracted condition as said shaft member is moved upwardly. (a) an elongated hollow tubular assembly having upper and lower opposite end portions, said lower end portion of said tubular assembly being insertable in said locking tube and including a guide member and a plurality of locking tube lifting members; (b) said guide member of said tubular assembly having a central axis and being composed of an elongated hollow tubular element having an open lower end and a guide element interfitting said tubular element at said open lower end and having a body portion projecting therefrom, said tubular element at a region thereof spaced from said lower end having a plurality of apertures defined at circumferentially spaced locations about said tubular element, said end of said tubular element and said body portion of said guide element having substantially the same outside diameter so as to provide a continuous smooth transition from said guide element body portion to said tubular element end for facilitating insertion of said guide and tubular elements of said guide member into the hollow locking tube without catching on an upper edge of the locking tube at said transition of said guide member; (c) said lifting members of said tubular assembly extending within said hollow tubular element of said guide member and composed of a plurality of finger elements being movable between expanded and contracted conditions away from and toward said central axis of said guide member and catch elements defined on said respective finger elements projecting radially outwardly from said central axis of said guide member and aligned with said apertures in said tubular element of said guide member, said catch elements projecting from said tubular element through said apertures for underlying and engaging a lower edge of the locking tube when said finger elements are at said expanded condition and retracted from said apertures within said tubular element for disengaging and retraction from the lower edge of the locking tube when said finger elements are at said contracted condition; and (d) means extending through said hollow tubular assembly and being actuatable for causing expansion of said finger elements from said contracted to expanded condition and permitting contraction of said finger elements from said expanded to contracted condition; (e) said lower end portion of said actuator assembly including an elongated shaft member extending between and past said lifting members of said tubular assembly lower end portion, said shaft member having upper and lower tandemly-arranged segments, said upper segment being larger in outside diameter than said lower segment such that downward movement of said shaft member removes said lower segment from between said lifting members and inserts said upper segment therebetween causing engagement therewith and expansion of said lifting members from contracted to expanded condition, whereas upward movement of said shaft member removes said upper segment from between said lifting members and inserts said lower segment therebetween permitting contraction of said lifting members from the expanded to contracted condition. said lifting members of said tubular assembly lower end portion have tapered tips; and said shaft member extending between and past said lifting members includes a retractor member mounted at a lower end of said shaft member, said retractor member having a tapered portion for engaging said tapered tips of said lifting members and ensuring that said lifting members move from said expanded to contracted condition as said shaft member is moved upwardly. a force-receiving member attached to said upper end portion of said tubular assembly; and a force-imparting member disposed about said upper end portion of said tubular assembly and slidably movable therealong in a reciprocating manner for delivering at least one forceful impact against said force-receiving member. means attached to said upper end portion of said actuator assembly and being operable to cause axial movement of said elongated shaft thereof. (a) an elongated hollow tubular assembly having upper and lower opposite end portions, said lower end portion of said tubular assembly being insertable in said locking tube and including (b) an actuator assembly having upper and lower end portions and being mounted through said hollow tubular assembly for axial movement therealong, said lower end portion of said actuator assembly including a force-receiving member attached to said upper end portion of said tubular assembly; and a force-imparting member disposed about said upper end portion of said tubular assembly and slidably movable therealong in a reciprocating manner for delivering at least one forceful impact against said force-receiving member. means attached to said upper end portion of said actuator assembly and being operable to cause axial movement of said elongated shaft thereof. 2. The tool as recited in claim 1, wherein said tubular assembly lower end portion includes a guide element interfitting said tubular element at an open lower end thereof and having a body portion projecting therefrom, said end of said tubular element and said body portion of said guide element having substantially the same outside diameter so as to provide a continuous smooth transition from said guide element body to said tubular element end for facilitating insertion of said guide and tubular elements of said tubular assembly lower end portion into the hollow locking tube without catching on an upper edge of the locking tube at said transition. 3. The tool as recited in claim 2, wherein said guide element body portion has an upper cylindrical segment and a lower conical nose, said cylindrical segment having a section of reduced diameter being inserted into said end of said tubular element. 4. The tool as recited in claim 1, wherein: 5. A tool for removing a hollow locking tube from a top nozzle of a nuclear fuel assembly, said tool comprising: 6. The tool as recited in claim 5, wherein said guide element body portion has an upper cylindrical segment and a lower conical nose, said cylindrical segment having a section of reduced diameter being inserted into said end of said tubular element. 7. The tool as recited in claim 5, wherein said actuatable means includes an actuator assembly having upper and lower end portions and being mounted through said hollow tubular assembly for axial movement therealong. 8. The tool as recited in claim 5, wherein: 9. The tool as recited in claim 7, further comprising: 10. The tool as recited in claim 9, wherein said force-receiving member is a bail assembly including a generally flat plate fixed to said upper end portion of said tubular assembly and a handle connected to and extending upwardly from said plate. 11. The tool as recited in claim 10, further comprising: 12. For use with a reconstitutable fuel assembly including a top nozzle with an adapter plate having at least one passageway, at least one guide thimble with an upper end portion and an attaching structure having a hollow locking tube for releasably locking the upper end portion of the guide thimble within the passageway of the top nozzle adapter plate, said locking tube having upper and lower opposite edges, a tool for removing the locking tube from its locking position, comprising: 13. The tool as recited in claim 12, further comprising: 14. The tool as recited in claim 13, wherein said force-receiving member is a bail assembly including a generally flat plate fixed to said upper end portion of said tubular assembly and a handle connected to and extending upwardly from said plate. 15. The tool as recited in claim 14, further comprising: 16. The tool as recited in claim 13, wherein said guide element body portion has an upper cylindrical segment and a lower conical nose, said cylindrical segment having a section of reduced diameter being inserted into said end of said tubular element. |
abstract | There is provided a collector system. The collector system includes a first collector mirror and a second collector mirror. The first collector mirror receives EUV light from a light source at a first aperture angle via a first beam path, and reflects the EUV light at a second aperture angle along a second beam path. The first aperture angle is larger than or substantially equal to the second aperture angle. The second mirror receives the EUV light from the first mirror at the second aperture angle. The collector is an oblique mirror type normal incidence mirror collector system. |
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summary | ||
claims | 1. A system including at least one control drum for a nuclear reactor, the at least one control drum comprising:an outer shell;an inner shell;a plurality of tubes, the plurality of tubes including at least one neutron absorbing tube and at least one neutron scattering tube,the at least one neutron absorbing tube configured to contain neutron absorbing material, the neutron absorbing material including at least one of boron material, carbide material, hafnium material, gadolinium material, or any combinations thereof, andthe at least one neutron scattering tube configured to contain neutron scattering material, the neutron scattering material including at least one of beryllium material, graphite material, europium material, or any combinations thereof; andthe at least one baffle plate including a plurality of perforations, wherein each perforation of the plurality of perforations is configured to receive and support a corresponding tube of the plurality of tubes. 2. The system of claim 1, whereinthe plurality of perforations are arranged along at least one ring of the at least one baffle plate, the at least one ring including at least a first sector and a second sector, the first sector including a plurality of neutron absorbing tubes and the second sector including a plurality of neutron scattering tubes, the plurality of neutron absorbing tubes including the at least one neutron absorbing tube, and the plurality of neutron scattering tubes including the at least one neutron scattering tube. 3. The system of claim 1, whereinthe at least one baffle plate is a plurality of baffle plates; andthe plurality of baffle plates are sequentially arranged between the outer shell and the inner shell along a first direction of the control drum. 4. The system of claim 1, whereineach perforation of the plurality of perforations includes at least one spring configured to allow for expansion of a corresponding supported tube. 5. The system of claim 1, further comprising:a drive mechanism including a drive shaft, the drive shaft configured to mate with the inner shell via a magnetic coupling; andthe drive mechanism is further configured to rotate the control drum via the drive shaft such that the at least one neutron absorbing tube faces at least one nuclear fuel rod during a first state of the nuclear reactor and the at least one neutron scattering tube faces the at least one nuclear fuel rod during a second state of the nuclear reactor. 6. The system of claim 5, further comprising:at least one torsional spring attached to the inner shell, the at least one torsional spring configured to rotate the control drum via the drive shaft such that the at least one neutron absorbing tube faces the at least one nuclear fuel rod during a third state of the nuclear reactor. 7. The system of claim 6, wherein the third state is a fail-safe state of the nuclear reactor where at least one of the magnetic coupling or the drive mechanism has failed. 8. The system of claim 1, whereinthe neutron absorbing material has a form of a powder, pellets, or a solid; andthe neutron scattering material has a form of a powder, pellets, or a solid. 9. The system of claim 1, wherein the control drum is horizontally mounted in a reflector region surrounding a nuclear fuel assembly of the nuclear reactor, the control drum horizontally mounted with respect to the nuclear reactor. 10. The system of claim 1, wherein the control drum is configured to be installed in the nuclear reactor, the nuclear reactor being a mobile nuclear reactor. 11. A nuclear reactor comprising:a plurality of nuclear fuel rods;a plurality of control drums, each control drum of the plurality of control drums attached to a drive shaft of a plurality of drive shafts, and at least one control drum of the plurality of control drums includes,a plurality of tubes, the plurality of tubes including at least one neutron absorbing tube and at least one neutron scattering tube,the at least one neutron absorbing tube configured to contain neutron absorbing material, the neutron absorbing material including at least one of boron material, carbide material, hafnium material, gadolinium material, or any combinations thereof, andthe at least one neutron scattering tube configured to contain neutron scattering material, the neutron scattering material including at least one of beryllium material, graphite material, europium material, or any combinations thereof, andat least one baffle plate arranged between an outer shell and an inner shell, the at least one baffle plate including a plurality of perforations, wherein each perforation of the plurality of perforations is configured to receive and support a corresponding tube of the plurality of tubes; anda plurality of motors attached to the plurality of drive shafts, at least one motor of the plurality of motors configured to rotate the at least one control drum such that the at least one neutron absorbing tube of the at least one control drum faces the plurality of nuclear fuel rods during a first state, and the at least one neutron scattering tube of the at least one control drum faces the plurality of nuclear fuel rods during a second state. 12. The nuclear reactor of claim 11, whereinthe plurality of perforations are arranged along at least one ring of the at least one baffle plate, the at least one ring including at least a first sector and a second sector, the first sector including a plurality of neutron absorbing tubes and the second sector including a plurality of neutron scattering tubes, the plurality of neutron absorbing tubes including the at least one neutron absorbing tube, and the plurality of neutron scattering tubes including the at least one neutron scattering tube. 13. The nuclear reactor of claim 11, whereinthe at least one baffle plate is a plurality of baffle plates; andthe plurality of baffle plates are sequentially arranged between the outer shell and the inner shell along a first direction of the at least one control drum. 14. The nuclear reactor of claim 11, whereineach perforation of the plurality of perforations includes at least one spring configured to allow for expansion of a corresponding supported tube. 15. The nuclear reactor of claim 11, whereinthe inner shell is configured to mate with the drive shaft via a magnetic coupling. 16. The nuclear reactor of claim 15, further comprising:at least one torsional spring attached to the inner shell, the at least one torsional spring configured to rotate the at least one control drum via the drive shaft such that the at least one neutron absorbing tube faces the plurality of nuclear fuel rods during a third state. 17. The nuclear reactor of claim 16, wherein the third state is a fail-safe state where at least one of the magnetic coupling or the at least one motor has failed. 18. The nuclear reactor of claim 11, whereinthe neutron absorbing material has a form of a powder, pellets, or a solid; andthe neutron scattering material has a form of a powder, pellets, or a solid. 19. The nuclear reactor of claim 18, wherein the plurality of control drums are horizontally mounted in a nuclear reactor core of the nuclear reactor. 20. The nuclear reactor of claim 18, whereinthe neutron absorbing material includes at least one of boron carbide, hafnium, gadolinium, or any combinations thereof; andthe neutron scattering material includes at least one of beryllium, graphite, europium, or any combinations thereof. |
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description | The present invention relates to an optimized laser cutting method for cutting a part from a material by means of a cutting system comprising a laser source for producing a laser beam having a certain power, and a cutting head comprising an end nozzle for the passage of the cutting laser beam. An application of lasers for cutting materials, particularly metals, is known. The cutting is created by evaporation and melting of the material subjected to the laser beam. A laser cutting system comprises in a conventional manner a head, comprising an internal chamber and an inlet of gas under pressure and filling the chamber before being ejected towards the material to be cut. The gas makes it possible to evacuate the material that was melted and vaporized by the laser beam. The molten material, which is in part oxidized, re-solidifies outside of the cutting. The following may be distinguished: slag adhering to the surface of the cut-out material, in other words solid particles constituted by ejection of the molten material from the cutting groove and compounds thereof formed by oxidoreduction as well as accumulation of impurities during the melting of the cut-out material; sedimented slag, in other words slag not adhering to the surface of the cut-out material and falling by gravity; aerosols, in other words solid particles formed by the projection and the vaporization of a part of the material from the cutting and the constitution of compounds formed by oxidoreduction during the melting of the cut material, but remaining in suspension in a surrounding gaseous environment, in other words having a negligible fall velocity. Adherent slag, being linked to the cut material, is collected together with the cut-out material. On the other hand, dispersion of sedimented slag makes the collection thereof difficult. It will be easily understood that when laser cutting is used for dismantling nuclear facilities, sedimented slag is even more undesirable, because it contains irradiated or contaminated materials, which are consequently radioactive. No optimized laser cutting method exists at the present time that makes it possible to limit the production of residues, particularly of sedimented slag. The invention proposes overcoming at least one of these drawbacks. To this end, according to the invention an optimized laser cutting method for cutting out a piece from a material by a cutting system is proposed and comprises a laser source for producing a laser beam having a power, and a cutting head comprising an end nozzle for the passage of the cutting laser beam,wherein the method is characterized in that it comprises a step of determining a cutting power Pd such that:Pd=Max(Popt;λe)where Max is the mathematical operator of the maximum, Popt is an optimal power of the laser beam of the cutting system, determined in accordance with the piece to be cut out, and/or with cutting parameters and/or with the type of system,in order to minimize the mass defect per unit length of the piece during cutting of said piece, λ is a leading coefficient representing the number of kW required for cutting out the piece per mm of piece thickness, and e is the thickness of the piece, in mm. The invention is advantageously completed by the following characteristics, taken alone or in any technically possible combination thereof: the cutting power Pd is of the form:Pd=Max(Λ;λe).where Λ is a predetermined constant, λ is a leading coefficient representing the number of kW required for cutting the piece per mm of piece thickness, and e is the thickness of the piece, in mm. for a cutting system comprising a laser source of the yttrium aluminium garnet YAG type for the production of a laser beam having a wavelength of the order of 1 μm, the power Pd is of the form:Pd=Max(4.75;0.1·e). the method comprises an initial step of determining an expression of the laser beam power in accordance with the piece to be cut-out, and/or with cutting parameters, and/or with the type of system, wherein according to this initial step: the system performs a plurality of test cuttings of a piece while varying the power of the beam, and/or the piece to be cut-out, and/or the cutting parameters and/or the type of system; a sensor performs a plurality of corresponding measurements of the mass defect during each test cutting of the piece, a computer expresses the mass defect per unit length during each test cutting of the piece in accordance with the power of the beam, and/or with the piece to be cut-out, and/or with cutting parameters and/or with the type of system; performs a partial derivative of the expression of the mass defect per unit length, with respect to the power of the laser beam, and determines (S4) the expression making it possible to cancel out said partial derivative in accordance with the piece to be cut-out, and/or with cutting parameters and/or with the type of system. The invention has numerous advantages. It makes it possible indeed to favour formation of adherent slag, instead of sedimented slag, to facilitate the evacuation of the cutting residues with the cut-out material. The dispersion of the sedimented slag in the surrounding environment is reduced. Indeed, the mass defect per unit length Mp is defined as the mass defect of the piece, after the cutting, per unit of cut-out length, expressed in g/m. Mp is obtained by weight difference before Mi and after Mf the cutting of length L: M P = M f - M i L . Mp takes into account both the loss of material by transformation of the metal into sedimented slag and into aerosols, but also a gain in weight by oxidation of the adherent slag. It will be understood however that by minimising the mass defect per unit length, the proportion of adherent slag is increased compared to sedimented slag. The reduction particularly of sedimented slag is advantageous in the case where laser cutting is used for dismantling nuclear facilities. The collection of wastes is thereby facilitated. On the other hand, the cutting is always carried out. In all of the figures, similar components bear identical reference signs. FIG. 1 illustrates a cutting system 10 mainly comprising a laser source 11 for producing a laser beam 111 having a power, for cutting out a piece 1 of a certain thickness e, and a cutting head 12 comprising an end nozzle 13 for the passage of the cutting laser beam 111. The piece 1 is represented flat, but can also have any profile, for example curved. The beam 111 must typically have a power of 1 kW per cm (10 mm) of thickness of piece to be cut out, particularly for stainless steel. The diameter of the nozzle 13 is referenced DB. It is in general 3 mm or 6 mm, but other values are obviously possible. The source 11 is connected to the head by two optic fibres. The first one, situated outside the worksite or the dismantling cell, is not contaminable, unlike the second one, situated inside the worksite. Both fibres are connected by means of an optical coupler 16. Each fibre is provided at each of the ends thereof with a connector that enables its dismantling: from the source, from the coupler and from the optic head. The replacement of the head 12 and the second fibre is thereby facilitated, notably when the head 12 or the fibre has been used for dismantling nuclear facilities, and is thus likely to be radioactive by contamination. The fibre 18 may be of any length, but is in general of a length of 100 m from the source 11 up to the coupler 16 (for a diameter of 400 μm for example), and of 20 m from the coupler to the head 12 (for a diameter of 600 μm for example). The cutting position of the beam 111 is preferentially at right angles with respect to the piece 1 to be cut out, for example horizontally. The distance H between the end of the nozzle 13 and the piece 1 may be of different values, but is in general comprised between 5 and 30 mm, as a function of the power of the beam 111 and of a piece thickness e. The head 12 may be moved along five axes (three translations and two rotations) to carry out the cutting, by means of an actuator 17 (typically a five axis robot or a remote manipulator). The movement of the head 12 takes place at a certain standard cutting speed V, during a cutting. It will be understood that a certain time is necessary to cut a thickness e of a piece 1, for a given power of the beam 111: if the head 12 moves too quickly with respect to the piece 1, the cutting is not performed correctly throughout the whole thickness e. The limit cutting speed VL designates the speed above which the piece 1 cannot be cut out. The limit speed VL is never, in practice, reached on a dismantling site, to ensure a safety margin and to guarantee the cutting of the piece. A coefficient k is thus defined, called limit speed coefficient of the cutting head 12, as the ratio between the standard cutting speed and the limit cutting speed of the part 1, i.e.: k = V V L . k takes in general the values of 0.5 or 0.7 on work sites, but can have any value below 1 according to the desired safety margin. In a conventional manner, the head 12 comprises an internal chamber 123 and an inlet 122, which is generally lateral, for pressurised gas 121 and for filling the chamber 123. The nozzle 123 also lets the pressurised gas 121 that surrounds the beam 111 escape. The gas 121 makes it possible to evacuate the material from the piece 1 that is melted and vaporised by the beam 111. The gas 121 is in general air. A gas passage flow rate is of the order of 400 L/min. The chamber 123 comprises a collimating lens 124 for the beam 111 and a focusing lens 125 for the beam 111. The focal length of the collimating lens 124 may be 80 mm for example, and the focal length of the focusing lens 125 may be 253 mm for example. During the cutting of the piece 1 by the beam 111, residues are created, particularly aerosols 20 and sedimented slag 21, which creates a mass defect Mp per unit length. General Principle of the Invention The invention makes it possible to minimise the Mp defect during a cutting of the piece 1, by determining an optimal power Popt of the laser beam 111 of the cutting system 10. To this end, the optimal power Popt of the laser beam 111 of the cutting system 10 is expressed, in an expression, in accordance with the piece 1 to be cut out, and/or with cutting parameters, and/or with the type of system 10,in order to minimize the mass defect per unit length during a cutting of the piece 1. The minimisation of the mass defect per unit length will mainly maximise the production of adherent slag. The piece 1 is for example represented in the aforementioned expression by the value e of the piece thickness, but could for example also be represented by another parameter representative of the piece, such as the type of material and/or alloy making up the piece (304 or 316 L for example). The cutting parameters in the aforementioned expression are for example the distance H between the nozzle 13 and the piece 1, or the coefficient k, or also for example the impact diameter of the beam. The type of system is for example represented in the aforementioned expression by the diameter DB, linked to the flow rate of gas 121, but other parameters could also be taken into account, such as the flow rate of gas, the nature of the gas, the cut-off pressure at the surface of the piece 1, or the type of optic fibre. The cut-off pressure is linked to the kinetic energy of the gas 121 and results from the flow rate of gas 121, the diameter DB and the geometry of the nozzle 13, as well as the distance H. A minimal cut-off pressure is required to ensure the cutting. Thus, for an impact diameter of the laser beam on the piece comprised between 2 and 4 mm, a pressure of 0.08 bars is required to cut out 10 mm of stainless steel. In practice, 0.25 bars is a value that ensures a satisfactory robustness for piece thicknesses below 80 mm. A pressure of 0.8 bars enables stainless steel pieces up to 100 mm to be cut-out. A higher cut-off pressure admittedly improves the productivity of the cutting, by providing a higher limit speed for the same power (a pressure of 0.8 bars thus makes it possible to double the limit cutting speed obtained with the impact pressure of 0.25 bars for thicknesses comprised between 20 and 80 mm), but it reduces the proportion of adherent slag to the profit of sedimented slag by favouring the evacuation of the molten material and slag that form inside the cutting edge. The value of 0.25 is thus recommended for thicknesses below 80 mm and the value of 0.8 bar is recommended for thicknesses from 80 to 100 mm or from 20 to 80 mm if enhanced productivity is required. General Principle of Determining the Power In order to be able to determine the power Popt, it will be understood that it is necessary to go through an initial step of determining the expression of the power in accordance with the piece 1 to be cut out, and/or with cutting parameters, and/or with the type of system 10. According to the initial step of determination represented in a schematic manner in FIG. 2, the system 10 carries out, during a step S1, a plurality of test cuttings of a piece 1 while varying the power of the beam 111, and/or the piece 1 to be cut out, and/or the cutting parameters and/or the type of system 10. During step S1, the system 10 thus performs cuttings with for example: different power values of the beam 111, and/or different values of thicknesses e of piece 1, and/or different values of k, and/or different values of H, and/or different values of DB. During a step S2, a sensor 14 performs a plurality of corresponding measurements of the Mp defect during each test cutting of the piece 1. The sensor 14 used during step S2 comprises in particular a balance making it possible to weigh the piece 1 before and after the cutting. A computer 15 known to those skilled in the art, comprising all conventional memory and processing means, and connected to the sensor 14, makes it possible to plot the curves of FIG. 3 representing in a general manner the evolution of Mp as a function of the power of the beam 111. The inventors have noted that the curves of FIG. 3 seem to indicate that the mass defect per unit length is minimal for a certain power, called Popt in the present description. Steps S3 and S4 thus make it possible to determine Popt as a function of the measurements of the sensor 14. To this end, during step S3, the computer 15 expresses Mp during each test cutting of the part 1 in accordance with the power of the beam 111, and/or with the piece 1 to be cut, and/or with cutting parameters and/or with the type of system 10. If the parameters taken into account are the power P, the thickness e, the coefficient k, the distance H and the nozzle diameter DB, an expression of the following type is obtained by means of the computer 15:MP=f(P,e,k,H,DB) (E1) To find the expression E1, the computer 15 thus performs the construction of a mathematical model as a function of the measurements stemming from the sensor 14, using a plurality of regressions, for example linear, logarithmic, square, or other, and keeps the expression giving the model closest to the measurements, in other words the model gives a known mathematical correlation coefficient R2, between the measures and the values given by the model, such as for example:R2>0.9.Since the curves of FIG. 3 show that Mp as a function of the power comprises a minimum, the computer 15 finds this minimum by cancelling out, during step S4, a partial derivative of the expression (E1) of Mp with respect to the power of the laser beam. It is then known that the cancellation of the partial derivative corresponds to the optimum of P. The computer 15 thus firstly performs: ∂ M P ∂ P = ∂ ∂ P f ( P , e , k , H , DB ) ( E2 ) The computer 15 then also determines during step S4 the expression making it possible to cancel out said partial derivative in accordance with the piece 1 to be cut, and/or with cutting parameters and/or with the type of system 10. The computer 15 thus performs: ∂ M P ∂ P = 0to find the expression of P minimizing the mass defect per unit length.Example of Determination During step S1, the system 10 performs a plurality of test cuttings of a piece 1 of thickness e, with a diameter DB of the nozzle 13 and a limit speed coefficient k of the cutting head 12. The system 10 used comprises a laser source 11 of the yttrium aluminium garnet YAG type, for example with disc, capable of producing a laser beam 111 having a wavelength of the order of 1 μm. During cuttings, the power P of the laser beam 111 is variable from 1 to 8 kW (powers of 3 kW, 5 kW and 8 kW are for example used), taking account of the fact that it is necessary, to have around 1 kW for each cm (10 mm) of thickness of the piece for the cutting to be effective. The flow rate of gas 121 has been maintained constant and equal to 400 L/min during the plurality of cuttings. The piece 1 to be cut out is a piece made of 316 L stainless steel (reference AFNOR standard: X2 Cr Ni Mo 18-10 1.4404) with thickness e, representative of the constituent components of a nuclear facility to dismantle. During test cuttings, the thickness e varies for example from 10 mm to 80 mm. During test cuttings, the diameters DB take the values 3 mm or 6 mm. The values 0.1; 0.25; 0.5 and 0.7 are taken for k. By way of information, the values of VL are reproduced in the following tables 1 and 2, for a 316L stainless steel piece—H 30 mm—Flow rate 400 L/min: limit speed (mm/min). TABLE 1limit cutting speeds for the nozzle of DB 3 mm.eVLVLVLVLVL(mm)8 KW6 KW4 KW3 KW2 KW1007.580206040204012575203020012575402045030017510020101200900600400200 TABLE 2limit cutting speeds for the nozzle of DB 6 mm.eVLVLVLVLVL(mm)8 KW6 KW4 KW3 KW2 KW100801060301540756010301751255020203502501257010101000800600400200 Table 3 hereafter reproduces the measures of the sensor 14 concerning the mass defect per unit length of the piece 1 (final column). TABLE 3MasseSurfaceMass perMassPlateper unitmass ofunit lengthVolumedefect ofNozzleLaserthick-DistanceCuttinglength ofaerosolH2adherentthe platediameterpowernessplaque-speedICICICslagICN°DBPeHkVRM,(M,)m2(M2)[H2]([H2])VMp(Mp)planmmkWmmmm—m/ming/mg/mg/m2g/m2mg/mmg/mcm3/mg/mg/m032.110300.500.1092.260.229.01.916.7110213310300.250.1093.550.253592511.73.116.9153426310150.480.1791.270.0912894.61.935.164433810150.500.4774.100.284152913.62.42.3179446810300.250.2492.230.16226167.92.316.5169446810300.250.2492.250.16228167.52.224.8193446810300.250.2492.390.17242174.52.026.2185446810300.250.2492.340.16237174.72.0171453320300.490.04412.50.96114231.66.017.5292753320300.490.04410.90.85343724.26.737.22751066320150.250.01512.20.9595435.26.291.91062073820150.240.1098.840.884324319.85.820.93931086820300.500.1992.900.20141106.22.9126.0276686820300.500.1993.410.24167125.02.9148.5247693830300.250.05615.71.15043713.78.358.058622103860300.240.01032.43.35335527.919.3133642113830300.490.10910.60.73402420.96.115.369211113830300.490.10910.00.73222320.05.522.166210123860300.500.02133.82.45564067.314.9135.392522133880300.570.01040.42.95043859.722.5157645143530300.700.0708.410.592701916.26.185.439010153830300.710.1596.470.372081219.76.511.065711173860300.710.03032.51.95343162.715.099.0116224186820300.090.0359.640.69471344.05.8104.57919196810300.500.4971.380.10140102.91.848.71054206310300.500.1890.550.045642.81.836.5534216310100.110.0404.910.34497353.24.318.1829226310300.110.0402.860.21290210.043.35310233860150.240.01043.73.27195527.619.012594325 6*31000.110.0404.750.33481344.54.9479 During step S2, the computer 15 constructs the curves of FIG. 3, according to the measurements of table 3 stemming from the sensor 14 for the cuttings of step S1. In FIG. 3, the black curve represents the upper limit of the validity domain of the curves, to take account of the constraint of 1 kW of power per cm (10 mm) of thickness of piece to be cut. During step S3, the computer 15 expresses the mass defect during each test cutting of the piece 1 as a function of the power P of the beam 111, the thickness e of the piece 1 to be cut out, the distance H, the coefficient k and the nozzle diameter DB of the system 10. The computer 15 uses a linear regression from the measures of step S2 to give the best model. In our example, the equation (E1) is for example expressed in the following manner, from the measurements of table 3:MP=237−93.2·(DB−4.40)+46.6·(P−6.20)+11.9·(e−24.7)−5.26·(DB−4.40)·(e−24.7)+16.0·(P−6.20)2 (E1) The correlation coefficient of (E1) with respect to the measures of table 3 isR2=0.9724. The equation (E2) corresponding to the partial derivative of the expression of (E1) is thus ∂ ∂ P M P = [ 46.6 + 32 P - 198.4 ] = [ 32 P - 151.8 ] ( E2 ) The computer 15 then determines the expression making it possible to cancel out said partial derivative, in accordance with the piece 1 to be cut out, with cutting parameters and with the type of system 10. By putting down: ∂ ∂ P M P = 0the computer 15 thus finds a constant Λ such that: Popt = 151.8 32 = 4.75 ( E3 ) Thus at step S5, the cutting power Pd is determined preferably such that it is equal to Popt, which minimises the mass defect per unit length during a cutting of the piece 1, while ensuring that the power is at least equal to the cutting power (it will be recalled that it is necessary to have around 1 kW for each cm of thickness of the piece for the cutting). A minimal power Pmin=λ·e may thus be defined, with λ representing the number of kW required for cutting the piece per mm of thickness of the piece 1 (in our example 0.1), and e the thickness of the piece, in mm. If Popt is such thatPopt<Pminthen Pmin will be taken for Pd. On the other hand, if Popt is such thatPopt>Pminthen Popt will be taken for Pd. |
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summary | ||
abstract | An apparatus and related method for supporting X-ray imaging. The apparatus comprises an input interface (IN) for receiving an X-radiation scatter measurement obtained by an X-ray sensor (SXi) during operation of an X-ray imager (XI) for imaging a first object (PAT). A predictor component (PC) is configured to predict, based on said measurement, whether or not: i) a second object (P) is present, or ii) there is sufficient X-ray exposure of said first object (PAT). The apparatus comprises an output interface (OUT) for outputting a predictor signal indicative of an outcome of said prediction. |
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description | The present invention relates to a charged particle microscope and a method for correcting measurement images by utilizing the charged particle microscope. The charged particle microscope is widely used for observing the structure of substances at a high magnification. However, drift can sometimes occur due to the characteristics of the specimen and the equipment stage. Charged particle microscopes generally shift the imaging field of view by moving a stage carrying a specimen placed on a sampling stage. However, due to problems with mechanical precision, the stage does not suddenly stop even if stop operation was initiated and still continues to moves even though only a small distance. Drift is caused mainly by the slight movement of the sampling stage after stop operation. In the charged particle microscope, the capture of the image for observation requires a long time ranging from a few seconds to several dozen seconds and moreover is imaging that is enlarged to a high magnification so that even just a slight amount of drift causes distortion to appear in the image. However, finding what extent of image displacement has occurred due to drift, or finding in what direction the displacement amount occurred, just in the image where distortion occurred was impossible (in the related art) so preventing distortion from entering the image at the time of measurement or some type of method for correcting distortion in the image is needed. One way to prevent distortion from entering an image during measurement, is to start the observation after waiting for the sampling stage to come to a complete stop after operation to stop the sampling stage movement however the image capture efficiency in that case is extremely poor. In order to resolve the problem, a variety of methods to correct image distortion due to drift were contrived. The patent literature 1 for example discloses a drift correction method to correct slow-scan images by utilizing results from finding the drift amount (drift speed) per unit of time in the X direction and the Y direction from two fast-scan images (television scanning image) in order to find the displacement amount, due to drift in the image captured by slow-scan. Patent literature 1: Domestic Re-publication of PCT International Application WO2003/004821 (U.S. Pat. No. 703,296) The method that starts measurement after waiting for movement of the sampling stage to stop requires waiting for the sampling stage to come to a complete stop whenever the sampling stage moves and so there is a drastic drop in operability during observation. The drift correction method disclosed in patent literature 1 on the other hand, corrects one image from among at least three images, however the number of captured images relates to the direct device throughput and damage to the specimen and so a drift correction method is needed that achieves the drift correction from as few images as possible. The present invention has the object of providing drift correction from as few captured images as possible. In order to resolve the aforementioned problems, a reference image is measured in order to correct distortion at a position and magnification identical to the image acquired for observation. The reference image is at this time measured within a shorter time than the actual observation image in order to lessen the effects of drift. The reference image performs the measurement in a short time compared to the observation image so the signal volume decreases and the reference image cannot be used for making observations. However the reference image has little distortion compared to the observation image and so correctly reflects the shape of the specimen. The shape of the observation image is corrected by comparing the shapes of the reference image and observation image, and then correcting the shape of the observation image to match the shape of the reference image. The present invention does not require waiting until the sampling stage has fully stopped before starting observation. Moreover, the correction of the related art required the acquisition of three images however the present invention can perform correction with two images. The embodiments of the present invention are described next while referring to the drawings. FIG. 1 is a drawing showing the structure of the device for achieving the present invention. A charged particle microscope 101 is a microscope utilizing charged particles and that obtains image at a high magnification by irradiating charged particles onto the specimen. Generally known microscopes that use an electron beam are the scanning electron microscope and transmission electron microscope. Images acquired by the charged particle microscope 101 are obtained by the image input device 102. Images acquired by the image input device 102 can be stored in the storage device 105. The images measured by using the charged particle microscope 101 and the image input device 102 are corrected on the arithmetic logic unit (ALU) 104 based on correction conditions input by the correction condition input device 103, and the images are output from the image output device 106. FIG. 2 is images showing the content implemented by the present invention. An observation image 201 is an image measured for observation purposes using the charged particle microscope 101. The measurement is made in a period ranging from a few to several dozen seconds in order to obtain high image quality and so contains much distortion. The correction reference image 202 like the observation image 201 is an image measured by using the charged particle microscope 101. The correction reference image 202 measures faster than the observation image 201 in order to alleviate the effects of drift in the sampling stage. The observation image 201 for example measures at 40 seconds and so when the correction reference image 202 measures at 40 milliseconds the effect of drift in the sampling stage is 1/1000th. A corrected observation image 203 is acquired by comparing the shape of the observation image 201 with the correction reference image 202 and making corrections. The sampling stage drift movement is for here the case where moving in the lateral direction and where moving in the vertical direction, and moreover in the diagonal direction utilizing the lateral and vertical components. FIG. 3 is illustrations for describing the correction of lateral drift. A profile position 301 is set along the line for verifying the shape of the image at a desired location laterally on the observation image 201. A profile position 302 is also set on the correction reference image 202 at the same position as the profile position 301 set on the observation image. The profile 303 is an image profile for the profile position 301 that was set on the observation image 201. The profile 304 is an image profile for the profile position 302 that was set on the correction reference image 202. The amount of movement in the lateral direction can be found by comparing the shape of these two profiles. The drift amount for the entire image can be obtained by sequentially detecting the drift amount on each line while shifting the position where the profile was set from the top edge to the bottom edge of the image. FIG. 4 is a drawing showing the drift amount in the lateral direction along the entire image. The actual measured drift amount 402 contains an error due to effects of noise and fluctuations in the shape during measurement, etc. However, the drift amount continuously fluctuates when the cause of the drift is established as the drift from the time after the stop of sampling stage stop operation until, the sampling stage stops. Variations due to the effects of noise and so on can be alleviated by calculating the approximate curve from the measured drift amount. The method for calculating the approximate curve is the least squares approximation. FIG. 5 is illustrations showing the method for correcting the drift in the lateral direction. In the observation image 201, a lateral drift-corrected image 501 can be obtained by shifting laterally one line at a time in the lateral direction according to the approximation curve of the drift amount found from FIG. 4. FIG. 6 is a drawing for describing the method for correcting the drift amount in the vertical direction. The lateral drift amount is featured in being measured as the image displacement however vertical drift appears as extensions and contractions in the image. Methods that simply compare profiles on a line therefore cannot detect the vertical drift amount. The vertical drift amount is therefore calculated by measuring what section in the correction reference image 202 matches a sectional area within the observation image. In this method, a reference region 601 is first of all set at an optional position in the lateral drift-corrected image 501. A correction position search region 602 is next set at a position identical to the reference region 601 that was set in the lateral drift-corrected image 501 on the correction reference image 202. The vertical drift amount can next be calculated by detecting a position having the same shape as the reference region 601 from the correction position search region 602. The correction position search region 602 is however set to a region wider than the reference region 601 in order to detect a section having the same shape as the reference region 601 within the correction search region 602. A method to search for an identical shape is widely known as template matching using a normalized correlation. The drift amount for the entire image can be calculated by calculating the drift amount while shifting the reference region 601 and correction position search region 602 up and down. In the vertical drift amount, effects from noise and so on may sometimes cause variations in the drift amount the same as with the lateral drift amount so an approximate curve is also found for the vertical drift amount. FIG. 7 shows the method for correcting the vertical drift amount. In the lateral drift-corrected image 501, the vertical drift amount can be performed by shifting the image vertically according to the approximate curve for the drift amount found in FIG. 6. However, drift in the vertical direction cannot be corrected just by merely shifting and copying the vertical line. Such correction is not possible because the drift amount in the vertical direction appears on the image as extensions and contractions in the shape. A corrected observation image 203 can be obtained by calculating which position in the lateral drift-corrected image 501 matches which of the respective pixels in the corrected observation image 203 according to the approximate curve for the drift amount found in FIG. 6; and calculating the value of each pixel in the corrected observation image 203 from performing interpolation calculation on the lateral drift-corrected image 501. The areas (profile position 301, 302, reference region 601, correction position search region 602) for detecting the drift amount and method for finding the correction curve from the drift amount must be set in order to calculate the lateral and vertical drift amounts. One way to set these areas and method is to establish fixed areas and a method beforehand. Moreover, a method to find the area for detecting the drift amount and find the correction curve from the detected drift amount can be input by the correction condition input device 103. FIG. 8 shows the processing flow of the present invention. An acquire reference image 801 obtains the correction reference image 202. An acquire observation image 802 obtains the observation image 201. The lateral drift amount is then calculated by the calculate lateral drift amount 803 using the correction reference image 202 obtained by the acquire reference image 801 and the observation image 201 obtained by the acquire observation image 802. A find approximate lateral drift amount 804 sets an approximate lateral drift amount calculated by calculate lateral drift amount 803 and calculates an approximate curve for the lateral drift amount. In the correct the lateral drift correction 805, the lateral drift amount is corrected by using the approximate curve for the lateral drift amount and the observation image 201, and acquires the lateral drift-corrected image 501. A calculate the vertical drift amount 806 calculates the vertical drift amount by using the lateral drift-corrected image 501 corrected in the lateral drift correction 805 and the correction reference image 202 obtained in acquire reference image 801. The find approximate vertical drift amount 807 approximates the vertical drift amount calculated by vertical drift amount 806, and calculates the approximate curve of the vertical drift amount. The correct the vertical drift 808 corrects the vertical drift amount by utilizing the lateral drift-corrected image 501 and the approximate curve for the vertical drift amount, and obtains the corrected observation image 203. The second embodiment of the present invention is described next using FIG. 9. In addition to the methods described so far, user support is further provided by way of the operating screen. A warning screen 901 contains a warning character string 904, a first button 902, and a second button 903. If drift was detected in the image then the warning screen 901 is displayed to the user to show that the captured image contains drift. The user at this same time selects on the warning screen 901 whether or not to execute drift correction of the captured image. To make the selection, the user clicks a first button 902 or a second button 903 to permit or abort drift correction. 101 Charged particle microscope 102 Image input device 103 Correction condition input device 104 Arithmetic logic unit (ALU) 105 Storage device 106 Image output device 201 Observation image 202 Correction reference image 203 Corrected observation image 301 Observation image profile position 302 Correction reference image profile position 303 Observation image profile 304 Correction reference image profile 401 Lateral drift amount graph 402 Lateral drift measurement amount 403 Lateral drift amount approximate curve 501 Lateral drift-corrected image 601 Reference region 602 Correction position search region 603 Vertical drift amount graph 801 Acquire reference image 802 Acquire observation image 803 Calculate lateral drift amount 804 Find approximate lateral drift amount 805 Correct the lateral drift 806 Calculate the vertical drift amount 807 Find approximate vertical drift amount 808 Correct the vertical drift 901 Warning screen 902 First button 903 Second button 904 Character string |
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abstract | A system for storing and/or transporting high level radioactive waste, and a method of manufacturing the same. In one aspect, the invention is a ventilated vertical overpack (“VVO”) having specially designed inlet ducts that refract radiation back into the storage cavity. A clear line-of-sight does not exist through the inlet ducts and, thus, the canister can be supported on the floor of the VVO. Also disclosed is a method of manufacturing a variable height VVO that falls within a regulatory license previously obtained for a shorter and taller version of the VVO. |
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050892207 | summary | BACKGROUND OF THE INVENTION The present invention relates to a fuel assembly for a boiling reactor. The fuel assembly comprises a bundle of elongated fuel rods retained by a number of so-called spacers placed with a certain distance between each other along the bundle. A coolant, for example water, is adapted to flow from below and upwards through the fuel assembly which normally is arranged vertically and, upon a nuclear reaction, to cool the fuel rods arranged in the fuel assembly. The object of the invention is to increase the efficiency of this cooling of the fuel rods. In a boiling type nuclear reactor the steam formation in the fuel assembly increases more and more towards the upper part of the assembly, as is clear from FIG. 1 which shows, in rough outline, a cross section of part of a fuel assembly. In FIG. 1, 1 designates a fuel rod and 2 spaces between the rods. This space 2 is in the lower part of the fuel assembly (corresponding to the lower part of the core of the reactor), filled with coolant, in this case water. Further up in the fuel assembly, steam bubbles 3 are formed in the water which, still further up, is transformed into water steam in the region 4. As long as so-called dry out does not take place, however, there is always a film 5 of the cooling water on the fuel rods. It is important that this film 5 is maintained at all points of the rods 1. If at some point it disappears by dry out, serious damage at this point of the fuel rod 1 will rapidly arise. In FIG. 1, 6 designates the wall of the fuel assembly. Also this is normally coated with a water film 5. However, this film 5 is not entirely necessary since the wall 6 of the assembly is considerably more insensitive to superheating compared with the fuel rods. This fact has been observed and attempts have been made to make use of it in some known designs, as, for example, in U.S. Pat. No. 4,749,543, column 8 and FIG. 9. In these designs, the cooling water flowing along the wall 6 of the fuel assembly is diverted towards the centre of the bundle by means of elevations on the wall 6 or recesses in the same. Also fins on the downstream side of the spacers are used to achieve a diversion or deflection of the cooling water. All these embodiments have certain drawbacks. Thus, for example, the elevations may increase the pressure drop in the cooling water and thus reduce the cooling effect, whereas recesses in the wall entail certain difficulties from the point of view of manufacturing technique. Further, a deflection of the cooling water flowing along the assembly wall 6 should take place as early as possible in relation to each separate spacer and, in any case, preferably not immediately after the same viewed in the direction of flow. This is due to the fact that dry outs normally occur immediately upstream of a spacer or possibly in the same. SUMMARY OF THE INVENTION The present invention relates to a device, in a spacer (13) known, inter alia, from Swedish patent 8601982-5 (see FIG. 1 of that patent), for achieving the desired deflection in a simple manner. According to the invention, the spacer is equipped with deflection fins arranged in the windows of the spacer. |
056028857 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to the field of optical inspection systems for assessing the quality of welds. In particular, the invention concerns an automated part handling and image processing system responsive to sharply localized variations in reflectance, for selecting, rejecting and/or monitoring the quality of girth welds on nuclear fuel tubes. 2. Prior Art Nuclear fuel is packaged in long narrow stacks of enriched uranium pellets carried in zirconium alloy tubes. The tubes are plugged at the ends and sealed, and in addition to the pellets, contain an inert gas. A number of such tubes are mounted parallel to one another in a fuel assembly having supporting grids spaced along the tubes, with receptacles holding the tubes parallel to one another and at a space sufficient to admit control rods that are movable into spaces between the tubes for damping nuclear flux. Zirconium is similar to aluminum in that it is a malleable metal that develops an oxide film or coating on surfaces exposed to oxygen, the coating generally forming a barrier that protects the metal underneath. When subjected to heating in oxygen or in water (for example when used in a reactor for generating heat), the oxide coating thickens and the tube becomes black, the rate of oxidation being related to the temperature and time of heating and the availability of oxygen. Zirconium alloy (e.g., Zircalloy) is a preferred material for fuel rods because it presents a low cross section to nuclear particles such as neutrons and therefore does not in turn become a strong source of nuclear radiation when it is irradiated. Oxidation of the fuel rod tubes is advantageous because a thick oxide coating protects the tubes, giving them a harder surface. Reactor coolant pumps and convection currents in an operating reactor produce a powerful and turbulent flow of coolant in pressurized water reactors and boiling water reactors. The flow can carry along pieces of metal and the like, which can impact against the fuel rod tubes. It is important to avoid a breach in the walls of the fuel rod tubes, which can lead to release of radioactive material into the coolant. Oxidized tubes are harder and less subject to fretting damage from debris carried along in the reactor coolant. The tubes are particularly vulnerable to fretting damage when they are new and relatively unoxidized. The tubes are also vulnerable on their surfaces immediately adjacent to the supporting grids of the fuel assembly on the upstream side relative to coolant flow. Loose debris in the coolant can be caught by the supporting grids and fretted against the tubes in this area. Such fretting damage to the tubes has been found to occur most often on the upstream side of the foremost supporting grid of the fuel assembly, which is the lowermost grid due to the upward flow path of the coolant, namely at the lower end of the fuel rod tube. To combat the danger of fretting damage at the upstream end of the tubes, it has been proposed to treat the tubes preliminarily at their upstream ends, for example along the lowermost six or eight inches (15-20 cm), to protect the tubes from fretting damage. U.S. Pat. No. 5,171,520 --Bryan et al proposes to coat the tubes with zircon, a hard refractory material (ZrSiO.sub.4) also known as zirconium silicate. U.S. Pat. No. 5,265,137 --Busch proposes to treat at least the ends of tubes by heating them with one or more of carbon, nitrogen and oxygen to form a protective layer. During production, an end plug is welded to the fuel rod tube, which can be an automated process as disclosed in U.S. Pat. 4,857,260 --Schoenig, Jr. et al. The fuel pellets are loaded, together with an inert gas, and the opposite end plug is welded in place. The attachment of the end plugs to the tube is made along a circumferential line and the weld is termed a girth weld. Typically, the fuel rods are visually inspected for quality, including for the integrity of the girth welds and other aspects such as dimensions. Insofar as the tubes or the ends of the tubes are treated for surface hardness by oxidizing and thereby blackening the tubes, or by applying a protective coating, the girth welds are obscured. It is possible to inspect the tube ends using X-rays, which are not sensitive to the appearance of the surface. Such a technique is disclosed in U.S. Pat. No. 4,957,691 --Brashier et al. For optimal visual inspection, it is appropriate to inspect the tubes before such treatments, and preferably immediately before, so that the potential for handling damage between inspection and treatment is minimal. It would be advantageous to automate the production of fuel rods as much as practicable, including the inspection of the tubes. Automated handling and processing steps, however, are generally inconsistent with visual inspection for dimensions, defects, weld quality and the like. The present invention seeks to provide an automated technique for optical inspection of girth welds which can be accomplished with minimal handling steps using a compact apparatus that rotates the tube on its longitudinal axis while pixel data is collected by a line scan camera triggered by a shaft angle encoder coupled to the tube rotation drive. Each scan line is digitized, obtaining a line of pixel data over a longitudinal length encompassing the girth weld, and the scans are collected at equally spaced angular positions around the tube. The invention further seeks to analyze reflectance data on the tube in the area of the weld, in a manner that amplifies the system's reaction to local flaws that are equal to or larger than a predetermined minimum, by determining the average value of a number of pixels and counting the number of adjacent pixels, in two mutually perpendicular directions, which exceed the average reflectance value by a predetermined proportion (i.e., profiling defects parallel to the axis of the tube and along a circumference). The number of pixels and their pitch or spacing corresponds to the minimum flaw size, profiled in the X and Y directions in the collected matrix of pixel data. The image is thereby analyzed for detection missing or gapped welds, automatically determining the quality of the plug/tube seam. SUMMARY OF THE INVENTION It is an object of the invention to facilitate the automatic inspection of plug-to-tube girth welds in nuclear fuel rods, using a compact apparatus that is readily integrated into an automatic manufacturing process. It is also an object of the invention m reduce quality assurance reliance on visual manual inspection of fuel rods, particularly such that inspection can be conducted prior to protective process steps that would obscure defects in a girth weld. It is another object of the invention to discriminate for local defects in girth welds using an averaging and X-Y profiling technique that emphasizes sensitivity to optically detectable variations equal to or greater than a predetermined size in one or both perpendicular directions. It is also an object of the invention to examine a girth weld with sensitivity to variations around a circumference, while rotating the robe to collect pixel data in circumferential lines or slices. It is still another object of the invention to detect missing welds or gaps. These and other objects are accomplished by a method and apparatus for the inspection of girth welds joining the end plugs to the tubes of nuclear fuel rods. The welds are inspected automatically using a technique that sums reflectance values to examine a predetermined number of pixels in one direction (e.g., along the tube axis) to test for minimum defect width, and then perpendicularly to test for minimum defect length (around the circumference), whereby defects at least as large as the minimum defect size are emphasized and smaller defects are de-emphasized. The minimum defect size can be different in the two perpendicular directions, for example with the maximum acceptable reflectance variation spanning three pixels in width and ten pixels in length. The plug is welded to the tube along a circumferential girth weld in a plane perpendicular m a longitudinal axis of the tube. The girth weld area is illuminated and rotated for at least one revolution at an inspection station, for example prior to oxidation treatments of the tube that may conceal defects. A line scan camera coupled to a digitizer collects repetitive scans during rotation, preferably scanning parallel to the longitudinal axis of the tube and being triggered by a shaft angle encoder at regular angular positions. A matrix of reflectance data is thus collected at and adjacent to the girth weld, including numeric reflectance values of rows of pixels along a longitudinal length of the tube spanning the weld, and columns of pixels along circumferences of the tube. A processor sums the predetermined number of pixels in columns (i.e., circumferential segments) for a measure of the average reflectance over an area the size of the minimum defect size and compares the sum or average to a selection standard for accepting or rejecting the tubes and welds. Adjacent pixels that exceed the average reflectance value of the matrix by a predetermined proportion (e.g., 75%) are counted in the X and Y directions. The standard deviation of the average and other factors can also be examined. Preferably, the entire matrix of pixel data is averaged to calculate the reflectance setpoints. The calculations also can be made in a running average manner contemporaneously with collection of the reflectance values and with rotation of the tube. |
description | This invention in general is related to atomic cells and nuclear batteries. Prior art atomic cells and nuclear batteries are limited because they generate low currents. Another drawback is that expensive radioisotopes are obtained from a nuclear reactor in their construction. The present invention overcomes the aforementioned limitations by utilizing an alpha fusion reaction and radon emissive material. The alpha fusion reaction economically generates high power densities. The present invention provides a unique concept that offers improved performance over prior art direct nuclear conversion systems. The new and novel invention that will be described utilizes an alpha fusion reaction that generates practical and useful electrical current. Devices that convert ionizing energy to electrical current have been used in prior art, but with poor results. Atomic cells generate electric currents by utilizing charged particles that are ejected from radioactive substances. The Direct Conversion of Energy was published by the GPO in 1964. On pages 28-29 William R. Corliss discusses the direct use of charged particles that are ejected from radioisotopes. He states that high velocity beta particles ejected from 38Sr90 generates a flow of electrical current. The negative charges on the particles become neutralized when they strike a metallic cylinder. The neutralized particles find their way back to the 38Sr90 becoming again ionized. This cycle repeats itself so long as the 38Sr90 remains radioactive. U.S. Pat. No. 2,926,268 describes a self-powered electron tube that generates secondary electrons when high-energy radiations, primarily from beta particles strike a semi-conductive material. The power generated by the above two sited examples generate high-voltage but produce extremely low amperage. There are numerous patents issued world wide relating to the direct conversion of charged atomic particles that generates electrical current but all produce low power densities in the millionth of a watt range. Despite the prior art that exists in this technology, it is believed that there has not previously existed a small, compact electrical device capable of generating a high power output. It is the object of this invention to provide a method embodying a new and novel device to furnish an efficient and economical source of electrical power. The present invention resolves limitations of the prior art. The primary object of the present invention is to provide a method that directly utilizes charged particles to produce electrical current, and a new and novel device for utilizing an alpha-fusion nuclear reaction to generate the charged particles. The present invention relates to a method that generates electrons which can be converted to electrical energy and more particularly, to electrical power generation through the fusion of alpha particles with carefully chosen target elements, compounds, or alloys. The present invention may serve as a source of electrical current that is consistent a full 24 hours per day which is safe and non-polluting. The present invention is an original approach to the generation of electrical current, which relies upon an alpha fusion reaction. It is the main object of the present invention to provide a method and device for generating electrical energy that result from the reaction of alpha particles with specific materials. It is generally accepted that helium gas will not form compounds in any chemical combination. This gas generally is believed to be chemically inert. What is not readily realized is that helium will react with a few substances when sufficiently excited. It is a well-established fact; helium is a gas that accompanies all radioactive minerals in an excited state. The name for a high-energy helium atom is called an “alpha particle” in the scientific literature. Until now, its role in nuclear transformations has not been fully realized. The quantity of energy that is released under certain conditions is considerable. This conclusion was reached by the early scientific community because the small amount of ejected particles coming from radioactive matter possesses an enormous velocity, carrying with them enormous amounts of energy. The alpha particle reaction is a liberator of an enormous reserve of stored atomic energy. An example of an alpha fusion reaction can be demonstrated by depositing radon gas onto a beryllium wire. The resulting reaction was used to generate neutrons in the early days of atomic energy to initiate a fission reaction using fissile 92U235. The reaction is expressed in the following equations;4Be9+2He4 (high energy alpha particle)→6C12 0n1 (fast neutron)4Be9+0n1 (fast neutron)→4Be8+20n1 4Be8→22He4 (high energy alpha particles) In these equations, beryllium reacts with an excited alpha particle generating a fusion reaction with neutrons as its by-product. Enrico Fermi describes this reaction in his U.S. Pat. No. 2,206,634 Process for the Production of Radioactive Substances. The atoms are not fragmented in the above expressed reaction as is the case when a fission reaction is created. A fusion reaction can produce non-radioactive stable by-products, along with a supply of useful electrons, unlike a fission reaction that creates a number of radioactive deadly waste products. In the present invention a germanium plated, negatively charged corona cathode wire or thin rod, used in conjunction with a palladium or graphite positively charged anode concentric cylinder, can be utilized in its construction. Other materials can be used and this will not depart from the spirit of the present invention. Germanium used as a target material is a good choice because 32Ge72 will react with alpha particles generating stable 34Se77 and high-energy electrons within the process, in which:32Ge72+2He4 (high energy alpha)→34Se75+0n1 34Se75 EC (e− capture)→33As75 stable32Ge76+0n1→32Ge77 32Ge77→33As77+beta (high energy electron)33As77→34Se77 stable+beta (high-energy electron) It takes at least 6.06 MeV of energy to generate a 32Ge72 alpha fusion reaction. Alpha particles are ejected from Po212 with the energy release of 8.78 MeV, Po214 with the energy release of 7.68 MeV, and Po216 with the energy release of 6.78 MeV; these elements can be used to generate 32Ge72 alpha fusion reactions. Therefore, Po218 with the energy release of 6.00 MeV cannot be used to generate a 32Ge72 alpha fusion reaction. Po210 with the energy release of 5.30 MeV cannot be used to generate a 32Ge72 alpha fusion reaction. These two later radioisotopes cannot be used to generate a 32Ge72 alpha fusion reaction because their energy levels are below the threshold of 6.06 MeV that is required to initiate the reaction. Rn220 with the energy release of 6.29 MeV of energy and can also be used to generate a 32Ge72 alpha fusion reaction. It is a good choice because it is the daughter product of Th228, which is abundant on the earth. It is a daughter product of Th232, which is said to be more abundant than lead. The sited equations are a few theoretical examples from whence the present invention obtains its energy. Numerous reactions are possible. Other radioisotopes, than what is sited herein, might also be used and this will not depart from the spirit of the present invention. A number of electron emitting and electron collecting materials can be used and this will not depart from the spirit of the invention. Other cathode and anode geometries may also be used and this will not depart from the spirit of the invention. However, the target material or cathode must be a delta-ray emitter. In the scope of the present invention, “a delta ray is characterized by very fast electrons produced in quantity by alpha particles. Collectively, these electrons are defined as delta radiation when they have sufficient energy to ionize further atoms through subsequent interactions on their own.” In the present invention, a new and novel improvement in the art of the direct conversion of nuclear energy is made apparent. The present invention generates electrons that are the result of atomic reactions that are efficiently converted to electrical current, which is novel in the field. Converted atomic energy within the scope of the present invention is directly available for driving motors, lighting, production of heat, and can be used in electrochemistry, etc. . . . It is a further object of this invention to provide a device for generating electrical current that results from a self-generating electron source that is simple in construction and compact. Thus, in accordance with the present invention there is provided a method of generating delta rays, or secondary electrons through the prescribed fusion reaction. The present invention provides a method and device that gives improved performance over prior art that utilizes the direct conversion of atomic reactions to obtain electrical power. The method to generate electrical energy includes a cathode which reacts with alpha particles generating electrically charged particles. The device that will be described includes an electron generating cathode and alpha source that allows for a practical and compact power supply. Atomic reactions are converted to electrical energy with extreme efficiency within the scope of the present invention. Furthermore, it will be understood that the generated electrical current can be directly converted into a useful voltage and amperage. The conversion of the electrons that are emitted from said cathode generates useful electrical current that will be made apparent and that the alpha fusion valve is unique in generating electrical power. It will be made apparent in the following descriptions; Referring now to FIG. 1 of the drawings, the said invention consists of a vessel 1 that is made out of an electrically insulating airtight material, such as glass, ceramic, plastic or the like. It is preferred that a natural alpha source be used but an artificial alpha source might also be used and this will not depart from the spirit of the present invention. Vessel 1 includes a corona wire 2, made out of a delta-ray emissive element, compound, or alloy, such as germanium, silicon, or lead-sulfide, etc. . . . delta-ray emissive substances emit delta-ray electrons when bombarded with alpha particles. The vessel 1 contains a high work function electron-collecting cylinder 3, preferably made out of palladium because this metal can absorb a large volume of gas. After a period of time, the alpha particles lose their charge, become helium gas, build up, and the present invention eventually becomes electrically blocked. This is because helium gas is electrically non-conductive. A high work function material that has the ability to absorb gas will delay this process. Other alternative electrical collector materials, such as activated carbon, which has the ability to absorb large volumes of gas, may be used and this will not depart from the spirit of the invention. Radon gas emissive radioactive material 4 is placed at the base inside vessel 1. The radioactive material 4 can be placed in a number of locations within vessel 1 and still not depart from the spirit of the invention. The electron emitter 2 can take the form of a wire, rod, cylinder, disc, plate, etc. . . . The electron collector 3 can also take the form of a wire, rod, cylinder, disc, plate, etc. . . . I do not stake my claim on the form or geometry of the electron emitter or electron collector. I stake my claim on the method used to generate electrical power using an alpha fusion reaction. In the instant invention a negative charge of one-thousand volts or higher is applied to pin 5, which is electrically connected to corona wire 2. Respectively, a positive charge is applied to pin 6 which is electrically connected to a high work function electron collection cylinder 3. This has the effect of attracting and concentrating radon gas onto the corona wire 2 which becomes an abundant supply of alpha reactive particles. A lower voltage may also be applied across pin 5 and pin 6. The applied voltage will depend on the parameters of the wattage design of the present invention, which are too numerous to mention. Electrically conductive pin 5 and pin 6 exit through an airtight seal at the bottom of vessel 1, not shown. There are a number of sealants that are available in the field. The inner cavity of vessel 1 is evacuated of air at a low pressure of about 1/10th of an atmosphere. The amount of air that is evacuated is not critical but care must be taken not to obtain too low of a vacuum because this can result in the generation of undesirable x-ray emission. There are a number of high voltage sources that can be used to apply the required activating potential through pin 5 and pin 6 and this will not depart from the spirit of the present invention. I stake my claim to my new and novel method that directly generates electrical power which results from the alpha fusion process and I do not stake my claim to the activating external voltage source thereof. The speed in which the present invention will build up power depends on the potential difference that is applied to it and type of radon gas that it contains. The quantity of the alpha particle source determines the amount of amperage that is generated. The target material 2 is also a determining factor of how much current will be generated. When the target material 2 temperature rises, a greater number of electrons are emitted from its surface. The heated cathode 2 increases the odds of alpha particles hitting head on with its atoms, thus, producing a greater number of alpha fusion reactions, which further increases the surface heat boiling off additional thermally generated electrons. The surface area of the cathode 2 and anode 3 is also a determining factor of how much electrical current will be obtained. The present invention generates a high voltage direct current. The present invention also generates a greater amperage per given density from what has been obtained from any previously known method or device in the prior art. The instant invention described can be slightly modified to convert high voltage, high frequency, and radio frequency currents into a direct current. This feature is accomplished by adding an electrically conductive substance such as mercury, not shown, into the electrically non-conducting vessel 1. Any number of electrically conductive substances that will form a vapor or gas when heated can be used and this will not depart from the spirit of the invention. Said modification can also be utilized without the use of the radioactive substance 4, if the input source has enough energy to excite the vapor or gas into its electrically conductive state. The present modification of the primary invention is more efficient than the prior art in converting alternating or oscillating currents because there is less electrical resistance in the conversion process. Therefore, energy can be more efficiently received and converted into a direct current. Referring now to FIG. 2 of the drawings; The present invention is named alpha fusion valve 8 in the block diagram that follows: The block diagram shown illustrates an example of how an alpha fusion valve 8 can be utilized in a practical application. Many differing types of systems are made possible using the present invention and will not depart from the spirit of the invention. The alpha fusion valve 8 must be energized by an external potential difference to function if it is initially inactive or is allowed to become inactive after it has been producing power, not shown. This can be accomplished by applying a high voltage charge obtained from an electronic power supply 7. The reactions will build up within the alpha fusion valve 8 to the point where the surface of its internal electron emitter is totally bathed with radon gas. The alpha fusion valve 8 has to be primed with a potential difference to begin generating electrical power. The alpha fusion valve 8 produces a high voltage direct current. The output of the alpha fusion valve 8 can be used to charge a high voltage capacitance 9. The high voltage is then lowered to twelve volts through a step-down converter 10. The twelve volts then charges a low voltage capacitance 11 which can be a set of parallel-connected twelve-volt storage batteries. A set of parallel-connected high farad capacitors could also be used. The stored energy in capacitance 11 can be used to provide power to electrical loads that require a twelve-volt direct current or it can provide a twelve-volt power supply to an inverter 12. The output of the inverter 12 can be designed by methods known in the art to provide a voltage and frequency that is required by specific electrical loads 13. It is preferable that an electronic voltage source be used to keep the alpha fusion valve 8 in a constant energized state, which can be alternating or non-alternating. Numerous electronic circuit designs may be used to supply the potential difference required to energize the alpha fusion valve 8. Such electronic circuits are known in the field and are not what I stake my claim to. Alternatively, a strong enough source of alpha, beta, gamma radiation or a combination thereof may also be used to energize the alpha fusion valve 8. A simple earth ground and antenna raised to a suitable height can be used to take advantage of the potential difference that exists between the planet and its atmosphere, although this is not always practical. Charging capacitance 9 with this method is unpredictable and slow. Any suitable circuit may be used to supply the required potential difference to energize the alpha-fusion valve 8 and this will not depart from the spirit of the invention. |
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052271307 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT In FIG. 1, 1 designates a fuel assembly for a pressurized-water reactor. The fuel assembly 1 comprises a number of fuel rods 2 as well as guide tubes (not shown). These are retained into a bundle by spacers 3. The bundle is arranged between a top nozzle 4 and a bottom nozzle 5, which are provided with openings (not shown) for the coolant flow through the fuel assembly 1. How these openings can be throttled in order to control the coolant flow through them is shown in Swedish patent specification 8801141-6. According to the invention, the fuel assembly 1 has been provided with partial fuel boxes 6 and 7, each of which, according to FIG. 1, extending in over two ordinary spacers 3. The fuel boxes 6, 7 have been provided with bevelled corners 8 to prevent the fuel boxes 6, 7 from hooking into adjacent fuel assemblies 1. The lower fuel box 7 has been provided with a number of holes 9 at its upper edge to equalize the difference in cooling water pressure inside the fuel assembly when changing from a partial box to a boxless state. FIG. 2 shows the lower part of the fuel assembly 1 which is enclosed by a partial fuel box 7. This fuel box 7 is intended to comprise two ordinary spacers 3 but the upper one of them has been replaced by two partial spacers 10, 11 of the type described in more detail in Swedish patent 8802305-6. These spacers 10, 11 are welded to the box wall and at the same time constitute an inner support for the walls of the partial fuel box 7, which walls are conceived to be made from thin plate of Zircaloy (.about.1 mm). Thus, the box walls are conceived to serve as spacer frames. Otherwise, the upper part of the box walls is provided with inwardly-bent studs 12 to facilitate withdrawal of the fuel assembly 1 from the reactor core. FIG. 3 shows the peripheral partial spacer 10 seen from above and containing fuel rods 2 and guide tubes 13. In the spacer 10 only the outer fuel rods 2 and the guide tubes 13 are fixed in a lattice-work 14 whereas the fuel rods 2 and the guide tubes 13 in the centre of the bundle freely pass through the spacer 10. The frame of the spacer 10 is formed by the wall of the fuel box 7. FIG. 4 shows the central partial spacer 11. In this, only the centrally located fuel rods 2 and guide tubes 13 of the bundle are fixed in a lattice-work 15. Certain of the crossing plates 16 forming this central lattice 15 are extended to the walls of the surrounding fuel box 7 and fixed therein as support for the walls. When need arises, additional partial spacers may be arranged between the spacers 10 and 11 and the spacers 11 and 3, respectively, as support for the walls of the box 7. It would also be possible to arrange a simpler form of support structure at these locations. FIG. 5 shows a partial fuel box 7 seen from above in a section 5--5 in FIG. 6 which shows the walls of the fuel box 7 unfolded into a plane surface. The walls in FIGS. 5 and 6 have given the corresponding designations a, b, c, d. The figures show how a fuel box 7 can be perforated with holes 17. These holes 17 are intended to facilitate ocular inspection of the bundle inside the box walls and also to reduce the pressure difference on each side of the walls of a single fuel box. FIGS. 5-7 also show how the holes 17 are to be placed to prevent them from getting just opposite to each other when several partial fuel boxes 7 are placed adjacent to each other, as shown in FIG. 7. If the coolant flow in the fuel assembly 1 is throttled, the partial fuel boxes 6, 7 now prevent the coolant flow from adjacent unthrottled fuel assemblies from flowing over to the fuel assembly 1 already in the vicinity of the bottom nozzle 5. With both the partial fuel boxes 6, 7 in position, passage of coolant flow between adjacent fuel assemblies can only take place in the open area between the inlet box 7 and the outlet box 6. The effect of the inlet throttling is thus moved upwards and the effect of the outlet throttling is moved downwards in the bundle such that the cooling of the unthrottled highly loaded assemblies is considerably improved. |
claims | 1. A core of a light water reactor, comprising:a plurality of first fuel assemblies including a plurality of isotopes of transuranic nuclides, said plurality of first fuel assemblies being all fuel assemblies disposed in said core,wherein said transuranic nuclides included in said plurality of first fuel assemblies are different in number of a recycle frequency,wherein a plurality of second fuel assemblies being a part of said plurality of first fuel assemblies and including said transuranic nuclides having a smallest number of recycle frequency are disposed at a central part of said core,wherein a plurality of third fuel assemblies being said plurality of first fuel assemblies except said plurality of second fuel assemblies and including said transuranic nuclides having larger number of recycle frequency than said number of recycle frequency of said transuranic nuclides included in said plurality of second fuel assemblies, are disposed between said central part and an outermost layer zone of said core,wherein between said central part and said outermost layer zone, said plurality of third fuel assemblies including said transuranic nuclides having larger number of recycle frequency among said plurality of third fuel assemblies are disposed on a side of said outermost layer zone, andwherein said plurality of second fuel assemblies disposed at said central part include a nuclear fuel material including the highest rate of Pu-239 in said transuranic nuclides, and between said central part and said outermost layer zone, the third fuel assemblies having a nuclear fuel material including a lower rate of Pu-239 in said transuranic nuclides than said rate of Pu-239 of said plurality of second fuel assembly. 2. The core of a light water reactor according to claim 1,as said number of recycle frequency of said transuranic nuclides included in said plurality of third fuel assemblies become large, said plurality of third fuel assemblies including this transuranic nuclides are disposed on said side of said outermost layer zone. 3. A core of a light water reactor, comprising:a plurality of first fuel assemblies having nuclear fuel material including a plurality of isotopes of transuranic nuclides, said plurality of first fuel assemblies being all fuel assemblies disposed in said core,wherein said transuranic nuclides included in said plurality of first fuel assemblies are different in number of a recycle frequency,wherein a plurality of second fuel assemblies being a part of said plurality of first fuel assemblies and including said transuranic nuclides having a smallest number of recycle frequency are disposed at a central part of said core,wherein a plurality of third fuel assemblies being said plurality of first fuel assemblies except said plurality of second fuel assemblies and including said transuranic nuclides having larger number of recycle frequency than said number of recycle frequency of said transuranic nuclides included in said plurality of second fuel assemblies, are disposed between said central part and an outermost layer zone of said core, andwherein said plurality of second fuel assemblies disposed at said central part having include a nuclear fuel material including the highest rate of Pu-239 in said transuranic nuclides, and said plurality of third fuel assemblies disposed between said central part and said outermost layer zone include a nuclear fuel material including a lower rate of Pu-239 in said transuranic nuclides than said rate of Pu-239 of said plurality of second fuel assemblies. 4. The core of a light water reactor according to claim 3,as said number of recycle frequency of said transuranic nuclides included in said third fuel assemblies become large, said third fuel assemblies including this transuranic nuclides are disposed on said side of said outermost layer zone. |
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050948024 | claims | 1. In a nuclear fuel assembly having a plurality of fuel rods held in a spaced array by grid assemblies, guide tubes extending through the grid assemblies and attached at their upper and lower ends to an upper end fitting and a lower end fitting, the end fittings having openings therethrough for coolant flow, and a debris filter, the debris filter comprising: a. a plate attached to the top periphery of and spanning the lower end fitting; and b. said plate having a plurality of substantially pie-shaped flow holes therethrough that each measure approximately 0.181 inch in diameter with the majority of said pie-shaped flow holes arranged in groups of four to define circular clusters that each measure approximately 0.405 inch in diameter whereby the portions of said plate between said flow holes in each cluster are diagonally oriented relative to the sides of the plate. |
claims | 1. A method of manufacturing X-ray lenses comprising the steps of: a) providing a layer of liquid on a flat surface of a first substrate, b) arranging a plurality of pipe-shaped lens components in a row following an axis which extends parallel to said flat surface in said layer of liquid; and c) holding said pipe-shaped lens components between a flat surface of a second substrate and the flat surface of said first substrate, such that said liquid fills in spaces formed by an exterior surface of said pipe-shaped lens components and the flat surface of said first substrate or the flat surface of said second substrate. 2. The method of claim 1 , in which the pipe-shaped lens components are carbon nanotubes. claim 1 3. The method of claim 1 , in which the liquid is silicon grease having a viscosity reduced by addition of a solvent. claim 1 4. An X-ray lens comprising: a) a plurality of pipe-shaped lens components which transmit X-rays, b) a liquid which holds said pipe-shaped lens components and c) a pair of substrates for holding said liquid and said plurality of pipe-shaped lens components, each having a flat surface facing said pipe-shaped lens components, and said pipe-shaped lens components are lined up following a same direction. 5. The X-ray lens of claim 4 , wherein said pipe-shaped lens components are carbon nanotubes. claim 4 6. The X-ray lens of claim 4 , in which the liquid is silicon grease having a viscosity reduced by addition of a solvent. claim 4 |
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052232093 | abstract | A method for pressure relief of a containment of a nuclear power plant includes heating a washing fluid in a filter disposed inside a containment at a rated heating power through a thermal bridge, with a gas-steam mixture filling the containment, prior to initial operation of the filter. The thermal bridge is rendered substantially ineffective in an operating state of the filter, leaving the washing fluid with a continuous rated heating power being negligible in terms of filtration. In a nuclear power plant having a containment, a system for pressure relief of the containment includes a filter being disposed inside the containment and having a container. At least part of the container has two walls defining a chamber between the walls. A heat-conducting fluid at least partly fills the chamber during a heating period and is at least half evaporated after attainment of an operating temperature. |
049869604 | abstract | An end fitting for a fuel assembly. A lower plate having a pin extending perpendicular thereto at each corner slidably receives a main body portion on the pins. Stop washers or nuts on the pins prevent the main body portion from sliding off the pins. A hairpin shaped spring along each side of the main body portion is positioned between the plate and main body portion to resiliently bias them apart. The end of the spring against the lower plate is offset from the end of the plate to preload the spring tension. |
062427474 | description | DETAILED DESCRIPTION OF A PREFERRED EMBODIMENT Referring now to the drawings, FIG. 1 shows a block diagram representative of an ion implantation apparatus 1 utilizing an RF linac 23. A user input device 10 sends signals to a control calculation device 11 that receives (and stores) calculation codes from a storage device 18. The control calculation device sends signals to an operator display device 17. In addition, the control calculation device 11 sends signals to an amplitude control device 12, a phase control device 13, a frequency control device 14. Still further, the control calculation device sends signals to a convergence/divergence lens power supply 16 that powers a convergence/divergence lens 28. The amplitude control device 12 and the phase control device 13 send signals to the RF power supply 15 that powers the RF linac 23. The frequency control device 14 sends signals to the RF power supply 15 that powers the RF linac 23, and sends signals to the RF resonator portion 23-1 of the linac 23. In FIG. 2, a plan view of the ion implantation apparatus 1 of FIG. 1 is shown. An ion beam, represented by line 29, is extracted out of an ion source 21 and then passes through a mass analysis electromagnet 22 and is directed to an RF linac 23, which applies RF acceleration only on desired ions that pass through the mass analysis electromagnet 22. RF linac 23 can accelerate or decelerate an ion beam using the effect of RF fields, in a known manner. The accelerated or decelerated ion beam is deflected by an energy analysis electromagnet 24 and then undergoes energy analysis using a separation slit 25. Ions that pass through separation slit 25 are implanted into a wafer 27 in an implantation process chamber 26. A number of convergence/divergence lenses 28 for efficiently transporting the ion beam are placed in, in front of, or behind RF linac 23. Referring back to FIG. 1, the control system of RF linac 23 and convergence/divergence lens 28 is explained. Constituting the elements necessary for controlling RF linac 23 and lens (or lenses) 28 are: an input device 10 used for entering necessary conditions by an operator, a control calculation device 11 used for calculating values of various parameters from the entered conditions and for further controlling each constituting element, an amplitude control device 12 used for adjusting the RF amplitude, a phase control device 13 used for adjusting the RF phase, a frequency control device 14 used for adjusting the RF frequency, an RF power supply 15, a convergence/divergence lens power supply 16 used for convergence/divergence lens 28, a display device 17 used for displaying operation parameters, and a storage device 18 used for storing determined parameters. Moreover, numeric value calculation codes (programs) for calculating values of various parameters are stored in storage device 18 in advance. As previously discussed, RF linac 23 includes one or more RF resonators 23-1. Next, the operation of the ion implantation apparatus 1 is explained. An operator or a higher level computer enters into input device 10 the desired type of ions, the ionic valence value of ions, the extraction voltage of ion source 21, and the ion or ion beam energy value which is needed at the process chamber end of the machine. Using the internally stored numeric value calculation codes in parameter storage device 18, logic in the control calculation device 11 simulates the ion beam acceleration or deceleration, and the diversion/dispersion of the ion beam and calculates the RF linac operational parameters (amplitude, frequency and phase) for obtaining an optimum transport efficiency. At the same time, the control calculation device 11 calculates operational parameters (at least the electrical current or electrical voltage) of convergence/divergence lenses 28 for efficiently transporting an ion beam. The calculated various parameters are displayed on display device 17. As for the acceleration or deceleration conditions which are beyond the capability of RF linac 23, a message indicating that there are no solutions is displayed on display device 17. Among the parameters, the parameter related to the amplitude is sent from control calculation device 11 to amplitude control device 12, which adjusts the amplitude of the output of RF power supply 15. The parameter related to the phase is sent to phase control device 13, which adjusts the phase of the output of RF power supply 15. The parameter related to the frequency is sent to frequency control device 14. Frequency control device 14 controls the output frequency of RF power supply 15 while it also controls the resonance frequency of RF resonator 23-1 of RF linac 23. Control calculation device 11 also controls convergence/divergence lens power supply 16 using the calculated parameters for the convergence/divergence lenses 28. Ions which enter RF linac 23 and convergence/divergence lenses 28, whose operations are controlled as described above, are accelerated or decelerated to the desired energy and deflected by energy analysis electromagnet 24. Then, the ions undergo energy analysis using separation slit 25. The ions that pass through separation slit 25 are implanted into wafer 27 in implantation process chamber 26. The various parameters that are calculated using the numeric value calculation codes are stored in parameter storage device 18, after the calculation or the actual operation to obtain a beam. The control calculation device 11 simulates the acceleration or deceleration of an ion beam based on the numeric value calculation codes which are stored in advance, and automatically calculates at least one of the RF parameters of amplitude, frequency and phase. The control calculation device 11 can then operate the ion implantation apparatus by reading the stored parameters. Thus, thereafter, the desired ion beam can be obtained merely by reference to the stored parameters and without numeric calculations. Specific conditions (such as the geometrical dimensions, number of acceleration stages, a utilized frequency band, the maximum value of the amplitude, the number of convergence/divergence lenses, the maximum values thereof and so forth) of the RF linac and the convergence/divergence lens system of the ion implantation apparatus, can be incorporated into the numeric value calculation codes which are stored by control calculation device 11 in storage device 18. In this manner, a set of the codes can be switched for various types of RF linac systems and convergence/divergence lens systems. Next, with reference to FIG. 3, the calculation procedure based on the numeric value calculation codes is explained. Here, the explanation is performed for a case in which RF resonators 23-1 consist of the first through fourth RF resonators. The process includes nine steps, referenced herein as S1-S9. In Step S1, an operator or a higher level computer enters the calculation conditions into input device 10. Here, an ion source extraction voltage, an ion mass, and an ionic valence value of ions are entered as incoming beam conditions, and the final energy value EF of the ions or ion beam is entered as an outgoing beam condition. In Step S2, the initialization calculation is performed. In other words, a plurality of outgoing beam energy values (E1 through E8) are calculated using the predetermined eight combinations of phase and voltage for the given incoming beam conditions. Here, E1 is the theoretically the lowest energy and E8 the largest energy. The combinations of phase and voltage are determined so that the outgoing energy levels E1 through E8 are separated by approximately the same energy incremental values. In Step S3, the final energy value EF and each of the calculated outgoing beam energy values (E1 through E8) are compared. In Step S4, conversion calculation is performed. In the conversion calculation, if for example, E4<EF<E5, then the value of voltage or phase is altered between the conditions of E4 and E5 until an outgoing beam energy becomes equal to the desired final energy value EF. In Step S5, temporary operational parameters for the RF linac are obtained as a result of repeated calculations of Step S4. In Step S6, the optimization of the bunching phase (first resonator) of the linac is performed. In other words, using the temporary parameters as the initial set, the phases of the resonance frequencies of the second through fourth RF resonators are varied until a phase combination which maximizes the transport efficiency of RF linac 23 is found. In Step S7, RF linac operational parameters are obtained as the result of Step S6. In Step S8, optimization for convergence/divergence lenses 28 is performed. In other words, simulation for the ion beam is performed by varying the parameters of convergence/divergence lenses 28 against the RF parameters of RF linac 23 which are obtained in the above step. The simulation includes the lateral spread of the ion beam. Thus, the strength of convergence/divergence lenses 28 for the maximum transport efficiency is obtained. In the final step S9, the final parameters are obtained. This is done by combining the RF parameters with the parameters for the convergence divergence lenses 28. As previously discussed, in the prior art, parameters are determined within an ion implantation apparatus and the determination provides analytical solutions (in other words, the solution of equations). Conversely, the most prominent feature of the present invention lies in the improvement by which numeric value calculation codes have been developed so that they can be applied to an RF acceleration system or convergence lens system for which analytical solutions cannot be obtained. Simulation utilizing numerical calculation is performed within an ion implantation apparatus and thereby parameters can be automatically determined. Acceleration parameters of an RF system or parameters of convergence/divergence lenses for totally new acceleration conditions were conventionally obtained by expending a very large amount of effort and time through a procedure such as the one illustrated in FIG. 6. The present inventions allow such parameters to be automatically determined by merely an operator or a higher level computer entering acceleration conditions (e.g., ionic valence value of ions, a desired energy value, etc.). FIG. 4 briefly illustrates the entire procedure. An operator enters the acceleration conditions and a final energy value into the input device 10, and the control calculation device 11 determines the optimum solution (by numeric simulation), and determines the linac and convergence/divergence lens operational parameters. In other words, the operation can now be performed with the same ease as the operation performed for a prior art ion implantation apparatus that accelerates ions utilizing an electrostatic field. Thus, regarding the process to determine a new set of linac operational parameters, there are advantages. The time required to determine the parameters is drastically reduced (approximately one minute according to the experiment results.) Effort by an operator to determine parameters is almost eliminated. Optimum parameters can be determined without iteration by trial-and-error. The quality of determined parameters does not depend on the skill of an operator and hence, is reproducible. Even when a higher level computer enters acceleration conditions or a final energy value, the apparatus of the present invention can automatically determine the parameters. Hence, it is possible to achieve completely automatic operation of the apparatus. As explained hereinabove, according to the present invention, operating conditions of an ion implantation apparatus that utilizes an RF acceleration method can be determined with ease in a short period of time. Moreover, an ion beam having any energy value can be obtained in a short period of time. Accordingly, a preferred embodiment has been described for a method and system for optimizing linac operational parameters in an ion implantation apparatus. With the foregoing description in mind, however, it is understood that this description is made only by way of example, that the invention is not limited to the particular embodiments described herein, and that various rearrangements, modifications, and substitutions may be implemented with respect to the foregoing description without departing from the scope of the invention as defined by the following claims and their equivalents. |
052664945 | summary | FIELD OF THE INVENTION This invention relates to a method and apparatus for cleaning particulate materials such as soils which are contaminated with a variety of contaminants such as heavy metals, radioactive compounds and organics, often in combination, through a combination of leaching, washing, attrition scrubbing, countercurrent flow size separation and density separation. This invention further relates to the recovery of such contaminates following removal from the soils, for additional processing, recycling or disposal. Most particularly this invention relates to a bench scale method of evaluating soil for contaminants and other conditions in order to determine the best approach for washing the soil. BACKGROUND INFORMATION Contaminated soil is becoming a more serious environmental problem every day. The contaminants can include heavy metals, such as for instance, copper, lead and mercury; radioactive species such as for example, radium, uranium and thorium; and organics such as for example, oils, polychlorinated biphenyls, (PCB's) flue soot and others. Various techniques have been developed to remove specific contaminants from soil. For instance, heavy metals are known to be found predominantly in the silt, humic or clay fraction of soil. Hence, they can be removed by size separation such as tiltable tables, concurrent flow in a mineral jig and chemical techniques, such as the use of leachates. The radioactive compounds when originating as a spill can be removed to a large extent by leaching. Since these compounds are often also present in the finer particles, the most severely contaminated fraction can also be removed by countercurrent flow size separation. Organics can be removed by washing with surfactants, thermal treatment or biological processes. Special problems develop when the different types of contaminants are present in the same soil. Generally, biological or thermal processes are more effective for removing organics than washing, in the case of finer grain soils and clays. However, toxic inorganics such as lead or chromium (+6), if present, tend to deactivate biological systems due to their toxicity and aggravate air pollution problems endemic to thermal destruction process. In addition, thermal processes may mobilize contaminants that were otherwise fixed in the treated soil. Radioactive contamination (e.g., uranium, thorium radium, etc.) can be removed by soil washing. Soil washing provides a means to process soils having multiple contaminants. The washed soil is compatible with subsequent biological or thermal treatment. Inorganic and radioactive compounds may be separated from organics for separate sale or disposal. Many soil washing processes are presently available. Most use mine equipment to provide intimate soil/extractant contact. U.S. Pat. No. 4,783,253 discloses a process for separating radioactive contaminants from soil using a concurrent flow of water to float away lighter uncontaminated particles from heavy contaminated particles. The slurry of lighter particles is dewatered using a spiral classifier, centrifuge, filter or the like. U.S. Pat. No. 4,783,263 is directed to a process for removing toxic or hazardous substances, in particular organics, from soils and the like by converting the material to a slurry, adding surfactants and/or alkaline agents, and concentrating the toxic substance in the liquid phase preferably with a modifier in a froth flotation cell. Some of the limitations of the currently used processes are that they are optimized for removing only one type of contaminant or for cleaning only one type of soil, they are geared to cleaning the larger particles while concentrating the fines in a fraction for later disposal, and they often use filtration for water removal which is a capital intensive operation with high operating costs. Once the contaminants have been removed from the soil or other particulate material they must in turn be recovered for further processing, such as mining and/or smelting in the case of heavy metals, or disposal, for example, through mixing with a fixative material such as concrete. The ability to recover contaminants from the cleaning system is to a large extent dependent upon the method by which the contaminants were removed from the soil in the first instance. Mineral extraction in general and soil washing in particular often require the oxidation of the metals and sometimes the organic fraction of the soil for the removal of the metals. Radioactive metals are also included with heavy metals requiring oxidation, since most radioactive materials are also heavy metals, such as uranium, thorium or radium. Some typical oxidants for heavy metal removal include nitric acid, sodium hypochlorite and calcium hypochlorite. However, the use of nitric acid is generally not practical due to the fact that nitric acid is nonselective in its action, dissolving the rock matrix as well as oxidizing and dissolving the metal of interest, it is expensive, and results in nitrate-laden waste liquors which can present environmental hazards unless treated. Sodium hypochlorite is expensive to use because commercial solutions are supplied as a 15% liquid which increases the freight costs. Calcium hypochlorite introduces large amounts of calcium ion into the leachate solution when used in quantities sufficient to oxidize the metal, and the calcium ions can then precipitate if carbonate bleach liquors are used or if the leachate solution is left standing in contact with air. This calcium carbonate precipitate is difficult to handle and can clog processing equipment. In addition, if common soaps are used to remove organics, the high calcium ion content tends to precipitate some of the soap which requires use of additional soap. In addition to the above problems, frequently, it is not immediately clear which of the several possible soil washing techniques, or combinations thereof, should be used for a soil of particular interest. Even if the precise nature of contamination is known (and it may not be) the most efficient method of removing that contaminant or contaminants may depend on a host of variables and trial-and-error solutions. Attempting to make a determination of the economics of particular processing methods and parameters in the field is impractical, and would require shifting and replacing relatively large pieces of equipment, refitting pumps and piping, etc. until the best soil washing approach is determined for each particular site. Furthermore, field equipment often requires large batches of soil to operate effectively, further increasing the inefficiency of determining an optimal washing method for the particular soil. There is a need therefore for an improved process and apparatus for treating particulate materials, such as soil and the like, contaminated with a mixture of wastes such as radioactive materials, organics and heavy metals. There is a further need for such a process and apparatus which separates organic and inorganic contaminants thereby allowing for optimum disposal routes or post treatment strategies to be used on the concentrated contaminated fractions. There is also a need for such a process and apparatus which produces a high solids content fines stream. There is yet another need for such a process and apparatus which is not capital intensive, is economical to operate and can be made portable for on-site treatment. There is a further need for a system that can effectively recover the contaminants once they have been removed from the soil, requiring a minimal amount of equipment, chemicals, and being portable to the job site, which further allows for the processing of recovered contaminants, such as metals, through mining and/or smelting operations, and allows for effective leach-resistant fixation of contaminants which are to be disposed. There is also a need for a scaled-down soil washing evaluation system which may be run on relatively small batches of contaminated soil and which quickly and accurately provides optimal soil washing parameters for the particular contaminated particulate matter being evaluated, be it soil, sludge or other solids. OBJECTS OF THE INVENTION It is an object of the invention to provide a bench or small scale process for treating contaminated soils which contain organic and/or inorganic contaminants and is able to remove such contaminants from a few pounds or less of soil sample. It is a further object to provide a process which evaluates the distribution of contaminants in the soil of interest. It is yet another object of the invention to provide a process which evaluates the extraction efficiency of a variety of extractants while modeling the soil washing process of interest. It is still a further object of the invention to provide a process which evaluates the effectiveness of the various leachate treatment steps. It is another object of the invention to provide a process which determines the volumes and probable product recoveries and contaminant concentrations in those product streams. It is a further object of the invention to determine the amount of extractants, acids, flocculent, and other chemicals needed for operation of any soil washing system. These and other objects and advantages of the present invention will become more readily apparent as the following detailed description of the preferred embodiments proceeds. SUMMARY OF THE INVENTION According to the present invention, a method of characterizing contaminated soil in order to determine an effective treatment approach for removing contaminants from the soil is disclosed. The method comprises the steps of obtaining a representative contaminated soil sample from a site containing the contaminated soil; identifying particle size ranges for the contaminated soil by passing at least a portion of the representative contaminated soil sample through a series of particle size classifiers; identifying contaminants to be removed from the contaminated soil; identifying an effective soil washing extractant by passing at least a portion of the representative contaminated soil sample through a bench scale soil washing process adapted to substantially correspond to a full-scale soil washing process for the contaminated soil, the bench scale soil washing process being adapted for removing contaminants from contaminated soils having particle size ranges corresponding to the ranges identified in the previous step, the bench scale soil washing process including washing a known quantity of the representative soil sample with a first extractant, then repeating the bench scale soil washing process using at least one additional extractant (which may be a different concentration of the first extractant) and recovering and identifying at least one of the contaminants from the representative contaminated soil sample, in order to determine those soil washing conditions favoring extraction of the contaminant from the contaminated soil; and identifying a suitable leachate treatment process for treating at least one contaminant removed in the previous step from the representative contaminated soil sample. In one preferred embodiment of the invention the particle size classifiers comprise a set of screens including a first screen and at least one additional screen arranged in order of decreasing screen size relative to the direction of passing the representative soil sample through the screens. |
claims | 1. A protective shroud covering a radiation shield of a nuclear pharmacy generator, the radiation shield having seams between components of the radiation shield, the shroud comprising;a first shroud end comprising a first opening that receives the radiation shield;a second shroud end that is disposed oppositely of and spaced from the first shroud end, wherein the second shroud end comprises a second opening that provides access to a portion of the radiation shield when the shroud is installed on the radiation shield, wherein the first and second openings face in opposite directions, wherein the second shroud end further comprises at least one flange that extends inward from the tubular body and that defines the second opening for the second shroud end;a tubular body that extends between the first and second shroud ends and that is positioned about the radiation shield;a raised boss disposed on the flange, wherein the raised boss is configured to store at least one radiopharmacy tool;at least one hoop, wherein a first part of the hoop is disposed on the flange, and wherein a second part of the hoop extends outwardly from the tubular body and accommodates storing at least one radiopharmacy tool. 2. The protective shroud according to claim 1, wherein a diameter of the second opening is less than a diameter of the first opening and is less than a diameter of the radiation shield. 3. The protective shroud according to claim 1, in combination with a nuclear pharmacy generator. 4. The protective shroud according to claim 1, wherein the shroud covers substantially all of the seams of the radiation shield. 5. The protective shroud according to claim 1, wherein the tubular body has a continuously smooth exterior surface. 6. The protective shroud according to claim 1, wherein the shroud includes a material that prevents damage to the radiation shield. 7. The protective shroud according to claim 6, wherein the material includes one or more of vulcanized rubber, neoprene, polyurethane, plastics and silicone. 8. A nuclear pharmacy generator assembly comprising:a radiation shielding body comprising a first end, an oppositely disposed second end, and a sidewall extending between the first and second ends, wherein the sidewall of the radiation shielding body comprises a plurality of seams;a nuclear pharmacy generator disposed within said radiation shielding body;an elution tool that houses an elution container that is fluidly connected with the nuclear pharmacy generator; anda shroud installed on the radiation shielding body and comprising:a first shroud end comprising a first opening, wherein the radiation shielding body is received within the first opening of the shroud;a second shroud end that is disposed oppositely of the first shroud end and that comprises a second opening, wherein an entirety of the shroud is located between the first opening on the first shroud end and the second opening on the second shroud end, wherein the first and second openings of the shroud face away from each other in opposite directions, wherein the second opening of the shroud is disposed beyond the second end of the radiation shielding body with the second end of the radiation shielding body being located between the first end of the radiation shielding body and the second shroud end, and wherein the elution tool is accessible through the second opening on the second shroud end from an exterior of the nuclear pharmacy generator assembly and through the second end of the radiation shielding body, all when the shroud is installed on the radiation shielding body; anda tubular body that extends between the first and second shroud ends and that is positioned about the sidewall of the radiation shielding body. 9. The nuclear pharmacy generator assembly of claim 8, wherein the shroud partially enshrouds the radiation shielding body. 10. The nuclear pharmacy generator assembly of claim 8, wherein the radiation shielding body comprises a plurality of rings which in turn comprises a base ring at the first end of the radiation shielding body, wherein a seam of the plurality of said seams is located between each adjacent pair of rings of the plurality of rings, and wherein substantially all of the base ring is covered by the shroud. 11. The nuclear pharmacy generator assembly of claim 8, wherein the shroud substantially covers each seam of the plurality of seams. 12. The nuclear pharmacy generator assembly of claim 8, wherein the second end of the shroud further comprises a flange that extends inwardly from the tubular body and forms the second opening of the shroud. 13. The nuclear pharmacy generator assembly of claim 12, wherein the radiation shielding body does not extend through the second opening of the shroud. 14. The nuclear pharmacy generator assembly of claim 8, wherein the shroud further comprises a first hoop positioned alongside the tubular body, and wherein the nuclear pharmacy generator assembly further comprises one of a vial holder and a recovering tool positioned within the first hoop. 15. The nuclear pharmacy generator assembly of claim 8, wherein the shroud further comprises first and second hoops that are each positioned alongside the tubular body, wherein the nuclear pharmacy generator assembly further comprises a vial holder positioned within the first hoop and a recovering tool positioned within the second hoop. 16. The nuclear pharmacy generator assembly of claim 8, further comprising:an eluant container;an input connector to the nuclear pharmacy generator, wherein the eluant container is fluidly connected with the input connector; andan output connector from the nuclear pharmacy generator, wherein the elution container is fluidly connected with the output connector. 17. The nuclear pharmacy generator assembly of claim 16, wherein the input connector comprises an input needle, and wherein the output connector comprises an output needle. 18. The nuclear pharmacy generator assembly of claim 17, wherein the eluant container stores an eluant fluid, wherein the eluant fluid exits the eluant container and enters the nuclear pharmacy generator through the input needle, and wherein a radioisotope-containing fluid exists the nuclear pharmacy generator through the output needle and enters the elution container. 19. The nuclear pharmacy generator assembly of claim 16, wherein the eluant container stores an eluant fluid, wherein the eluant fluid exits the eluant container and enters the nuclear pharmacy generator through the input connector, and wherein a radioisotope-containing fluid exists the nuclear pharmacy generator through the output connector and enters the elution container. |
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description | This application is a continuation in part (CIP) of co-pending U.S. application Ser. No. 15/488,983, filed Apr. 17, 2017, which claimed priority to U.S. application Ser. No. 14/190,389, filed Feb. 26, 2014, which has issued as U.S. Pat. No. 9,636,524 on May 2, 2017, which claimed priority to U.S. application Ser. No. 13/532,447, filed on Jun. 25, 2012, now abandoned, which claimed priority to provisional U.S. patent application 61/571,406 filed Jun. 27, 2011. The present application also claims priority to U.S. patent application 62/749,875, filed Oct. 24, 2018. This invention is in the technical area of apparatus and methods for Boron Neutron capture therapy for cancer. Boron Neutron Capture Therapy (BNCT) is not new in the art, as thermal neutrons have been used for cancer therapy for the destruction of cancer tumors. These neutrons interact with boron-10 that has been placed at the cancer site. The neutrons interact with the boron to produce fission events whereby alpha particles and lithium nuclei are created. These massive ionized particles are then released, destroying the chemical bonds of nearby cancer tumor cells. At present the neutrons created in a reactor or accelerator pass through a moderator, which shapes the neutron energy spectrum suitable for BNCT treatment. While passing through the moderator and then the tissue of the patient, the neutrons are slowed by collisions and become low energy thermal neutrons. The thermal neutrons undergo reactions with the boron-10 nuclei at a cancer site, forming compound nuclei (excited boron-11), which then promptly disintegrate to lithium-7 and an alpha particle. Both the alpha particle and the lithium ion produce closely spaced ionizations in the immediate vicinity of the reaction, with a range of approximately 5-9 micrometers, or roughly the thickness of one cell diameter. The release of this energy destroys surrounding cancer cells. This technique is advantageous since the radiation damage occurs over a short range and thus normal tissues can be spared. Gadolinium can also be considered as a capture agent in neutron capture therapy (NCT) because of its very high neutron capture cross section. A number of gadolinium compounds have been used routinely as contrast agents for imaging brain tumors. The tumors have absorbed a large fraction of the gadolinium, making gadolinium an excellent capture agent for NCT. Therefore, GNTC may also be considered as a variation in embodiments of the present invention. The following definitions of neutron energy ranges, E, are used frequently by those skilled in the art of producing and using neutrons for medical, commercial and scientific applications: Fast (E>1 MeV), Epithermal (0.5 eV<E<1 Mev) and Thermal (E<0.5 eV) neutrons. BNCT has the potential to treat previously untreatable cancers such as glioblastoma multiforme (GBM). In the US brain tumors are the second most frequent cause of cancer-related deaths for males under 29 and females under 20. GBM is nearly always fatal and has, until now, no known effective treatment. There are approximately 13,000 deaths per year due to primary brain tumors. If conventional medicine is used where the glioblast is excised, new tumors almost invariably recur, frequently far from the original tumor site. Effective radiation therapy, therefore, must encompass a large volume and the radiation must be uniformly distributed. Conventional radiation treatment is usually too toxic to be of use against GBM. For distributed tumors, effective radiation therapy must encompass a larger volume and the radiation must be uniformly distributed. This is also true of liver cancers. The liver is the most common target of metastases from many primary tumors. Primary and metastatic liver cancers are usually fatal, especially after resection of multiple individual tumors. The response rate for nonresectable hepatocellular carcinoma to traditional radiation treatment or chemotherapy is also very poor. However, recent results indicate that the thermal neutron irradiation of the whole liver with a 10B compound, to be bombarded with low-energy neutrons, could be a way to destroy all the liver metastases. Recent research in BNCT has shown that neutron capture therapy can be used to treat a large number of different cancers. BNCT has been found to be effective and safe in the treatment of inoperable, locally advanced head and neck carcinomas that recur at sites that were previously irradiated with traditional gamma radiation. Thus, BNCT could be considered for a wider range of cancers. BNCT holds such promise because the dose to the cancer site can be greatly enhanced over that produced by γ-radiation sources. This is a consequence of the fact that the neutron-boron reaction produces the emission of short-range (5-9 um distance) radiation, and consequently normal tissues can be spared. In addition, boron can achieve a high tumor-to-brain concentration ratio, as much as ten or more, thereby preferentially destroying abnormal tissue. BNCT has been tested using either nuclear reactors or accelerators to produce the neutrons, which are not practical or affordable for most clinical settings. Reactors also do not produce an ideal neutron spectrum and are contaminated with γ-radiation. Fusion generators produce fast neutrons from the deuterium-deuterium (DD) or the deuterium-tritium (DT) reactions and are, in general, smaller and less expensive than accelerators and reactors. Fast neutrons thus produced must be moderated or slowed down to thermal or epithermal neutron energies using, for example, water or other hydrogen bearing materials. The fusion neutron generator has three basic components: an ion source, an electron shield and an acceleration structure with a target. The ions are accelerated from the ion source to usually a titanium target using a high voltage potential of between 40 kV to 200 kV, which can be easily delivered by a modern high voltage power supply. An electron shield is usually disposed between the ion source and the titanium target. This shield is voltage biased to repel electrons being generated when the positive D+ ions that strike the titanium target. This prevents these electrons from striking the ion source and damaging it due to electron heating. The target uses a deuterium D+ or tritium T+ absorbing material such as titanium, which readily absorbs the D+ or T+ ions, forming a titanium hydride. Succeeding D+ or T+ ions strike these embedded ions and fuse, resulting in DD, DT or TT reactions and releasing fast neutrons. Prior attempts at proposing fusion generators required the use of the DT reaction with the need for radioactive tritium and high acceleration powers. High yields of fast neutrons/sec were needed to achieve enough thermal neutrons for therapy in a reasonable length of time of therapy treatments. These prior schemes for achieving epithermal neutron fluxes are serial or planar in design: a single fast neutron generator is followed by a moderator, which is followed by the patient. Unfortunately, since the neutrons are entering from one side of the head, the planar neutron irradiation system leads to a high surface or skin dosage and a decreasing neutron dose deeper into the brain. The brain is not irradiated uniformly, and cancer sites have lower thermal neutron dosage the further they are from the planar port. A conventional planar neutron irradiation system 14 and its operation is shown in FIG. 1 labeled Prior Art. Conversion of fast neutrons 22 to thermal neutrons 30 takes place in a series of steps. First the fast neutrons 22 are produced by a cylindrical fast neutron generator 20 and then enter a moderating means 18 where they suffer elastic scatterings (collisions with nuclei of the moderating material's atoms). This lowers the fast neutrons to epithermal neutron 24 energies. A mixture of epithermals 24 and thermal neutrons 30 are emitted out of a planar port 16 and then enter the patient's head 26. The epithermal neutrons 24 are moderated still further in the patient's brain and moderated further to thermal neutrons, finally being captured by the boron at the tumor site. The fission reaction occurs, and alpha and Li-7 ions are released, destroying the tumor cells. The epithermal and thermal neutrons reach the patient's head through a planar port 16 formed from neutron absorbing materials that form a collimating means 28. The thermal and epithermal neutrons strike the patient's head on one side, and many neutrons escape or are not used. One escaping neutron 38 is shown as representative. This is an inefficient process requiring a large number of fast neutrons to be produced in order to produce enough thermal neutrons for reasonable therapy or treatment times (e.g. 30 min). To achieve higher yields of fast neutrons the planar neutron irradiation system 14 requires that one use either the DD fusion reaction with extremely high acceleration powers (e.g. 0.5 to 1.5 Megawatts) or the DT reaction which has an approximate 100-fold increase in neutron yield for the same acceleration power. The use of tritium has a whole host of safety and maintenance problems. Tritium gas is radioactive and extremely difficult to eliminate once it gets on to a surface. In the art of producing fast neutrons this requires that the generator be sealed and have a means for achieving a vacuum that is completely sealed. The generator head cannot be easily maintained and usually its lifetime is limited to less than 2000 hours. This reduces the possible use of this generator for clinical operation since the number of patients who could be treated would be small before the generator head would need replacement. On the other hand, the use of the DD fusion reaction allows one skilled in the art to use an actively-pumped-vacuum means with roughing and turbo pumps. The generator can then be opened for repairs and its lifetime extended. This makes the DD fusion reaction neutron generator optimum for clinical use. The downside for the DD fusion reaction is that high acceleration powers are required to achieve the desired neutron yield required by prior art methods. Improving the efficiency of producing the right thermal neutron flux at the cancer site is imperative for achieving BNCT in a clinical and hospital setting. In an embodiment of the invention a neutron generator is provided, comprising a pre-moderator block of moderating material having an upper surface, a lower surface, a first and a second end, first and second side surfaces, a first length, a first width substantially less than the first length, and a first thickness, a cylindrical acceleration chamber having a first diameter substantially the first width of the pre-moderator block, sealed at one end to the upper surface of the pre-moderator block adjacent the first end of the pre-moderator block, with a vertical axis perpendicular to the upper surface, the acceleration chamber having a height and a top cover at a second end away from the pre-moderator block, a vacuum pump engaging the acceleration chamber, evacuating the acceleration chamber to a moderately high vacuum, a plasma ion chamber opening into the acceleration chamber through an ion extraction iris through the top cover of the acceleration chamber on the vertical axis of the acceleration chamber, a gas source providing deuterium gas to the plasma ion chamber, a microwave energy source ionizing the gas in the plasma ion chamber, a cylindrical primary isolation well extending a substantial distance into the pre-moderator block from the upper surface, centered on the vertical axis of the acceleration chamber, a secondary isolation well substantially in a shape of a hollow cylinder surrounding the primary isolation well, to a depth somewhat less than the substantial distance of the primary isolation well, within the first diameter of the acceleration chamber, a water-cooled titanium target disk having a target surface orthogonal to the axis of the acceleration chamber, the target disk having a diameter substantially smaller then a diameter of the isolation well, positioned at a lower extremity of the isolation well, the target disk biased to a substantial negative DC voltage, and electrically grounded metal cladding covering all otherwise exposed surfaces of the pre-moderator block. Ions extracted through the ion extraction iris are accelerated to bombard the titanium target at the lower extremity of the primary isolation well, producing energetic neutrons that pass through and are moderated by the pre-moderator block, and wherein any path along a surface of the from the titanium target to an electrically grounded element is necessarily maximized by the primary and secondary isolation wells. In one embodiment the material of the pre-moderator block is Ultra-High-Molecular-Weight Polyethylene (UHMWPE or UHMW), or High-Density Polyethylene (HDPE), or polytetrafluoroethene (PTFE). Also, in one embodiment surfaces of the primary and secondary isolation wells are formed in continuous curves and are roughened to enhance resistance to high-voltage flashover. In one embodiment the generator further comprises supply and return water channels lengthwise through the pre-moderator block from the second end of the pre-moderator block to the titanium target, providing cooling water cooling the target. In one embodiment the neutron generator further comprises a female socket for a high-voltage male connector at the second end of the pre-moderator block, coupled to a high-voltage bus bar implemented lengthwise through the pre-moderator block, to the target, for biasing the target to a substantial negative DC voltage. Also, in one embodiment both the first and second side surfaces of the pre-moderator block are angled inward from vertical by thirty degrees for at least a portion of the height, enabling six neutron generators to be placed with the angled side surfaces fully adjacent, forming a closed ring about a center point. Also, in one embodiment both the first and second side surfaces of the pre-moderator block are angled inward from vertical by forty-five degrees for at least a portion of the height, enabling eight neutron generators to be placed with the angled side surfaces fully adjacent, forming a closed ring about a center point. In another aspect of the invention a boron neutron cancer treatment system is provided, comprising a secondary moderator having a central treatment chamber for a subject, and six substantially identical neutron generators, each comprising a pre-moderator block of moderating material having an upper surface, a lower surface, a first and a second end, opposite side surfaces angled inward by thirty degrees along at least a portion of the height, a first length, a first width substantially less than the first length, and a first thickness, a cylindrical acceleration chamber having a first diameter substantially the first width of the pre-moderator block, sealed at one end to the upper surface of the pre-moderator block adjacent the first end of the pre-moderator block, with a vertical axis perpendicular to the upper surface, the acceleration chamber having a height and a top cover at a second end away from the pre-moderator block, a vacuum pump engaging the acceleration chamber at a right angle to the vertical axis, evacuating the acceleration chamber to a moderately high vacuum, a plasma ion chamber opening into the acceleration chamber through an ion extraction iris through the top cover of the acceleration chamber on the vertical axis of the acceleration chamber, a gas source providing deuterium gas to the plasma ion chamber, a microwave energy source ionizing the gas in the plasma ion chamber, a cylindrical primary isolation well extending a substantial distance into the pre-moderator block from the upper surface, centered on the vertical axis of the acceleration chamber, a secondary isolation well substantially in a shape of a hollow cylinder surrounding the primary isolation well, to a depth somewhat less than the substantial distance of the primary isolation well, within the first diameter of the acceleration chamber, a water-cooled titanium target disk having a target surface orthogonal to the axis of the acceleration chamber, the target disk having a diameter substantially smaller then a diameter of the isolation well, positioned at a lower extremity of the isolation well, the target disk biased to a substantial negative DC voltage, and electrically grounded metal cladding covering all otherwise exposed surfaces of the pre-moderator block. The six neutron generators are positioned around the secondary moderator with the axis of each acceleration chamber passing through the center of the treatment chamber, and with the angled sides of the neutron generators fully adjacent. In one embodiment the system further comprises six substantially rectangular spacing blocks of moderator material, one spacing block placed between each adjacent neutron generator with sides of the spacing blocks fully adjacent with the angled sides of the neutron generators. Also, in one embodiment the secondary moderator is shaped to fill all volume between the neutron generators and the central treatment chamber. In one embodiment the secondary moderator is a block or blocks of solid moderator material. In one embodiment the secondary moderator is a container filled with heavy water. And in one embodiment the secondary moderator is a container filled with granulated moderator material. In yet another aspect of the invention a boron neutron cancer treatment system is provided, comprising a secondary moderator having a central treatment chamber for a subject, and eight substantially identical neutron generators, each comprising a pre-moderator block of moderating material having an upper surface, a lower surface, a first and a second end, opposite side surfaces angled inward by forty-five degrees along at least a portion of the height, a first length, a first width substantially less than the first length, and a first thickness, a cylindrical acceleration chamber having a first diameter substantially the first width of the pre-moderator block, sealed at one end to the upper surface of the pre-moderator block adjacent the first end of the pre-moderator block, with a vertical axis perpendicular to the upper surface, the acceleration chamber having a height and a top cover at a second end away from the pre-moderator block, a vacuum pump engaging the acceleration chamber at a right angle to the vertical axis, evacuating the acceleration chamber to a moderately high vacuum, a plasma ion chamber opening into the acceleration chamber through an ion extraction iris through the top cover of the acceleration chamber on the vertical axis of the acceleration chamber, a gas source providing deuterium gas to the plasma ion chamber, a microwave energy source ionizing the gas in the plasma ion chamber, a cylindrical primary isolation well extending a substantial distance into the pre-moderator block from the upper surface, centered on the vertical axis of the acceleration chamber, a secondary isolation well substantially in a shape of a hollow cylinder surrounding the primary isolation well, to a depth somewhat less than the substantial distance of the primary isolation well, within the first diameter of the acceleration chamber, a water-cooled titanium target disk having a target surface orthogonal to the axis of the acceleration chamber, the target disk having a diameter substantially smaller then a diameter of the isolation well, positioned at a lower extremity of the isolation well, the target disk biased to a substantial negative DC voltage, and electrically grounded metal cladding covering all otherwise exposed surfaces of the pre-moderator block. The eight neutron generators are positioned around the secondary moderator with the axis of each acceleration chamber passing through the center of the treatment chamber, and with the angled sides of the neutron generators fully adjacent. In one embodiment this system further comprises eight substantially rectangular spacing blocks of moderator material, one spacing block placed between each adjacent neutron generator with sides of the spacing blocks fully adjacent with the angled sides of the neutron generators. In one embodiment the secondary moderator is shaped to fill all volume between the neutron generators and the central treatment chamber. In one embodiment the secondary moderator is a block or blocks of solid moderator material. In one embodiment the secondary moderator is a container filled with heavy water. And in one embodiment the secondary moderator is a container filled with granulated moderator material. In yet another aspect of the invention a treatment system for evaluating boron sources for boron neutron cancer therapy (BNCT) is provided, comprising a substantially square secondary moderator having a central treatment chamber for a subject, and four substantially identical neutron generators, each comprising a pre-moderator block of moderating material having an upper surface, a lower surface, a first and a second end, opposite parallel side surfaces, a first length, a first width substantially less than the first length, and a first thickness, a cylindrical acceleration chamber having a first diameter substantially the first width of the pre-moderator block, sealed at one end to the upper surface of the pre-moderator block adjacent the first end of the pre-moderator block, with a vertical axis perpendicular to the upper surface, the acceleration chamber having a height and a top cover at a second end away from the pre-moderator block, a vacuum pump engaging the acceleration chamber at a right angle to the vertical axis, evacuating the acceleration chamber to a moderately high vacuum, a plasma ion chamber opening into the acceleration chamber through an ion extraction iris through the top cover of the acceleration chamber on the vertical axis of the acceleration chamber, a gas source providing deuterium gas to the plasma ion chamber, a microwave energy source ionizing the gas in the plasma ion chamber, a cylindrical primary isolation well extending a substantial distance into the pre-moderator block from the upper surface, centered on the vertical axis of the acceleration chamber, a secondary isolation well substantially in a shape of a hollow cylinder surrounding the primary isolation well, to a depth somewhat less than the substantial distance of the primary isolation well, within the first diameter of the acceleration chamber, a water-cooled titanium target disk having a target surface orthogonal to the axis of the acceleration chamber, the target disk having a diameter substantially smaller then a diameter of the isolation well, positioned at a lower extremity of the isolation well, the target disk biased to a substantial negative DC voltage, and electrically grounded metal cladding covering all otherwise exposed surfaces of the pre-moderator block. The four neutron generators are positioned around the substantially square secondary moderator with the axis of each acceleration chamber passing through the center of the treatment chamber. In one embodiment of this system the secondary moderator is a block or blocks of solid moderator material. In one embodiment the secondary moderator is a container filled with heavy water. And in one embodiment the secondary moderator is a container filled with granulated moderator material. In yet one more aspect of the invention a Boron neutron cancer treatment system is provided, comprising a moderator chamber filled with liquid or granular moderator material except for a central treatment chamber, the moderator chamber having parallel upper and lower surfaces, a plurality of neutron generators, each comprising a pre-moderator block of moderating material having an upper surface, a lower surface, a first and a second end, opposite side surfaces, a first length, a first width substantially less than the first length, and a first thickness, a cylindrical acceleration chamber having a first diameter substantially the first width of the pre-moderator block, sealed at one end to the upper surface of the pre-moderator block adjacent the first end of the pre-moderator block, with a vertical axis perpendicular to the upper surface, the acceleration chamber having a height and a top cover at a second end away from the pre-moderator block, a vacuum pump engaging the acceleration chamber, evacuating the acceleration chamber to a moderately high vacuum, a plasma ion chamber opening into the acceleration chamber through an ion extraction iris through the top cover of the acceleration chamber on the vertical axis of the acceleration chamber, a gas source providing deuterium gas to the plasma ion chamber, a microwave energy source ionizing the gas in the plasma ion chamber, a cylindrical primary isolation well extending a substantial distance into the pre-moderator block from the upper surface, centered on the vertical axis of the acceleration chamber, a secondary isolation well substantially in a shape of a hollow cylinder surrounding the primary isolation well, to a depth somewhat less than the substantial distance of the primary isolation well, within the first diameter of the acceleration chamber, a water-cooled titanium target disk having a target surface orthogonal to the axis of the acceleration chamber, the target disk having a diameter substantially smaller then a diameter of the isolation well, positioned at a lower extremity of the isolation well, the target disk biased to a substantial negative DC voltage, and electrically grounded metal cladding covering all otherwise exposed surfaces of the pre-moderator block, and mechanically adjustable carriers for the neutron generators, each carrier supporting one neutron generator, and enabled to translate the neutron generator toward and away from the central treatment chamber, and to rotate the neutron generators in a plane parallel to a plane of the parallel upper and lower surfaces of the moderator chamber. The modular generators and the mechanically adjustable carriers are fully immersed in the liquid or granular moderator material of the moderator chamber. In the following descriptions reference is made to the accompanying drawings that form a part hereof, and in which are shown by way of illustration specific embodiments in which the invention may be practiced. It is to be understood that other embodiments may be utilized, and structural changes may be made without departing from the scope of the present invention. Uniform Delivery of Thermal Neutrons to the Cancer Sites To achieve extremely high thermal neutron fluxes uniformly distributed across a patient's head, for example, a hemispherical geometry is used in one embodiment of the invention. This unique geometry arranges fast neutron sources in a circle around a moderator whose radial thickness is optimized to deliver a maximum thermal neutron flux to a patient's brain. This embodiment produces a uniform thermal neutron dose within a factor of 1/20th of the required fast neutron yield and line-voltage input power of a conventional planar neutron irradiation system. This arrangement permits using a relatively safe deuterium-deuterium (DD) fusion reaction (no radioactive tritium) and commercial high voltage power supplies operating at modest powers (50 to 100 kW). FIG. 2 is a cross sectional view of a hemispheric neutron irradiation system 36 according to one embodiment of the invention. Multiple fast neutron generators 68 surround a hemispheric moderator 34, which in turn surrounds the patient's head 26. Titanium targets 52 are distributed around the perimeter of the hemispheric moderator 34. Surrounding the moderator 34 and the fast neutron generators 68 is a fast-neutron reflector 44. In the moderator 34, moderating material such as 7LiF, high density polyethylene (HDPE), and heavy water are shaped in a hemisphere that is shaped around the head of the patient. The optimum thickness of the hemispheric moderator for irradiation purposes is dependent upon the material's nuclear structure and density. FIG. 3 shows a perspective view of a patient 58 on a table 54 with the patient's head inserted into hemispheric irradiation system 36. The patient 58 lies on the table 54 with his head inserted into hemispheric moderator 34. Surrounding the moderator is neutron reflecting material 44, such as lead or bismuth. Referring again to FIG. 2, fast neutrons 22 are produced by fast neutron generators 68. Generators 68 are composed of titanium targets 52 and ion sources 50. Ion beams are produced by ion sources 50 and accelerated toward titanium targets 52 which are embedded in hemispheric moderator 34. A DD fusion reaction occurs at the target, producing 2.5 MeV fast neutrons 22. The fast neutrons 22 enter the moderator 34 wherein they are elastically scattered by collisions with the moderator atom's nuclei. This slows them down after a few collisions to epithermal neutrons 24 energies. These epithermal neutrons 24 enter the patient's head 26 wherein they are moderated further to thermal neutron 30 energies. These thermal neutrons 30 are then captured by boron-10 nuclei at the cancer site, resulting in a fusion event and the death of proximal cancer cells. Fast neutrons 22 are emitted isotropically from titanium target 52 in all directions. Outwardly traveling fast neutrons 42 are reflected back (reflected neutron 48) by fast neutron reflector 44, while inwardly traveling fast neutrons 40 are moderated to epithermal energies and enter the patient's head 26, where further moderation of the neutrons to thermal energies occurs. A shell of protective shielding 56 is also shown in FIG. 2. In some embodiments, this may be necessary for shielding both the patient and the operator from excessive irradiation due to neutrons, x-rays and gamma radiation. The shielding can be made of a variety of materials depending upon the radiation components one wishes to suppress. In some embodiments, fast neutron reflector 44 is made of lead or bismuth. The fast neutron reflector also acts as a shielding means to reduce emitted gamma rays and neutrons from the hemispherical neutron irradiation system 36. As one skilled in the art will realize, gamma-absorbing or other neutron reflector means can be placed in layers around the hemispherical neutron irradiation system 36 to reduce spurious and dangerous radiation from reaching the patient 58 and the operator. Hemispheric moderator 34, fast neutron reflector 44 and head 26 act together to concentrate the thermal neutrons in the patient's head. The patient's head and the moderator 34 act in concert as a single moderator. With a careful selection of moderating materials and geometry, a uniform dose of thermal neutrons can be achieved across the patient's head and, if a boron drug is administered, a large and uniform therapeutic ratio can be achieved. The invention gives a uniform dose of thermal neutrons to the head while minimizing the fast neutron and gamma contributions. The required quantity of fast neutrons to initiate this performance is reduced compared to that of prior art planar neutron irradiation systems (see FIG. 1). A cross section perspective view of the hemispheric neutron irradiation system 36 in an embodiment of the invention is shown in FIG. 4. This cross-section view is of a radial cut directly through the patient's head 26 and hemispherical neutron irradiation system 36. As shown in this embodiment, ten fast-neutron generators 68 composed of ion sources 50 with titanium targets 52 are radially surrounding the hemispheric moderator 34 and the patient's head 26. The titanium target 52 in this embodiment is a continuous belt of titanium surrounding the moderator 34. The titanium targets can also be segmented, as was shown in FIG. 2. The ion sources in this embodiment are embedded in fast neutron reflector 44. There are a number of materials one could select for the moderator 34 to achieve maximum thermal neutron flux at the patient's head 26. The performance of HDPE, heavy water (D2O), graphite, 7LiF, and AlF3 was analyzed using the Monte Carlo Neutral Particle (MCNP) simulation. In general, there is an optimum thickness for each moderator material that generates the maximum thermal flux at the patient's head (or other body part or organ). The thermal neutrons/(cm2-s) was calculated for these materials as a function of moderator thickness d3, where d4=25 cm, and fast neutron reflector 44 is d1=50 cm thick and is made of lead. As in all our calculations, the combined fast neutron yield striking the area from all the fast neutron generators 68 is assumed in the MCNP to be 1011 n/s. The optimum thickness, range of thicknesses and maximum thermal neutron flux (E<0.5 eV) are given in Table I for various moderator materials. These are approximate values given to help determine the general dimensions of the moderator. TABLE IModerator ThicknessModeratorOptimumRange of thicknessMaximum FluxMaterialThickness d3 (cm)d3 (cm)(n/cm2-sec)HDPE6 4-107 × 108D2O15 9-252 × 108Graphite2019-209 × 1077LiF2520-303 × 107AlF33020-401.5 × 107 The calculation of the therapeutic ratio is also important and depends upon the organ in question (brain, liver) and the body mass of the patient. Although HDPE gives the highest flux, it gives a lower therapeutic ratio compared to 7LiF. The designer is expected to do calculations similar to this to determine the optimum geometry for the neutron irradiation system. The MCNP simulation was used to determine the delivered dose and therapeutic ratio to the patient 58 and compare it to a planar neutron irradiation system. In one simulation, moderator 34 is composed of 7LiF whose thickness is d3=25 cm. The inner diameter of the moderator (hole for head) is d4=25 cm. The spacing between hemispheric fast neutron reflector 44 and hemispheric moderator 34 is d2=10 cm. The head is assumed to be 28 cm by 34 cm. Fast neutron reflector 44 is made of d1=20 cm thick lead in one embodiment. Thicker values of d1 increase the tumor dose rate. At a thickness of 10 cm, the tumor dose rate is about one-half the value at a thickness of 50 cm. Fast neutron generators 68 are assumed to emit a total yield of 1011 n/sec. The combined titanium targets 52 give a total neutron emission area of 1401 cm2. In the MCNP simulation BPA (Boronophenylalanine) was used as a delivery drug. The concentration of boron in the tumor was 68.3 μg/gm and in the healthy tissue was 19 μg/gm. The calculated neutron dose rates in Gy-equivalent/hr are plotted in FIG. 5 as a function of distance from the skin to the center of the head. The calculated dose rates are comparable to those used for gamma radiotherapy, typically 1.8 to 2.0 Gy per session. For the same dosage, at a rate of 3 Gy-equivalent/hr, the session length would be from 30 to 40 min. long. These session times are considered reasonable for a patient to undergo. For this simulation, the therapeutic ratio for the hemispherical neutron irradiation system is plotted in FIG. 6 as a function of distance from the skin to the center of the skull. The therapeutic ratio is defined as the delivered tumor dose divided by the maximum dose to healthy tissue. A therapeutic ratio of greater than 3 is considered adequate for cancer therapy. The conventional planar neutron irradiation system requires larger fast-neutron yields (1012 to 1013 n/s) to achieve equivalent dose rates and therapeutic ratios. In FIG. 5, a planar neutron irradiation system 14 of FIG. 1 is compared with that of a hemispheric neutron irradiation system 36 (FIGS. 2, 3, 4) in one embodiment of the present invention, using the same source of fast neutrons (1011 n/s). As can be seen from FIG. 5, the hemispherical neutron irradiation system (called radial source in FIG. 5) achieves a dose rate of about a factor of 20 over that of the conventional planar neutron irradiation system 14. The planar geometry needs a fast neutron source of 2×1012 n/s to achieve the same results. Indeed, if a DD fusion generator is used, then the planar source requires a factor of 20× increase in wall-plug power or 2.0 MW, a prohibitively large power requirement. In addition, as can be seen from FIG. 5, over a ±5 cm distance across the head center, hemispheric neutron irradiation system 36 has less than a 10% variation in dosage. A uniform dose rate is crucial for the treatment of GBM, where we want to maintain a maximum therapeutic ratio and tumors may have distributed themselves across the brain. Hemispherical neutron radiation system 36 in embodiments of the invention also gives a more uniform therapeutic ratio (FIG. 6) across the brain. The ratio is more uniform for the radial source and requires only 1/20th of the fast neutron yield of the planar source (FIG. 1). Other materials can be used for hemispheric moderator 34 in alternative embodiments. As those skilled in the art will know, high density polyethylene (HDPE), heavy water (D2O), Graphite and 7LiF can also be used. In addition, combinations of materials (e.g. 40% Al and 60% AlF3) can also be used. Different thicknesses d1 of moderator can be used to optimize the neutron flux and give the highest therapeutic ratio. The term “neutron generator or source” is intended to cover a wide range of devices for the generation of neutrons. The least expensive and most compact generator is the “fusion neutron generator” that produces neutrons by fusing isotopes of hydrogen (e.g. tritium and deuterium) by accelerating them together using modest acceleration energies. These fusion neutron generators are compact and relatively inexpensive compared to linear accelerators that can produce directed neutron beams. Other embodiments depend upon the selection of the plasma ion source that is used to generate the neutrons at the cylindrical target. These are (1) the RF-driven plasma ion source using a loop RF antenna, (2) the microwave-driven electron cyclotron resonance (ECR) plasma ion source, (3) the RF-driven spiral antenna plasma ion source, (4) the multi-cusp plasma ion source and (5) the Penning diode plasma ion source. All plasma ion sources can be used to create deuterium or tritium ions for fast neutron generation. Cylindrical Irradiation System for the Liver and Other Cancer Sites. FIGS. 7A and 7B shows another embodiment of the invention which uses a cylindrical geometry to irradiate other organs and parts of patient 58, such as the liver 76. FIG. 7A is a cross sectional view of cylindrical neutron irradiation system 62 and FIG. 7B is a perspective view of the same embodiment. In this embodiment eight fast-neutron generators 68 surround a cylindrical moderator 46. These generators 68 all emit their fast neutrons at the surface of the moderator. A cylindrical fast neutron reflector 44 surrounds the cylindrical moderator 46. As in the case of the hemispheric moderator 34, the cylindrical moderator 62 can be composed of well-known moderating materials such as 7LiF, high density polyethylene (HDPE), and heavy water. These are shaped in a cylinder that surrounds the patient. The optimum thickness of the cylinder moderator for neutron capture purposes is dependent upon the material nuclear structure and density. In this embodiment, fusion neutron generators are used to supply the fast neutrons. Fast neutron generator 68 is composed of a titanium target 52 and an ion source 50 as before. The titanium targets are contiguous to the cylindrical moderator 46. Ion beams 60 are accelerated using a DC high voltage (e.g. 100 kV) to the titanium target 52 where fast neutrons are produced from the DD fusion reaction. The fast neutrons are emitted isotropically from the titanium targets 52 on the moderator, some moving out to the fast neutron reflector 44 and others inwardly to be moderated immediately to epithermal or thermal energies. Those reflected come back into the cylindrical moderator 46 where they are moderated to epithermal and thermal energies, making their way finally to the patient 58. Cylindrical neutron irradiation system 62 permits uniform illumination of a section of the patient's body (e.g. liver) as compared to the conventional planar neutron irradiation system. In the case of the brain, the body itself acts as part of the moderation process, thermalizing epithermal neutrons coming in from cylindrical moderator 46. As one skilled in the art will realize, other cancers, such as throat and neck tumors, can be effectively irradiated by a hemispherical neutron irradiation system such as system 36. The thickness and material content of the moderator can be adjusted to maximize the desired energy of the neutrons that enter the patient. For example, for throat and neck tumors, the moderator can be made of deuterated polyethylene or heavy water (D2O) to maximize thermal neutron irradiation of the tumor near the surface of the body. For deeper penetration of the neutrons one might make the moderator out of AlF3, producing epithermal neutrons. These would be optimum for reaching the liver and producing uniform illumination of that organ. Segmented Moderator In yet another embodiment, fast neutron sources with segmented moderators may be individually moved to achieve a uniform dose across the liver or other cancer site. This geometry produces a uniform thermal neutron dose with a factor of between 1/10th and 1/20th of the required fast neutron yield and line-voltage input power of previous linear designs. This again permits the use of the relatively safe deuterium-deuterium (DD) fusion reaction (no radioactive tritium) and off-the-shelf high voltage power supplies operating at modest power (≤100 kW). A segmented neutron irradiation system 70 in an embodiment of the invention is shown in FIG. 8. Ten fast neutron generators 68, each with a wedge-shaped moderator 74, surround the patient 58. The exact shape of each moderator can vary and can be of other geometries. Each generator and moderator pair can be moved independently of the others to achieve uniformity of the neutron flux across the liver, organ, or body part. In between the wedge-shaped moderators 74 more moderating material (“filler moderating material” 72) is inserted, forming a large single moderator. The “filler” moderating material 72 can be heavy water or powered moderating materials such as AlF3. Pie shaped fillers of moderating material can also be fitted into the spaces between the wedge-shaped moderator 74. Since neutrons scatter easily, there can be some space between the wedge-shaped moderators 74 and the pie shaped fillers without undue loss of neutron moderating efficiency. The neutron yield from and the position of each fast neutron generator 68 can be adjusted to achieve uniformity across the liver or body part. The position and the neutron yield of the generator can be varied to achieve the desired radiation dose at a particular location in the patient's body. Since the cancer can be located in any part of the body, this benefit can be particularly useful for optimizing the dose at the cancer site. Surrounding the entire fast neutron/moderator system is a cylindrical fast neutron reflector 44. Fast neutrons are produced by the fast neutron generators 68 and enter the moderators 74 where they are elastically scattered by collisions with the moderator atoms' nuclei, slowing them down after a few collisions to epithermal energies. As in the other embodiments, these epithermal neutrons enter the patient 58 and liver 76, wherein they are moderated further to thermal neutron energies. The invention in various embodiments provides a uniform dose of thermal neutrons to the liver, organ or body part while minimizing fast neutron and gamma contributions. The required number of fast neutrons (e.g. 2×1011 n/s) to initiate this performance is again reduced compared to that (e.g. 2×1013 n/s) needed for the planar neutron irradiation system of the prior art. Another embodiment of the segmented design is shown in FIG. 9. The shape of the neutron irradiation system 78 is elliptical, with six sources of fast neutrons shown as distributed targets embedded in the inside elliptical moderator 96. Fast neutrons 22 are emitted isotropically in all directions. Those fast neutrons 22 moving outwardly are reflected back (see arrow 48) by fast neutron reflector 44, while fast neutrons traveling inwardly 22 are moderated to epithermal energies and enter the liver 76, where further moderation of the neutrons to thermal energies occurs. The inside elliptical moderator 96, outside elliptical moderator 98, reflector 44 and patient's body 58 act together to moderate and concentrate the thermal neutrons into the patient's liver 76. With a careful positioning of the moderators and fast neutron sources 90, 92, 94, a uniform dose can be achieved across the patient's liver, and, with a boron drug administered to the tumor, an excellent therapeutic ratio can be achieved. Elliptical neutron irradiation system 78 in FIG. 9 is a simplified cross-sectional view of the patient 58 inside the elliptical moderator 96. This cross-section view is of a radial cut directly through the patient's torso and the moderator and fast neutron generator system. To maintain visual simplicity, only the titanium targets are shown and not the ion sources. Thus, six fast-neutron sources are represented by three flat titanium targets 90, 92, 94. The rest of the fast neutron generator is not shown. Other components (e.g. plasma ion source) are neglected in the analysis. The wedge-shaped moderators 74 (used in FIG. 8) are also not shown in FIG. 9. For a simple simulation of the neutron irradiation system, the targets 90, 92, 94 are the sources of the fast neutrons and are arranged in an elliptical material 96 (e.g. AlF3, LiF). The effect of the moderating material 96, the fast neutron reflector 44 and the patient's body 58 were calculated using a Monte Carlo N-particle (MCNP5) transport code to determine how fast the neutrons were converted to thermal neutrons in the neutron irradiation system. Dosage calculations were made along a central axis of the liver. The fast neutron sources (titanium targets) are 2 cm×2 cm in area, each producing 1011/N n/s, where N is the number of sources. The human body 58 dimensions are 35.5 cm along the major axis and 22.9 cm along the minor axis. The inner elliptical moderator 96 is made of 7LiF and 10 cm thick, while the outer moderator 98 is made of AlF3 and 40 cm thick. The fast neutron reflector 44 is made of lead 50 cm thick. Boron-10 concentration is 19.0 μg/g in the healthy tissue and 68.3 μg/g in the tumor. The six sources are located in cms at: (−15,18.06,0) (−15,−18.06,0) (−17,17,0) (−17,−17,0) (0,15.85,0) (0,−15.85,0). These measurements are made along the axis of the liver 76 from the point (−15,0,0) to (−5,0,0). In the x-direction, the first two sources 90 are centered about the left edge of the liver shown in FIG. 9, the two sources 92 are centered about the edge of the body, and the third two 94 are located above and below the origin. The origin is shown in FIG. 9 as a small cross+at the center of the body in the plane of the liver. FIG. 10 shows the therapeutic ratio for a large single dose, and the therapeutic ratio for multiple small doses (where the photon dose to healthy tissue is not included) plotted as a function of distance along the axis of the liver. The photon dose can be neglected if there is some amount of time between doses. Many of the body's healthy cells can self-repair and recover between doses. The expected therapeutic ratio is between these two curves when there is fractionation into multiple doses. In this simulation, BPA was again used as the delivery drug with the concentration of boron in the tumor at 68.3 μg/gm and in the healthy tissue at 19 μg/gm. FIG. 11 indicates that the goal of having an extremely uniform dosage to the tumor has been achieved, with about ±6% variation along the x-dimension. The calculated dose rates are comparable to those used for gamma radiotherapy, typically 1.8 to 2.0 Gy-equivalent per hour if we increase the total neutron yield to 2×1011 to 3×1011 n/s. Thus, at approximately 2×1011 to 3×1011n/s it is possible to obtain a therapeutic ratio and uniform dosage to a tumor. Approximately 10 to 20 treatments of 30 to 40 minutes would be required, with a good therapeutic ratio, uniformity of dosage, and the opportunity for healthy tissue repair between treatments. Once again, the planar neutron irradiation systems require high fast neutron yields to drive them. In one prior art system known to the inventors a fast neutron source of 3×1013 n/s is needed to obtain realistic treatment time of ˜1-2 hours. Using a D-T neutron source with a yield 1014 n/s, acceptable treatment times were obtained (30 to 72 minutes with single beam and 63 to 128 minutes with 3 beams of different direction). But these are impossible yields to achieve with realistic wall plug powers. Instead of 50 to 100 kW for the hemispheric and cylindrical neutron irradiation systems, it would take a minimum of 0.5 MW to achieve adequate yield for the planar geometry with a DT generator. These are high powers for clinics and hospitals. As one skilled in the art knows, other cancers, such as throat and neck tumors, can be effectively irradiated by the neutron irradiation system. The thickness and material content of the moderator can be adjusted to maximize the desired energy of the neutrons that enter the patient. For example, for throat and neck tumors, the moderator can be made of deuterated polyethylene or heavy water (D2O) to maximize thermal neutron irradiation of the tumor near the surface of the body. For deeper penetration of the neutrons one might make the moderator out of AlF3, producing epithermal neutrons. These would be optimum for reaching the liver and producing uniform illumination of that organ. Modular Generators As is shown in FIGS. 8 and 9, multiple modular generators may be encased in moderator material and may be arrayed to maximize thermal neutron flux at a cancer tumor location. Fast 2.5 MeV neutrons must be slowed (moderated) to energies (usually epithermal) that will penetrate to the cancer site without too many neutrons being lost in their travel to the cancer via capture by healthy tissue. These modular generators act as independent neutron sources and each may be optimized by adjustment of each individual beam's energy, direction and intensity. The modular generators can be arranged to fit a site in a particular subject's component location and structure. This is true also for cancer tumor location. The energy of the neutrons can also be adjusted by adding or subtracting moderator material. This can be done more easily than with a single beam LINAC or reactor, which usually has a fixed beamline that is integral to the neutron source. In the prior art some adjustment can be made, but the DD fusion generator in embodiments of the invention, being much smaller, can have more degrees of freedom in direction, intensity and moderation. This has an added benefit of aiding physicians in tailoring neutron radiation to the patient's cancer. Comparison to Linear Accelerators and Reactors. Modular generators in various embodiments of the present invention may also form and be part of the mechanical structure of a cancer irradiation system. This has an added benefit of moving the neutron sources as close as possible to the cancer site and the diseased body part, resulting in efficient use of the neutron source. The neutrons are being emitted in a 4π solid angle from the modular generators, so the closer to cancer site, the more of the fast neutron flux is being utilized. Linear accelerators (LINACs), which are somewhat collimated, are further from the cancer site and cannot provide this advantage. Compared to a linear accelerator, which can be several meters long or longer and may include large microwave power sources, the DD fusion sources in embodiments of the invention are less than one meter long and comprise compact microwave sources that can either be solid state microwave sources or small, inexpensive, single microwave oven magnetrons. The accelerator structure in embodiments of the invention is compact and includes a pre-moderator 118 that adds only from 5-10 cm of High-Density Polyethylene (HDPE) or 15-20 cm of polytetrafluoroethene (PTFE) Teflon to produce a first stage of neutron beam tailoring. The pre-moderator in these embodiments is an integral part of each modular generator, as is taught below with reference to several figures. In alternative embodiments other pre-moderator materials can be used such as AlF3, MgF2, 7LiF, and Fluental (trade name). Smaller, Nontoxic, Less Complex Targets for Neutron Production The modular DD fusion generator 118 in embodiments of the present invention uses a small titanium target (e.g. a 5 cm diameter disk of titanium backed by water-cooled copper fins) to produce neutrons. The target is supported directly on the pre-moderator, which is an integral part of the apparatus in this application, termed a modular generator. Linacs and other methods in the conventional arts use larger or toxic targets that require complex cooling and rotation. For example, the neutron source used by Neutron Therapeutics has a 2.6 MeV electrostatic proton accelerator and a rotating, solid lithium target for generating neutrons. In that prior-art process the Lithium becomes radioactive and toxic, and when exposed to air, it disintegrates. This prior art source has a large target chamber housing a large Li disk which is rotated in a powerful 2.8 MeV proton beam produced by a large accelerator. The Lithium wheel is roughly 2 meters in diameter and has been divided into pie-shaped sections that are removed by mechanical robotic means. In embodiments of the present invention, the Ti target is a relatively small diameter (˜5 cm) and is typically attached with 6-8 screws to the pre-moderator block and is sealed to the block with a Viton “O” ring. The Ti targets in embodiments of the invention can be easily manually removed and replaced. They also have a long lifetime and have been tested for over 4000 hours with no failures. Nuclear reactors are large structures with a substantial amount of shielding (water and concrete) and cooling systems to maintain the hot reactor core. Reactors provide primarily thermal neutrons that must be raised up in energy using an energy multiplier, and then the neutron beam must be improved to IAEA standards to produce epithermal neutrons with minimal gamma radiation. Optimizing Neutron Energy for Penetration and Minimum Damage to Healthy Tissue For tumors at depths in a subject of 3 cm or more, a goal for the moderator is to provide a neutron beam that has its energy clustered about 10 keV at the skin, in order to provide sufficient energy to penetrate a minimum of several centimeters into a human target while avoiding higher energies that are more damaging to human tissue. High conversion to epithermal energies occurs in HDPE at a thickness of approximately 5 cm, but it also produces a high yield of thermal neutrons and 2.2 MeV gammas that can damage the healthy tissue at the skin. Modular Generators In embodiments of the present invention modular generators are very important components. The modular generator combines multiple functions that were separate functions in the prior art. These integrated functions include both neutron production and beam tailoring. FIG. 12A is a perspective view of an individual modular generator 118 in an embodiment of the invention. FIG. 12B is a cross section of the modular generator 118 of FIG. 12A taken along an axis of an acceleration chamber 100 for ion beam generation and containment, and at a right angle to the axis of a turbo vacuum pump 124 that is part of the modular generator 118. FIG. 12C is a cross section of the modular generator 118 of FIG. 12A taken along the axis of the acceleration chamber 100, and along the axis of the turbo vacuum pump 124, at a right angle to the section of FIG. 12B. Each modular generator 118 can operate independently of the other modular generators and each possesses all required components to generate neutrons. Further, the various modular generators may have pre-moderators shaped to engage other building blocks of a project, such as adjacent generators or spacing moderators, as is described in enabling detail below. Viewed as in FIGS. 12A, B and C, each modular generator 118 comprises a pre-moderator 108 that is made of material known to moderate energy of energetic neutrons. In most embodiments the pre-moderator is a solid block of material, with a rather complicated shape for certain purposes. Modular generator 118 has three key elements: (1) a deuterium ion source 102, (2) an acceleration chamber 100, through which deuterium ions may be accelerated, and (3) a titanium target 106 (shown in FIGS. 12B and 12C) that is bombarded by the deuterium ions to produce high-energy neutrons. The deuterium ion source 102 has an attached microwave source 160, and microwave slug tuners 172, connected by a cable 178. Deuterium gas is leaked slowly into a plasma ion chamber 174 at the upper end of the acceleration chamber, where microwave energy ionizes the gas, creating deuterium D+ ions. The gas is ionized by microwave energy, and Deuterium (D+) ions are created and accelerated out through an ion extraction iris 138 into acceleration chamber 100, and through an electron suppression shroud 180 which deflects back-streaming electrons from being accelerated back into the plasma source, which could damage the apparatus. Electrons are being created by collisions of the D+ ions in the deuterium gas that are being created in the acceleration chamber. The deuterium ions are positively charged, and target 106 is negatively charged to a level of from 120 kV to 220 kV, and the D+ ions are strongly attracted to negatively biased target 106. Acceleration chamber 100 is connected to a turbo vacuum pump 124 that provides a modest vacuum in one embodiment of about 10−6 Torr, minimizing scattering of the D+ ions as they travel from the extraction iris 138 to the target 106. Titanium target 106 is positioned in a primary electrically insulating well 181 at the bottom of the chamber embedded into the pre-moderator material, which may be UHMW, HDPE or Teflon, of the pre-moderator 108. There is further a secondary electrical insulating well 182 surrounding the primary electrical insulating well. The surface of the moderator material in the primary and secondary electrical insulating wells may be seen as a corrugated insulator causing any surface charge to follow a curved path taken in any direction. The purpose is to provide a very long surface path to prevent electrons from traveling from the target to acceleration chamber 100 wall or any grounded element, and to avoid surface electrical breakdown or flashover in that surface path. As those skilled in the art know, the wells form an electrical insulating path. Additional corrugations or wells can be added to lengthen the path. Pre-moderator 108 has a high voltage bus bar 122 and fluid cooling channels 120 to and from the target. The high voltage is introduced via a high voltage receptacle 130 which is connected to the high voltage bus bar. Pre-moderator 108 acts as a HV insulator and as a mechanical support for the target 106 at a high negative bias. The pre-moderator 108 has metal cladding 140 at ground potential to minimize high voltage breakdown through the pre-moderator plastics. When in operation the D+ ions in the ion beam are attracted to the titanium target 106, where fast (2.5 MeV) neutrons are produced in a resulting DD fusion reaction. FIG. 13A illustrates an assembly of six modular generators 118, wherein pre-moderators 108 are spaced apart by spacers 128 which are also made of moderator material. FIG. 13B shows the arrangement of FIG. 13A in perspective. FIG. 13C shows the arrangement of FIG. 13B with one modular generator 118 removed from the assembly. FIG. 13D is a more diagrammatic illustration showing an arrangement in which modular generators may be mounted on translation and rotation mechanisms to be positioned to maximum irradiation of a cancer site. As is shown in FIGS. 13A-D the modular generators in embodiments of the invention may be arranged in an array to form a complete and moveable system of irradiating neutron sources with pre-moderators. For example, as shown in FIG. 13A-C, in the simplest configuration of the array, the modular generators may form a circle around a human torso or body part. The modular generators can be moved into three dimensional arrays around the subject to maximize neutron flux to a cancer site 148 that may not be centered on a body part 146, illustrated as a human brain in FIG. 13D. Thus, depending upon body contour, shape and size, and cancer location and distribution, the modular generators may be moved to adapt to the shape and tumor location in order to maximize the dose to the cancer and to minimize the dose to the other body parts. Referring to FIG. 13D, rotation 150 and translation 151 of the modular generator 118 can be achieved with electrical motors attached to the modular generator 118. Seven Functions of the Pre-Moderator Because the titanium target is on the pre-moderator (first stage of moderation), fast neutrons coming from the target immediately enter the pre-moderator and quickly moderated to thermal or epithermal energies. The pre-moderator also provides mechanical support, high voltage supply and cooling fluid transport to the titanium target. Exemplary pre-moderator materials that may accomplish this are Teflon and HDPE. Both Teflon and HDPE are excellent high voltage dielectrics which can also support a HV bus bar 122 and water channels 120 to be used to transport HV and the cooling fluids to the Ti target, as shown in FIG. 12C. As shown in FIGS. 12A, B, C a single generator 118 consists of an acceleration chamber 100, an ion source 102 emitting deuterium ions, a titanium target 106 and a pre-moderator 108. Pre-moderator 108 also provides a function of being a high voltage insulator for high voltage bus bar 122 that delivers high voltage (e.g. 80 kV to 300 kV)) to titanium target 106, and a water channels 120 that deliver cooling fluid to the titanium target 106. The high voltage is delivered from a high voltage power supply through a standard HV receptacle 130 to the bus bar 122 and then on to the titanium target 106, all of which are mounted in the pre-moderator 108. In various embodiments of the invention the pre-moderator 108 performs seven functions: (1) moderation, (2) mechanical support of the titanium target, (3) cooling fluid transport to the target, (4) high voltage transport to the target, (5) minimum surface flashover, (6) and a portion of a high vacuum container (a wall) with no out gassing (7). These seven attributes permit a substantial reduction of distance and amount of material between the fast neutron source and the patient, thus helping to maintain a maximum neutron flux delivered to the patient. Modular Generators Around a Subject FIGS. 13A-D show how the generators may be arranged. In FIG. 13A, six modular generators 118 form a ring around a secondary moderator 112 and are part of a structure formed by secondary moderator 112, spacers 128, and pre-moderators 108. Pre-moderators 108 and secondary moderator 112 provide the moderation function by slowing the neutrons down to epithermal energies (function #1). These elements also form a mechanical support (function #2) for the entire generator and moderator system. Secondary moderator 112 may also be a separate section attached directly to the modular generator just after the pre-moderator, each separate from the other instead of being in a ring 112 as in FIG. 13A. As shown in FIG. 12B-C, fluid transport (function #3) is supplied through channels 120, which delivers cooling fluid to target 106 to maintain the target at an acceptable operating temperature. Each generator is supplied with a separate cooling fluid input and output, wherein cooling fluid is provided through a connector 132 shown in FIGS. 12A-12C. Thus, the pre-moderator supplies fluid transport (function #4). High voltage is delivered via high voltage bus 122, which passes through pre-moderator 108 (function 4, high voltage transport). HDPE, UHMW and Teflon are excellent insulators and withstand high voltage flashover (function #6). All three may be used in vacuum systems without excessive out gassing and may help maintain the system vacuum (function #7). The achievement of these seven functions provides a very compact and flexible neutron source. The Secondary Moderator Secondary moderator 112 (FIGS. 13A-C) may comprise any one of or a combination of multiple moderator materials that optimize both the maximum flux and neutron energy for maximum dose to the cancer site. Selection (material, size and shape) may be varied depending on depth of the cancer in the subject and a desired dose at the cancer site. The secondary moderator may be D2O (heavy water) for delivery of thermal neutrons to, for example, throat and neck cancers, or a combination of AlF3 and Teflon for delivery of epithermal neutrons to brain tumors. The recommended levels of fast, thermal and gamma emission by IAEA are given in Table I. TABLE 1IAEA Recommended values in the beam exit window.IAEA RecommendedBNCT beam port parametersvalueϕepithermal (n cm−2 s−1) ~109ϕepithermal/ϕfast >20ϕepithermal/ϕthermal>100Dfast/ϕepithermal (Gy cm2)<2 × 10−13Dγ/ϕepithermal (Gy cm2)<2 × 10−13Fast energy group (ϕfast)E >10 keVEpithermal energy group1 eV ≤ E ≤ 10Thermal energy group (ϕthermal)E <1 eV These IAEA recommended values depend upon older drugs, such as p-Boronophenylalanine (BPA) that have been approved for use in humans by the Food & Drug Administration (FDA) for other medical applications. Delivery of higher boron concentrations to a cancer site may depend to some extent on newer drugs to be developed, and may permit lower power, less efficient neutron beams to be used. Since treatment time might also be faster, the neutron beam quality need not be as high. DD fusion generators in embodiments of this invention have relatively low beam flux, thus permitting them to be used for cancer therapy. In some embodiments multiple modular generators may be distributed around a secondary moderator surrounding a central chamber holding a subject for treatment, providing an alternative to a completely integrated multi-ion beam system, and may have particular benefits in some circumstances. Benefits might include (1) an ability to quickly replace a single generator that has failed and needs repair; and (2) an ability to change alignment of the generators relative to one another, the moderator, and the subject. In regard to a subject, alignment of the generators may optimize dose distribution and density of neutrons at a cancer site, while at the same time minimizing spurious radiation, such as gamma rays that might be emitted external to the apparatus, or into healthy tissue of the subject. In the prior art, where reactor and accelerator neutron sources are used, careful attention has been given to achievement of high quality neutron beams to meet the IAEA standards for BNCT developed in 2001 for International Atomic Energy Agency (IAEA) (Current Status of Neutron Capture Therapy (2001) IAEA-TECDOC-1223. In embodiments of the present invention, where multiple modular DD fusion generators are used, these standards may be relaxed. The IAEA specification assumes that there is a single neutron beam that is used for all cancers and body locations. This results in standard values for the three neutron energies (thermals, epithermal and fast neutrons). Moderator and neutron spectral shifters are then designed to achieve these values for a particular fast neutron source as an input specification. This results in designs in the prior art that may not use the available fast neutrons economically and then may waste some of them to achieve the IAEA universal specs. For generators such as the DD fusion source in an embodiment of the present invention, early calculations have indicated that a single DD fusion generator would have difficulty achieving required fast neutron input to the moderating process. So, in embodiments of the invention, the use of multiple generators increases the total fast neutron yield available and allows the moderated dose to be distributed over a larger area of the body, instead of having the beam enter at one location of the body. For example, as shown in FIG. 13D, neutrons n are entering the head from many directions. This permits reduction of thermal neutron flux at any one point on the skin of the head while still achieving adequate epithermal flux to the cancer site. In early prior art reactor BNCT experiments, the thermal neutron flux burned the skin of subjects. When considering neutrons used for a particular cancer it is desirable to direct the maximum flux to the cancer site, and therefore, one must consider the specific cancer that is to be treated. This includes location and depth in the human body. Because of their relatively small size and large neutron yield, the modular generators in the embodiments of the present invention are particularly able to accomplish this by being positioned to maximize their flux at the cancer site. Since in embodiments of the invention generators are placed as close to the patient's body as practical to maximize flux at the cancer site, there is a more holistic problem. There are multiple parameters for each modular generator: (e.g. neutron flux, neutron energy, position relative to the body). What comes out of a single neutron beam pipe (1998 IAEA Standards, Table I) is not the only concern. A body part can now, in new implementations of the invention, be irradiated in all directions, and neutron intensity can be adjusted at each modular generator to achieve better flux and even more optimum neutron energy than with a single beam LINAC or a reactor. The direction of each neutron beam can be adjusted by rotating and displacing each modular generator 118. Each modular generator's yield can be adjusted electronically by varying the accelerator voltage and the ion beam current. Since the moderator size is relatively small and compact compared to the prior art, the neutron spectrum of each modular generator 118 can be adjusted by the selection of different moderator materials and thicknesses. Lowering of Required Beam Quality In embodiments of the present invention the subject's body is bombarded with neutrons from multiple directions. The neutrons can come from all sides of the body part, which minimizes the amount of distance each beam has to transverse. Unwanted neutrons striking the skin are now distributed over a larger area, reducing the skin dose of harmful components (e.g. gammas, and thermal and fast neutrons) per unit area. These components are simply delivered over a larger area of the skin. This permits adjustment of dose at the cancer site to be higher than that achieved with a single beam but with reduction of harmful components over a larger area of the skin. For a single beam case in the prior art, an argument might be made that one can rotate the patient for each exposure, but, due to possible patient movement, the neutrons would not be as accurately placed as in multi-beam embodiments of the present invention. For each placement the patient would have to be carefully re-oriented relative to the single neutron beam, which requires careful placement of the patient. In embodiments of the invention, multiple beam directions and an ability to adjust the neutron flux of each modular generator allow for optimum delivery to the cancer site while reducing harmful components. For example, if the cancer is located in the left lobe of the brain, the neutron flux to the tumor can be adjusted to deliver epithermal neutrons in the direction of that tumor. Since each modular generator neutron flux can be adjusted quickly by varying the accelerator's high voltage or the ion beam current, and by translation and rotation, this can be done easily with delivery determined by a computer program. In the present invention, a control computer monitors the ion beam current, the acceleration voltage and the output neutron yield, which can be automatically adjusted. Small modular generators in embodiments of the invention can make use of new boron drug delivery methods for higher concentrations of boron to the cancer sites. Higher concentrations of boron lower the required neutron dose and require shorter delivery time. Higher boron concentrations to the cancer site permit use of neutron generators with lower neutron yield such as the modular DD fusion generators in embodiments of the present invention. Each modular generator 118 is an independent device capable of producing neutrons independently of the other generators. This allows the total available power, P, to be distributed over N generators, resulting in the heat load being distributed safely without, for example damaging the titanium targets (unlike single target devices using lithium). In one example there are six modular generators, distributing total heat load per titanium target, since the number of neutrons per unit area is fixed by the ion beam power per unit target area. To properly treat a tumor in a subject, a large number of neutrons is required. For reasons of temperature management and stability, DD fusion generators are at present limited to fast neutron yields of less than 4×1010 n/sec. To increase the neutron yield, the number of neutron generators can be increased in embodiments of the present invention. Pre-moderators 108 can be shaped so that larger numbers of modular generators may be fitted around a subject to be treated. In the example shown by FIG. 13A there are six generators arranged equally spaced around a common secondary moderator 112, the subject cavity 116 and the subject 134. Spacing blocks 128, composed of moderator material that may be the same as that of pre-moderator 108 (e.g. Teflon or polyethylene), are placed between each pre-moderator to provide adequate spacing for fitting the subject cavity 118. The wedge angle, α, as indicated in FIG. 12A, on the pre-moderator in FIG. 13A determines the number of modules 118 with pre-moderators 108 that can fit in the circle around the patient and how close the sources may be to the patient. For example, a wedge angle of α=30° for 6 generators and α=22.5° for 8 generators. Moveable Sources with Fluid Moderator One embodiment of a system of modular generators is shown in FIGS. 13A and 13B. In FIG. 13A a plane view of six modular neutron generators 118 fitting into the cylinder (or ring) is shown. In FIG. 13B, a perspective view is shown. The modular generators can also be arranged in other patterns to maximize the dose in particular locations in the subject's body and deliver cancer therapy to selected body organs. In some embodiments of the invention the modular generators may be moved by electric motors and mechanical means to optimized locations to provide the maximum dose to the cancer site and tumor profile as determined by boron bio-distribution test biopsy and pathological analysis, Positron Emission Computed Tomography (PET-CT), Computed Tomography (CT) or magnetic resonance imaging (MRI). One may make use of moderating materials between movable modular generators. For clinical systems there should be moderator material between the modular generators. Ideally the material can quickly position itself to the new location of the modular generators and also be a moderating material. As shown in FIG. 13D, liquid moderator 156 can be used to surround the modular generators 118, acting as a secondary moderator. The moderating material is shown between the movable modular generators. The liquid is contained in an appropriate liquid container. Liquids that also have good moderating properties can be used and are easily displaced by the modular generators when moving. For example, different grades of 3M™ Fluorinert™ Electronic Liquid (e.g. FC-40), which is non-conductive, thermally and chemically stable fluid, can be inserted between generators. Like Teflon it contains primarily fluorine atoms, making it an excellent moderator, and no hydrogen. Stages of Moderation Use of multiple modular generators in embodiments of the invention permits efficient use of modulator material, reducing size of moderator and shielding material and, thus, the reduction and size of the entire system. It also reduces the required flux density of fast neutrons by bringing the neutron sources closer to the patient and directing the limited number of neutrons to the cancer site in a more efficient fashion. The subject's body also becomes part of the equation of the moderating process. The fact that the neutrons are coming from multiple directions reduces local skin dose and localized body dose of healthy tissue. Rather than coming into the body at one location, the neutrons are coming from roughly 360 degrees around the body. Moderation of fast neutrons in embodiments of the invention is a three-step process. In a first step (1) the pre-moderator 108 acts to reduce energy of the fast neutrons in as short a distance as practical with a minimal amount of gamma radiation produced in the process. The pre-moderator also serves as a medium to (2) transport high voltage and (3) cooling fluid to a fast neutron production titanium target 106. Combining these three functions ((1) moderation, (2) fluid transport and (3) high voltage transport) reduces distance and the amount of material between the fast neutron source and the patient, helping to maintain a maximum neutron flux finally delivered to the patient. Partially slowed neutrons can then pass into the secondary moderator 112 which continues the slowing process without undue production of gamma rays from, for example, hydrogen. In the case of small animal models, the selected moderator may be heavy water (D2O). Neutron energy reduction is continued by the D2O without the generation of ˜2.2 MeV gammas that would occur if materials composed of hydrogen were used. For the case of irradiating tumors of depth greater than 3 cm in a human body, the neutrons need to be moderated to epithermal neutron energies. The human body also acts as a partial, final moderator. Thus, the epithermal energy neutrons are slowed further as they move through the body, and finally are slowed to thermal energies at the tumor site. Those skilled in the art will understand that the moderation is a statistical and random process that reduces the neutron energy with a variation or spread of the neutron energies. The process can also result in undesired gamma ray components (e.g. 2.2 MeV gammas from hydrogen capture of neutrons) which damage health cells. In embodiments of the invention, selection of the moderator material depends at least in part upon the desired energy of the neutrons at the body's skin to achieve maximum penetration to the cancer site while reducing (1) excess thermal energy components at the skin, (2) the cost and availability of the moderator material, and (3) harmful gamma ray components. Each generator's energy, yield, direction and moderation can be determined from moderation materials, the generator's voltage and acceleration current. Unlike in the prior art, dimensions of the moderator and content may be quickly changed. In some embodiments of the invention a liquid moderator (e.g. Fluorinert FC40) or a granular (e.g. AlF3) moderator may be used. The modular generators are positioned in the liquid or granular moderator material, where they are free to move by mechanical means quickly between different cancer sites. In the prior art, the moderators and shields are large, massive and usually fixed relative to a single beam reactor or linear accelerator. The patient is usually moved relative to the fixed neutron source. Using liquid or granular moderator materials permits a more efficient reduction of fast neutrons to epithermal energies while minimizing thermals and fast neutrons. Selection of the pre-moderator material is important for optimum neutron beam quality. Generally speaking, beam quality involves minimization of harmful components of radiation that accompany the production of thermal neutrons at the cancer site but also the minimization of the fast and thermal neutrons at the skin surface. In this process gamma rays are produced and, depending upon the cancer site, fast neutrons must be converted to the right energy so that they penetrate the body and deliver thermal neutrons to the tumor site with minimal irradiation of healthy tissue. Moderating the neutrons to thermal energy can result in the skin being damaged. Indeed, the thermal neutron dose to the skin can be larger than the dose to the tumor. The body itself moderates and absorbs the neutrons as they penetrate the body. Selection of the moderator material requires materials that do not moderate the fast neutrons too quickly to thermal energies. Thermal neutrons can damage the skin, and if hydrogen atoms are present in the moderation process, then damaging gamma rays are also produced. Like the moderator, the human body also moderates and absorbs the neutrons. The desired required depth of penetration depends upon the location of the tumor in the body. Simulations show that penetration of thermal neutrons starting at the skin results in penetration depths of 3 to 5 cm before a large fraction of the neutrons are absorbed. Teflon Moderator for Clinical Machine When used as a Pre-moderator, Teflon (PTFE) can satisfy 6 of the 7 functions listed above. Indeed, on several of the attributes Teflon excels. For example, since Teflon does not have atomic hydrogen, gamma production is avoided, whereas the use of HDPE does have hydrogen and, therefore maximizes the thermal neutron moderation with and added 2.2 MeV gamma ray component. Selection of HDPE as the pre-moderator material results in production of thermal neutrons in a short distance from the Ti target, whereas the use of Teflon results in a slower rate of neutron energy reduction from 2.5 MeV permitting the production of epithermal neutrons for deeper penetration into the human body and no 2.2 MeV gammas. Teflon can have a minimum high voltage in which surface arcs (flashovers or surface discharges) momentarily short out the high voltage, and lead to damage to the Teflon surface and possibly damage to the high voltage power supply. This is primarily a materials problem and not a structural problem (shape of the accelerator and Teflon shape and structure). Surface discharge along solid insulators in a vacuum in high voltage devices determines the maximum voltage between an anode and a cathode. The voltage hold-off capability of a solid insulator in vacuum is usually less than that of a vacuum gap of similar dimensions. O. Yamamoto et. al (Yamamoto, O; Takuma, T; Fukuda, M; Nagata, S; Sonoda, T “Improving withstand voltage by roughening the surface of an insulating spacer used in vacuum,” IEEE TRANSACTIONS ON DIELECTRICS AND ELECTRICAL INSULATION (2003), 10(4): 550-556) has studied a simple and reliable method to improve surface insulation strength of a dielectric such as Teflon, PMMA, and SiO2 by roughening the surface of the dielectric. Some experimental results have revealed that in a vacuum, charging along the surface of an insulating spacer precedes the flashover. The charging takes place through a process in which electrons are released from a triple junction where the cathode, insulator and vacuum meet, and propagate toward the anode, causing a secondary emission electron avalanche (SEEA) along the insulator surface. The dielectric (e.g. Teflon or HDPE) can hold charge like a battery or capacitor and then release it along the surface. This limits the use of plastics such as Teflon and HDPE as insulators and moderators inside the vacuum chamber of the neutron generator's acceleration chamber 100. For short distances across Teflon (10 mm), Yamamoto found that roughing the surface (e.g. with sandpaper or sandblasting) affects the charging of various plastics (such as Teflon and HDPE), which decreases as roughness increases. Yamamoto used roughness up to 37.8 μm but had used lower voltage gradients and smaller dielectric thicknesses (10 mm). Studies in embodiments of the present invention find that larger surfaces (distances e.g. 8 inches) of Teflon can be roughened with roughness values of 5 microns and greater and achieve high voltages of 150-220 kV for distances greater than ˜2 cm without flashover. More importantly, the roughing method gives higher insulation strengths without time-consuming conditioning previously used. This provides a significant advantage and makes generators in embodiments of the present invention operational more quickly. Depending on maximum field strength required, conditioning by the roughing process could take minutes or days. Below 1 MV m−1, the conditioning process is relatively fast. Between 1 and 10 MV m−1, the conditioning process takes longer. The best way to monitor how conditioning is going is to record the number of transient discharges (or sparks) per hour. At very high fields the arc rate might never get better than a few arcs per hour. A tolerable arc rate depends on the application. If no high voltage breakdown (arcs) can be tolerated, then the system must first be conditioned to a higher field, and then when the voltage is reduced to the operating level the arc rate drops almost to zero. For very high field strengths above 10 MV m−1, it is very difficult to condition the electrodes to give an arc rate of zero. The electrode shape and material composition becomes very important at these field levels. The Importance of the Human Body in the Moderation Process The human body acts as a moderator to reduce the epithermal neutrons to thermal energies at the cancer site. The amount of neutron energy reduction by the human body depends at least in part upon the depth of the tumor in the body. This determines the maximum neutron flux for delivery to the patient. The desired reduction of the neutron's energy will depend upon the depth of the tumor in the human body. For example, with throat and neck cancers the reduction of the neutron energy to thermal energies is desired for maximum dose to the cancer site. For small animal models, thermal energies are also desired. Dimension in the body from the skin (epidermis) to the cancer site can vary, requiring the neutron energy to be large enough for penetration to the cancer while still primarily at thermal energies, permitting capture by the boron and the destruction of cancer cells. For small animal models or skin cancer in humans, the neutrons can be at thermal energies. For cancers at deeper depths in the body, epithermal neutrons (0.025 to 0.4 eV) can be used. For deep tumors in the torso, such as, for example, pancreatic tumors, epithermal neutrons are required. Pancreatic tumors are deep in the torso and require epithermal neutrons at entrance to the body to penetrate to the tumor. Moderation of the epithermal neutrons occurs as they pass though the body. Simulations in various embodiments show that there are materials at the right thicknesses, such as Teflon, 7LiF and AlF3, which produce the epithermal neutrons that penetrate the body and thermalize by the time they reach the depth of the tumor with a maximum neutron flux. In embodiments of the invention, this occurs while minimizing production of thermal neutrons at the skin. Shape of a Clinical Machine to Match a Human Body The shape of the patient's chamber in a machine may be contoured to fit the human body part to maximize radiation to the cancer site. The shape depends upon the body part to be irradiated and the location of the tumor. As shown in FIG. 13D, for glioblastoma 148 (brain cancer), modular generators 118 may be arranged in a close ring around the head 146 that maximizes neutron flux to the cancer site 148 in the brain. The intensity of each generator can be varied to achieve maximum thermal neutrons to the tumor while minimizing the dose to healthy tissue. As discussed above, applications in embodiments of this invention permit control of the distance of each generator from the cancer site. The cancer site may be mapped using radiographic means (CT scans) and/or MRIs. A treatment planning protocol can then be determined for the optimum use of the clinical neutron source. The intensity of the neutrons coming from each neutron generator can then be varied and the location of each individual generator can be optimized. As shown in FIG. 13 D, an improvement of the moderator surrounding the modular generators is to suspend or surround the modular generators with a liquid 156 that does not contain hydrogen (a gamma producing source), but has modest atomic-number atoms like Fluorine, Carbon or Nitrogen. Various kinds of Fluorinert (tradename), FC-70 or FC-40, or FC3839 can be used. The fluid may be put between the modular generators and by mechanical means each modular generator can move independently of the other generators to a certain extent. This fluid can also absorb some heat from modular generators. As shown in FIG. 13 D, an improvement of the moderator surrounding the modular generators is to suspend or surround the modular generators with a liquid 156 that does not contain hydrogen (a gamma producing source), but has modest atomic-number atoms like Fluorine, Carbon or Nitrogen. Various kinds of Fluorinert (tradename), FC-70 or FC-40, or FC3839 can be used. The fluid may be put between the modular generators and by mechanical means each modular generator can move independently of the other generators to a certain extent. This fluid can also absorb some heat from modular generators. Generator Alignment In embodiments of the present invention each stand-alone generator, as seen in FIG. 13D, for example, may be positioned and aligned to give a maximum flux and neutron distribution at the cancer site. Each generator is small enough in size and weight that the generators may be mechanically moved and positioned so that optimum neutron flux at the cancer site is achieved, depending upon the cancer's location and distribution. The generators may be arranged around a moderator whose radial thickness is optimized to deliver a maximum thermal neutron flux to the cancer site. Depending upon the body part being irradiated, the geometry can be circular or elliptical. By selecting the moderating material and radial thickness one can deliver thermal neutrons to the cancer site. FIG. 14A shows an on-axis view of an exemplary clinical neutron source using multiple modular generators 118 for BNCT of a human head. This example uses eight modular generators 118 and assorted moderator materials coupled with reflecting and shielding material (e.g. graphite 144). Secondary moderators (166 and 170) can be composed of one or more materials. There are moderator spacing blocks 128 in one embodiment composed of the same material High Density Polyethylene (HDPE), Ultra High Molecular Weight polyethylene (UHMW), or (PTFE (Teflon)) as the secondary moderators. Blocks of these materials fit in between the modular generators and are adjacent to each generator's pre-moderator. They act as mechanical spacers as well as moderator components. The outside of this region, between and behind the modular generators 118, is filled with either graphite or lead 144 to serve as a neutron reflector and shield. FIG. 14B also shows a side section view of the apparatus of FIG. 14A taken along a line through the top and bottom generators. There is additional moderator material in the front and behind the modular generators, extending a little above the pre-moderator. In our example, the cylindrical space 164 available for the patient's head is 52 cm deep and 30 cm in diameter. This space might be lined with 1-mm of cadmium 162 to shield against too large a thermal neutron dose to the patient's skin. Shield 162 is also shown in FIG. 14A. In other embodiments this space may be lined with 6LiF. The exemplary arrangement as illustrated in FIGS. 14A and B has a secondary moderator consisting of multiple layers of 40% Al and 60% AlF3 (166) and an additional moderating cylinder 170 of either 7LiF or D2O. These materials are shown to be concentric rings in FIG. 14A. Since 7LiF or D2O can be expensive, thicknesses were varied to obtain a desired neutron beam quality without over-using either 7LIF or D2O. In the example shown in FIGS. 14A and 14 B the thickness ratio between the two segments is altered, the total moderator thickness is 34 cm, and the sources are R=52.5 cm from the origin (center of the brain). The effect of doing this varying these materials is plotted graphically in FIG. 15. The reflector material graphite 144 is 30 cm thick in this example, the thickness of the Teflon 168, t, in front of the 2.5 MeV source is varied, and the portion 170 of the moderator is either 7LiF or D2O. As t changes, the thickness of the Al/AlF3 166 of the moderator changes, with all other dimensions remaining constant. The target is embedded in the Teflon 168, UHMW or HDPE. Sources are titanium targets 106 being bombarded by deuterium ion beams 5.0 cm in diameter. Each target is emitting 4×1010 neutron/sec. Eight modulator generators 118 emit 3.2×1011 n/s total emission. A concentration of 10B in the tumor and health tissue (e. g. skin) is known to be possible. 10B tumor concentration is assumed to be 50 ppm, while 10B in healthy tissue is 15 ppm. The relative biological effectiveness (RBE) for 10B in tumor is 2.7, and in healthy tissue is 1.3. Tumor and healthy tissue doses are calculated using the NRC and ICRP models for neutron RBE. The material 7LiF was the best performer and D2O was second best. An important main objective in these examples is to give a sufficient dose of neutrons to the cancer while minimizing the dose to the healthy tissue and not damaging it. FIG. 15 shows the performance for moderators with different values for tin cm and either 7LiF or D2O in the secondary moderator. The ordinate R is the ratio of tumor dose at the origin to healthy tissue skin dose, and the tumor dose at the center of the brain assumed to be the site of the cancer. As can be seen from FIG. 15, 7LiF outperforms D2O. The best performance is R=1.9 and a tumor dose in excess of 1.4 Sv/hr. A consequence of RBE is that a small percentage of fast neutrons is essential to obtain a high value for R; also, a reasonable number of epithermals is required to penetrate the target. Thus a combination of 7LiF and D2O may outperform either material alone. A Need for Small Animal Neutron Sources Development of boron delivery agents for BNCT is an ongoing and challenging task of high priority. A number of boron-10 containing delivery agents have been prepared for potential use in BNCT. With the development of new chemical synthetic techniques and increased knowledge of the biochemical requirements needed for an effective agent and their modes of delivery, a wide variety of new boron agents has emerged, but only two of these, oronophenylalanine (BPA) and sodium borocaptate (BSH) have been used clinically and have US FDA approval. Patient-derived xenograft (PDX) is created by transferring primary tumors from a patient into a mouse or small animal model. Tests of delivery and effectiveness of drugs to the cancer site can then be performed. In the prior art, only beamlines from nuclear reactors and linear accelerator structures can be used. A small laboratory neutron source, as in embodiments of this invention, is therefore valuable in the development and testing of new boron delivery drugs and their effectiveness in destroying the cancer site. As compared to a clinical delivery system, a smaller number of stand-alone generators such as generators 118 is needed for a delivery system for a small animal such as a mouse. The modular generators used have a slab wall angle of α=0 (see α defined in FIG. 12 A). The secondary moderator may be a separate container of heavy water (D2O). Since the small animal target is indeed small, the secondary moderator volume can be reduced, and the compact modular generators can be moved close to it permitting the modular generators to be closer to the animal target. Thus, the neutron flux at the cancer site is increased, and with proper selection of moderator material and size, will still be able to moderate the neutrons to IAEA standards. In addition, by moving closer, the number of generators can be reduced while still maintaining a high thermal neutron flux at the cancer site. In our example of the new art for a small animal source, we can use four modular generators 118 to emit enough thermal neutrons at the cancer site. We can use the modular generators of 12 A, B, C with the slab wall angle of α=0. This makes the pre-moderator 108 a rectangular cuboid (or “rectangular slab” of). FIG. 16A is a perspective view of a modular generator having such a rectangular pre-moderator 108, making it suitable for arrangements of four generators in a rectangular array, as shown in FIG. 16B. In FIG. 16B the four modular generators are arranged around a secondary moderator 112, which in one embodiment may be a container of heavy water or granulated moderator material. FIG. 16C is a cross section view of the arrangement of FIG. 16B, taken along section line 4 16C-16C of FIG. 16B. The elements previously annotated for modular generators are reused in FIGS. 16 A, B and C. FIG. 16D is an exploded view where the four generators 118 are moved back from the small heavy water moderator 112. Each generator 118 has a pre-moderator 108 with a fast neutron generator with a titanium target 106. A deuterium ion beam is generated by a plasma ion source 102 and accelerated in an acceleration chamber 100 to the titanium target 106, where the DD fusion reaction occurs releasing fast 2.5 MeV neutrons. This description is all common to the descriptions or other embodiments in the specification. The neutrons generated pass through a pre-moderator 108, where they are partially moderated to thermal neutron energies. They then pass into the moderator block 112 where they are further moderated, reducing the energy of fast neutrons to thermal neutron energies. The thermal neutrons then enter a cylindrical mouse chamber 114 where they enter the small animal 116. The pre-moderator is designed to slow the fast neutrons to thermal neutrons by scattering the fast neutrons via collisions with the hydrogen in the HDPE or UHMW plastics. The distance the 2.5 MeV neutrons have to traverse is approximately 3 to 5 cm, wherein approximately 50% of the neutrons lose enough of their energy to be classified as thermal neutrons. These neutrons, containing both thermal and fast neutron components, can then travel into the moderator box 112, where they are further moderated by collisions with deuterium atoms. Roughly speaking, the HDPE with its hydrogen-atoms moderates the neutrons to thermal energies over a short distance; the thermalized neutrons then penetrate the cylindrical chamber 114 wherein they place the small animal 116. The small animal model is used to test the delivery of boron to the cancer site. For the pre-moderator, high density polyethylene (HDPE) is optimum for producing the maximum flux of thermal neutrons. As in the case of the clinical generator, it is desired to produce a maximum thermal flux at the cancer site. A mouse is a small object, and penetration of thermal neutrons to the cancer site can easily be achieved. Moderation of the fast neutrons to thermal energies is desired with minimum production of gamma radiation, which is harmful to the healthy cells. As those skilled in the art will understand, hydrogen atoms are excellent at scattering fast neutrons, resulting in moderation of the neutrons to thermal energies in the shortest path length in the moderating material. Indeed, using 5-6 cm of high-density polyethylene (HDPE) or UHMW plastic results in moderation of about 50% of 2.5 MeV neutrons to thermal energies. Further moderation of the neutrons by longer distances in the HDPE results in more fast neutrons being converted to thermal energies. However, this results in reduction of the total flux (n/cm2) that is available since the neutrons are being emitted in a 47 solid angle. Hydrogen capture of neutrons produces high energy gamma radiation, which is destructive to both healthy and cancerous cells. Adding another moderator to further thermalize the neutrons is accomplished by the use of heavy water (D2O). The skilled person will understand that the embodiments described in this application are exemplary, and not limiting. Many variations may well fall within the scope of the invention, which is limited only by the scope of the following claims. |
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claims | 1. A method of measuring Doppler reactivity coefficient, comprising:measuring time-series data of neutron flux in which reactor power is increased by a prescribed amount by applying reactivity to a reactor core which is in sub-critical or achieves super critical, and obtaining time-series data of in-reactor average moderator temperature in which reactor power is increased by the prescribed amount by applying reactivity to the reactor core which is in sub-critical or achieves super critical, whereinin said measuring step, neutron flux during this period is measured as time-series data, andin said step of obtaining time-series data of in-reactor average moderator temperature, average moderator temperature in the reactor is obtained as time-series data;obtaining time-series data of reactivity in which the time-series data of reactivity is obtained from the measured time-series data of neutron flux, said time-series data of reactivity being calculated by applying an inverse kinetic method based on a one-point reactor kinetic equation;obtaining time-series data of reactor power by calculating the time-series data of reactor power based onsaid obtained time-series data of in-reactor average moderator temperature andthe time-series data of neutron flux,such that the obtained time-series data of reactor power matches said two time-series data which are the time-series data of in-reactor average moderator temperature and the time-series data of neutron flux,obtaining time-series data of fuel temperature in which the time-series data of fuel temperature subjected to a prescribed averaging is obtained from the time-series data of reactor power obtained above and a reactor kinetic model;obtaining time-series data of reactivity feedback contribution component in which the time-series data of reactivity feedback contribution component is calculated fromthe time-series data of reactivity obtained above in said step of obtaining time-series data of reactivity, andthe reactivity of constant reactor period; andobtaining Doppler reactivity coefficient, in which the Doppler reactivity coefficient is obtained fromthe obtained time-series data of in-reactor average moderator temperature,the obtained time-series data of fuel temperature subjected to said prescribed averaging,an isothermal temperature reactivity coefficient, andthe obtained time-series data of reactivity feedback contribution component. 2. The method of measuring Doppler reactivity coefficient according to claim 1, whereinmeasurement of the time-series data of neutron flux at said step of measuring neutron flux measures neutron flux as well as γ-ray; andsaid step of obtaining time-series data of reactivity has a removal procedure of removing influence of the γ-ray from the measured time-series data of neutron flux, and, from the time-series data with the influence of γ-ray removed, time-series data of reactivity is obtained from said inverse kinetic method with respect to the one-point reactor kinetic equation. 3. The method of measuring Doppler reactivity coefficient according to claim 2, wherein, in said removal procedure, an error function is evaluated by the least squares method in the following way:the error function is defined froma time-change numerically evaluated value calculated by a nuclear reactor kinetic equation based on a reactivity of constant reactor period and a γ-ray mixture rate as parameters related to reactor power response in a low power range in which reactivity feedback contribution is negligible, anda time-change part corresponding to the reactor power response of actually measured time-series data of neutron flux,the error function being calculated as a difference between these two in logarithmic value; anda combination of the reactivity of constant reactor period and the γ-ray mixture rate that minimizes the error function value is aimed. 4. The method of measuring Doppler reactivity coefficient according to any one of claims 1-3, whereinat said step of obtaining time-series data of in-reactor average moderator temperature, the average moderator temperature is obtained in the form of time-series data, when the reactor power is increased by a prescribed amount by applying reactivity to a reactor core which is in sub-critical or achieves super critical. 5. The method of measuring Doppler reactivity coefficient according to claim 4, whereinat said step of obtaining time-series data of reactor power,a time constant τsg,12 related to heat transfer from a primary side to a secondary side of a steam generator associated with the reactor, and an initial reactor power P0 are selected as parameters, anda combination of said time constant τsg,12 and the initial reactor power P0 is obtained that minimizes the value of the error function E(τsg,12, PO) represented by: E ( τ sg , 12 , P 0 ) = ( 1.0 - t p s t p m ) 2 + ( 1.0 - ΔT c , av s ΔT c , av m ) 2 wherein tsp represents analytical time of average moderator temperature to maximum, tmp represents measured time of average moderator temperature to maximum temperature, ΔTsc,av represents analytical value of maximum change width of average moderator temperature, and ΔTmc,av represents measured value of maximum change width of average moderator temperature. 6. The method of measuring Doppler reactivity coefficient according to claim 1, whereinat said step of obtaining time-series data of fuel temperature, volume-averaged fuel temperature is numerically evaluated from a fuel rod heat conduction equation related to average fuel rod temperature, and time-series data of reactor power is modified based on a correction coefficient obtained in consideration of distributions of neutron flux and adjoint neutron flux (neutron importance) in a moderator flow path direction in zero-power state,such that time-series data of fuel temperature subjected to prescribed averaging, based on the first-order perturbation theory, is obtained. 7. The method of measuring Doppler reactivity coefficient according to claim 1, whereinsaid prescribed averaging is importance power averaging in said step of obtaining time-series data of fuel temperature,wherein said time-series data of fuel temperature is used for obtaining the Doppler reactivity coefficient αf, such that:in said step of obtaining Doppler reactivity coefficient, the Doppler reactivity coefficient αf is obtained from the following equation:Δρfd(t)=αf(βTf,av(t)−ΔTc,av(t))+αitcΔTc,av(t)wherein in said equation Δρfd(t) represents a reactivity feedback contribution component related to Doppler reactivity feedback, αf represents a Doppler reactivity coefficient, ΔTf,av(t) represents a change amount of average fuel rod temperature, ΔTc,av(t) represents change amount of average moderator temperature, and αitc represents an isothermal temperature reactivity coefficient. 8. The method of measuring Doppler reactivity coefficient according to claim 7, whereinin said step of obtaining Doppler reactivity coefficient based on the equation of claim 7, the actual Doppler reactivity coefficient αf is estimated as the coefficient that minimizes the value of an error function Erdf which is defined as: E rdf = 1 N ∑ i = 1 N { 1.0 - α f ( c ip Δ T f , av ( t i ) - ΔT c , av ( t i ) ) Δρ fc ( t i ) } 2 wherein N represents the number of data, ti represents time corresponding to data i, Δρfc(t) represents a reactivity contribution component, and cip represents a correction coefficient defined as: c ip = Δ T f , av ip Δ T f , av wherein ΔTipf,av(t) represents a change amount of importance power-averaged value of fuel temperature, andwherein ΔTf,av(t) represents a change amount of average fuel rod temperature, and ΔTc,av(t) represents a change amount of average moderator temperature as recited in claim 7. |
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description | This application claims the benefit of U.S. Provisional Application Ser. No. 60/486,591, filed Jul. 11, 2003, entitled “Curved Honeycomb article, EUV Apparatus Including a Curved Honeycomb Article, and Method of Making a Curved Honeycomb Article,” which is hereby incorporated herein by reference in its entirety. 1. Field of the Invention The present invention relates generally to the manufacture and use of novel honeycomb articles. 2. Technical Background The desire for faster, more powerful integrated circuits has led to rapid development of advanced ultraviolet photolithographic methods. The performance of an integrated circuit increases as the feature size decreases; hence, a decrease in feature size allows more circuitry to be put on a chip of a given size, and reduces the power needed for operation. The feature sizes obtainable in a photolithographic process depend on the wavelength of radiation used in the exposure step; shorter wavelengths enable smaller feature sizes. As such, there has been a trend toward shorter wavelengths in photolithographic processes. Currently, photolithography systems based on wavelengths as low as 193 nm and 157 nm are being developed for commercial use. In order to further decrease the minimum feature size obtainable in photolithographic processes, it has been suggested to use extreme ultraviolet (EUV) radiation, for example, having a wavelength of about 13 nm. The use of EW radiation in photolithography, while greatly reducing feature size, forces a radical departure in design from currently used photolithographic apparati. First, no practical materials are sufficiently transparent to EUV radiation to be used as windows, lenses or photomasks. As such, any manipulation of the radiation must be performed using reflection. Mirrors constructed from alternating layers of molybdenum and silicon have been used as reflecting focusing mirrors, collimators, and photomasks in EUV apparati. Further, EUV sources tend to be rather dirty. In one conventional EUV source, a high energy laser is used to heat an object, which functions as a source of secondary emission of mainly shortwave radiation. This process releases undesirable particles and atoms, forming debris in the apparatus. The debris, known generally herein as “soot,” can deposit on the mirrors and on the photomask, wreaking havoc with the necessarily precise optical system of the EUV apparatus. Since there currently exists no suitable window material for EUV radiation, the design of a device to catch the particulate matter while transmitting the EUV radiation is not trivial. One suggestion for an EUV soot filter is described in international patent application publication WO 99/42904, which is incorporated herein by reference. The apparatus described therein is placed between the EUV source and optical system, and has a plurality of foils or plates arranged in a direction radial from the EUV source. The position of the plates allows any EUV radiation propagating directionally from the EUV source to the optical system to be transmitted, but will catch the randomly diffusing soot. The filter is assembled from a plurality of copper plates; such a filter would be difficult to assemble, especially as the desired size of the filter increases. It would be desirable to have an EUV soot filter that is simple to fabricate and easily adaptable to a number of EUV source geometries. One embodiment of the present invention relates to a method for making a curved, honeycomb article, the method including the steps of providing a honeycomb body having a first face, a second face, and a honeycomb structure having a plurality of substantially parallel channels formed from the first face to the second face; filling the channels of the honeycomb body with a filler material, thereby forming a filled honeycomb composite; molding the filled honeycomb composite to form a curved filled honeycomb composite; and finishing the curved filled honeycomb composite into the curved honeycomb article, wherein during the step of molding, the filler material has a plastic deformation behavior compatible with that of the honeycomb body. Another embodiment of the present invention relates to an EUV lithography apparatus including an EUV source; an optical system coupled to the EUV source; and a curved honeycomb article made using the process described above placed operatively between the EUV source and the optical system, such that a substantial fraction of the EUV radiation generated by the EUV source propagates through the channels of the curved honeycomb article. Another embodiment of the present invention relates to a curved honeycomb article having a first face, a second face, and a plurality of channels formed from the first face to the second face, each channel having a channel axis, the curved honeycomb article having a width of at least about 15 cm in at least one direction in the plane normal to a channel axis at its geometrical center, each channel having a virtual channel extension associated therewith, the virtual channel extensions defining a convergence area substantially smaller than the occluded area of the curved honeycomb article. Another embodiment of the invention relates to an EUV lithography apparatus including an EUV source; an optical system coupled to the EUV source; and the curved honeycomb article described above placed operatively between the EUV source and the optical system, such that a substantial fraction of the EUV radiation generated by the EUV source propagates through the channels of the curved honeycomb article. The curved honeycomb articles, EUV apparati, and methods of making curved honeycomb articles of the present invention result in a number of advantages over prior art devices and methods. The curved honeycomb articles of the present invention may be made to be quite large, and may be curved in many desirable geometries (e.g., spherical, elliptical, cylindrical, or more complex). The EUV apparati using the curved honeycomb articles of the present invention may utilize relatively large EUV sources. Further, use of the curved honeycomb articles of the present invention in an EUV apparatus can provide a high degree of debris mitigation in EUV photolithographic processes. The methods of the present invention can be used to easily and relatively rapidly form curved honeycomb elements with a minimum of waste. The methods can be adapted by the skilled artisan to make a wide variety of curved honeycomb geometries. Additional features and advantages of the invention will be set forth in the detailed description which follows, and in part will be readily apparent to those skilled in the art from the description or recognized by practicing the invention as described in the written description and claims hereof, as well as in the appended drawings. It is to be understood that both the foregoing general description and the following detailed description are merely exemplary of the invention, and are intended to provide an overview or framework for understanding the nature and character of the invention as it is claimed. The accompanying drawings are included to provide a further understanding of the invention, and are incorporated in and constitute a part of this specification. The drawings are not necessarily to scale, and sizes of various elements may be distorted for clarity. The drawings illustrate one or more embodiment(s) of the invention, and together with the description serve to explain the principles and operation of the invention. One embodiment of the present invention relates to a method of making a curved honeycomb article that is especially useful as a soot filter in EUV methods and apparati. The method according to one embodiment of the invention is shown in schematic cut-away view in FIG. 1. A honeycomb body 20 having a first face 22, a second face 24, and a honeycomb structure 25 formed from a plurality of substantially parallel channels 26 formed from the first face to the second face is provided. Channels 26 of honeycomb body 20 is then filled with a filler material 28 to form a filled honeycomb composite 30. Filled honeycomb composite 30 is molded (e.g. forged) to form a curved filled honeycomb composite 32, which comprises a curved honeycomb body 34 filled with filler material 28. Curved filled honeycomb composite 32 is finished to form the curved honeycomb article 36. As will be described in more detail below, the curved honeycomb bodies and articles made by the method of the present invention may have a variety of shapes, such as a spherical shell segment shape, or a cylindrical shell segment shape. The honeycomb body 20 may be made from a variety of materials and by a variety of methods. Ceramic honeycomb structures are widely used for applications such as catalyst substrates and honeycomb heaters. The fabrication of honeycomb structures by extrusion of plasticized batches of inorganic powders is well known, and is described in U.S. Pat. Nos. 3,790,654 and 3,905,743, each of which is incorporated by reference in its entirety. These extrusion methods typically involve compounding a batch material having a plastic consistency from inorganic powders, plasticizer, vehicle and binder components. The plasticized batch is thereafter forced through an extrusion die to form a honeycomb shape, which may be solidified by drying and firing. Dies for honeycomb extrusion typically have a die body incorporating a number of feedholes on an inlet face which extend through the body to convey the plasticized batch material to a discharge section on an opposing die outlet face. The discharge section incorporates a criss-crossing array of discharge slots, cut into the outlet face to connect with the feedholes within the die body. These slots reform the batch material supplied by the feedholes into the interconnecting cell wall structure of the desired honeycomb body. The composition of the inorganic powders used to formulate the plasticized batch for extrusion of the honeycomb body is not critical. The powders may be, for example, metallic, semi-metallic, ceramic, glass, polymeric, or mixtures thereof. The powders may be employed either in raw (mineral) or refined form. Specific examples of powder types include powdered glasses or powdered crystalline or semi-crystalline ceramic materials, or the precursors thereof, particularly including amorphous silicate, borate or aluminate glasses and/or crystalline oxides (e.g. silicates, aluminates, borates), carbides, borides, and aluminides. High purity glass soot formed, for example, by a flame hydrolysis process may also be used, as described in U.S. Pat. No. 6,468,374, which is incorporated herein by reference in its entirety. Particular powder materials may include mineral powders such as cordierite, spinel, various clays, talc, refined powders of alumina, silica, and the oxides of calcium, magnesium, boron, titanium, germanium, and the alkali and transition metals, and various mixtures or chemical combinations thereof. Powdered polymers (e.g., PTFE), and powdered metals (e.g., tungsten) may also be used. The plasticizing vehicle/binder system used to compound the plasticized powder batch will depend in part on the composition and morphology of the solid powder components of the batch. Aqueous binder systems comprising a water vehicle and a plasticizing additive such as a cellulosic binder, e.g., a methyl hydroxypropyl cellulose, can provide highly plastic batches, particularly if the powders include substantial proportions of kaolinitic clays. Batches of this type, disclosed for example in U.S. Pat. No. 3,885,997, which is incorporated herein by reference, are presently in large-scale commercial use for the production of cordierite honeycombs. Other components which may be present in these types of binder systems, both for metallic and for ceramic powders, include dispersants, surfactants, lubricants, polymers and/or additional water-miscible and/or water-immiscible organic vehicles. Specific types of compounds which may been included in these batches include the alkali stearates, oleic acid and its derivatives, and co-binders such as the polyvinyl alcohols and silicones. Non-aqueous binder systems that secure a degree of plasticity to an extruded honeycomb shape can also be used. These may include polymer, solvent, or wax-based binder systems, the latter including mixtures of waxes, wax-polymer blends, and solutions of waxes in various organic solvents. U.S. Pat. No. 5,602,197, for example, discloses extrudable ceramic and/or metal powder batches with particularly good post-forming plasticity, based on a gelling binder system comprising an elastomeric polymer component (e.g. a KRATON brand elastomeric polymer) dissolved in a low-melting wax vehicle. Batches formed from the selected inorganic powders and aqueous or non-aqueous binder systems may be compounded and conditioned for extrusion using known mixing and plasticizing methods and equipment, and may be formed into honeycomb bodies by conventional forming procedures such as molding or, more preferably, extrusion. Continuous extrusion procedures and apparatus, such as disclosed, for example, in U.S. Pat. Nos. 3,790,654 and 4,551,295, are particularly well suited to the production of precisely engineered honeycomb bodies at relatively low cost. The extrusion process can conveniently be used to make long honeycomb structures which may not be desirable (or necessary) for the molding process used in the present invention. As shown in FIG. 2, it may be desirable to use an extrusion process to make an elongated honeycomb structure 40, then slice the elongated honeycomb structure into thinner honeycomb bodies 42 more suitable for the molding processes described herein. The extruded honeycomb body may be used in its as-extruded plastic state (i.e. “wet-green”), or may be hardened to a “dry-green” state before the step of filling. Hardening may be accomplished, for example, by drying, gelling, or freezing. The hardening is desirably conducted in a reversible manner such that the honeycomb body can be returned to its plastic state before the molding step. The skilled artisan will recognize that other methods may be used to make the honeycomb bodies used in the present invention. For example, stack-and-draw methods and coextrusion methods as well as conventional machining and molding methods may be used to make the honeycomb bodies used in the present invention. As shown in FIG. 3, the honeycomb body 20 has a honeycomb structure 25. The honeycomb structure may have any desired geometry (e.g. square, triangular, hexagonal); however, the hexagonal cell geometry shown in FIG. 3 is especially desirable, as it is highly compliant and isotropic in the plane. Filling of the parallel channels of the honeycomb body with an appropriate filler material should be accomplished in a manner which will protect the cellular channel structure of the honeycomb body from inadvertent damage to or alteration of the initial channel size, shape and volume during the filling process. In the case of honeycomb bodies which are to be filled in their as-extruded plastic state, the filler material will typically be introduced into the honeycomb channel structure in liquid or dispersed form. The liquid filler material is then partly or completely solidified to convert it to a plastic state suitable for molding of the filled honeycomb composite. Solidification may result, e.g., from crystallization or gelling. In the case of gelling, the gelling process may be initiated by a change in temperature, or by the addition of a gelling agent, combined with the filler material before filling, or introduced therein from the channel walls of the honeycomb body. When the as-extruded honeycomb body is hardened before the filling step (e.g., by freezing or drying), conversion of the filler material to liquid form is not necessary. In such a case, filler materials having a semi-solid or paste consistency at filling temperatures can be introduced into the channels. Excess filler material should be trimmed away from the filled honeycomb composite. The filler material is selected such that it has a plastic deformation behavior compatible with that of the honeycomb body during the molding step. By compatible deformation characteristics is meant that the honeycomb body and the filler material deform as if they were one at the temperature of the molding step. If the filler material is marginally soft, reshaping of the honeycomb structure during molding may occur in a fracture manner rather than in the desired plastic manner. When reshaping is accompanied by fracturing sufficient to introduce voids or openings in the honeycomb channel walls, the honeycomb body and filler materials do not have compatible deformation characteristics. One suitable test for evaluating the physical suitability of a candidate filler material is the blunt indenter test. In this test, the end of a 2 mm diameter rod is pushed perpendicularly into a candidate filler material at a pre-determined molding temperature (i.e., the temperature to be used in the subsequent molding step) and the filler deformation mechanism is noted. An elastic response (e.g. a JELLO gelatin-like recovery) is undesirable. A brittle response (e.g. radiating fracture lines and haze) is also undesirable. A plastic response (e.g. a plastic heaving up of the material about the indenter) is desirable. The chemical composition of the filler material will be selected with due regard for the composition of the honeycomb body, most importantly the composition of the honeycomb binder system employed. The filler material should have little or no solubility in or miscibility with the binder components of the honeycomb body, nor should it exhibit substantial osmotic affinity or solvating activity for any components of the binder system. In general, these conditions will be favored if hydrophobic filler materials (e.g. wax-based) are used with aqueous binder systems (e.g. methylcellulose-based binders); or if hydrophilic filler materials (e.g. starch or polyethylene glycol based materials) are used with water-immiscible organic binders (e.g. elastomer-based binders). If the filler material and binder share the same solvent, they should be chosen to be in osmotic balance. Optimum filler material compositions for any particular binder system can of course be readily determined through routine experiment, using a suitable binder/filler material contact interval to identify interactions deleterious to the requisite properties of either. Examples of families of filler materials useful in combination with aqueous binder systems include heat softenable vegetable or animal fats, natural or synthesized fatty acids, polyalcohols and/or esters, paraffins (often blended with other components for improved flexibility and plasticity), other hydrocarbon waxes, both natural and synthetic, and synthesized thermoplastic polymeric materials. Particular examples of filler materials of these types range from butter to microcrystalline wax to crystalline waxes in combination with modifiers such as propylene glycol monostearate and mineral oil. Butyl rubber and/or poly(isobutylene) may be used as filler additives to modify the properties of the filler materials, particularly the microcrystalline waxes. Specific filler materials of wax type which are expected to offer good performance when used with aqueous honeycomb binder systems containing methylcellulose and/or hydroxypropyl methylcellulose as the principal plasticizing constituent are reported below in Table 1. Included in the Table for selected ones of the filler materials are the melting points of the materials (as determined by ASTM D-127), needle penetration at 25° C. (ASTM D-1321) viscosity at 99° C. (in Saybolt Universal Seconds (SUS) per ASTM D-445) and density at ambient (25° C.) and near-boiling (99° C.) temperatures. As typical of many commercial wax formulations, certain of the physical and thermal properties of the waxes are reported as ranges. These values are not controlled to discrete values by wax manufacturers, since the properties of the products may vary within relatively wide limits without adversely impacting utility for most commercial applications. In general the materials reported in Table 1, which may all be characterized as microcrystalline waxes, are ranked in general order from relatively hard and flexible to relatively soft and sticky in character. TABLE 1MeltingNeedleViscosityDensity at 25°FillerPointpenetrationat 99°C./99° C.Material(° C.)at 25° C.C. (SUS)(g/mL)Bareco Victory74/7920/3570/900.93/0.79WaxBareco63.9/74.421/3979/940.92/0.79Ultraflex WaxWitco W-44577/8225/3575/90waxWitco W-83574/7960/8075/90waxBlended wax7465BW-568 As previously indicated, the relative plasticities of the various filler material and honeycomb powder/binder mixtures used in the practice of the invention are typically temperature dependent. That is, each of the components of the filled honeycomb composite will have plastic characteristics which depend at least to some degree on the temperature of its constituent materials. This does not present a problem provided there is a least some temperature range over which the deformation characteristics of the filler material are sufficiently compatible with those of the honeycomb body so that the filler material will plastically and hydrostatically support the cellular honeycomb structure from buckling and/or fracturing during the process of molding. Further information regarding reshaping of filled honeycombs can be found in U.S. Pat. No. 6,299,958, which is incorporated herein by reference in its entirety. The filled honeycomb composite formed by filling the honeycomb body with the filling material desirably has a thickness (i.e. in the direction parallel to the channels) that is small enough to allow for rapid heat transfer through the entire thickness during the molding step. Using a filled honeycomb composite with a relatively small thickness in the present invention allows the skilled artisan to fabricate large curved honeycomb bodies (e.g. greater than 16 cm in diameter) without complications due to poor heat transfer to the interior of the filled composite. The filled honeycomb composite is desirably less than about 8 cm in thickness. More desirably, the filled composite is less than about 4 cm in thickness. In certain desirable embodiments of the invention, the filled honeycomb composite has a thickness between about 12 mm and about 3 mm. If the honeycomb body was hardened to a dry-green state before filling, it may be necessary to first reconstitute the dry-green body to a plastic wet-green body before filling. It is possible to return a dry-green honeycomb body earlier hardened by drying (but not sintering) to a plastic condition by introducing a liquid back into the structure. Where the dry-green body is originally plasticized with a water solution of a reversibly thermogellable binder such as a cellulose ether and thereafter dried, reconstitution may be accomplished by saturating the pores of the body with water at a temperature above the gel point of the binder. Any liquid water remaining in the channels should be removed (e.g., by draining) prior to reducing the temperature of the body to the gel point. For example, in the case of a honeycomb body formed from mineral powders bound together with a methyl cellulose binder additive, exposure to near-boiling water can develop the saturation necessary to provide plasticity without undesirable swelling of the binder. In one example of a suitable method form making the filled honeycomb composite, a hot wet-green honeycomb body is filled with hot wax, then cooled to solidify the wax in the channels and to redissolve the plasticizer into the channels. The filled honeycomb composite is then molded into a curved filled honeycomb composite using a method selected for its compatibility with the material of the honeycomb body and the filler material. The molding method may be a confined method (as in an isostatic pressing method), or an unconfined method (as in a sagging method). In a confined method, the volume of the material being molded is inhibited from increasing (by cracking), for example by application of a pressure to the entire filled honeycomb while molding. In an unconfined method, the volume of the material being molded may increase in the molding process; this can occur globally by an increase in porosity, or locally and coarsely when the material fractures on bending or stretching. When easy fracture on bending occurs, the honeycomb body is referred to as being “short”. A “short” honeycomb body coarsely and locally fractures rather than finely and uniformly fractures (unsaturates to become porous) or plastically (constant volume) deforms under unconfined deformation. An extrudable “short” material may be plastic under confined deformation, but may be brittle under unconfined deformation. Particularly for relatively “short” materials such as methyl cellulose-plasticized clay-based batches of the kind used for commercial cordierite honeycomb production, the preferred mode of molding is compressive (e.g. press molding) rather than tensile (e.g. sagging). As noted above, compressive or confined deformation methods tend to keep the volume of the material constant during deformation so as to not introduce a compressible phase (i.e. an open crack, pore or void). In cases where the material of the honeycomb body is quite “short” and/or the filler material is too soft, the filler material penetrates and/or opens fissures in the honeycomb channel walls. This permits the honeycomb body volume to increase, i.e., an decrease in open frontal area of the honeycomb body due to the increase in volume of the honeycomb wall structure and the decrease in volume of the channels. Particularly for many ceramic-powder-filled pastes, the honeycomb body is only plastic under constant volume (confined) deformation, and even in that case an overly-soft fill promotes fissuring and checking. This is most clearly observed with some of the softer waxes. With extreme softness the filler material tends to be expelled from the honeycomb channels during molding, causing complete or partial collapse of some or all of the channels. Conversely, if the filler material is too hard and brittle, then gross fractures/slips can form, as indicated by cell mis-registration throughout the curved honeycomb body. This can be observed with some of the harder fillers such as propylene glycol monostearate. Again, the optimum molding temperature and conditions for any filler material/honeycomb body system can readily be determined by routine experiment. An example of a desirable molding process for use with a filled composite based on a wax-filled cordierite honeycomb body (e.g., as described in U.S. Pat. Nos. 3,885,997 and 6,299,958) is shown in schematic view in FIG. 4. The filled honeycomb composite 50 is first vacuum bagged. The step of vacuum bagging may be desirable in order to help prevent cracks from opening during the molding step. Vacuum bagging helps keep the filled honeycomb composite in a compressed state during the molding step, thereby helping to ensure a confined molding environment. The filled honeycomb composite 50 (inside vacuum bag 52) is then clamp forged between convex tool 54 and concave tool 56 to form a bagged, curved filled honeycomb composite having a concave fact and a convex face. If the desired curvature has a small radius of curvature, it may be necessary to perform the clamp forging in a plurality of steps, using multiple tools of decreasing radius of curvature, thereby forming a progressively greater curvature in each step. When forging in a plurality of steps, it may be desirable to anneal the curved filled honeycomb composite (as described below) between forging steps. For certain materials, such as the cordierite material mentioned above, the forging step is desirably performed at a somewhat elevated temperature (e.g.30–40° C.). The bagged, curved filled honeycomb composite may then be annealed, desirably in an isostatic press chamber. In an alternative embodiment of the invention, the clamp forging is performed in an isostatic press chamber, with or without vacuum bagging. As the skilled artisan will recognize, the general molding technique described above may be modified in a number of ways. For example, a rubber bladder may be used as one of the tools (convex or concave); the bladder may be expanded using liquid or gas pressure to force the honeycomb body against a second tool. The honeycomb body may be kept in a compressed state during molding using methods other than vacuum bagging. For example, the honeycomb body may be encased in a wax casement. After the filled honeycomb composite is molded into the curved filled honeycomb composite, it is finished into a desired curved honeycomb article. As used herein, finishing includes whatever steps are necessary to covert the curved filled honeycomb composite into the curved honeycomb article. For example, if the curved filled honeycomb composite was vacuum bagged, the finishing step may include removing the vacuum bag. During the finishing step, the filler material is desirably removed from the filled honeycomb composite. Filler material removal may be carried out through the use of solvents or other chemical means if desired. However, for most heat-softenable fillers, removal is best affected by the application of a mild heat treatment to melt, liquefy and gravity drain and/or blow and/or blot the filler from the curved filled honeycomb composite to form a curved honeycomb body. It may be desirable to remove any remaining filler material using vacuum pyrolysis. The curved honeycomb body formed by removing the filler material from the curved filled composite may be treated to harden it to a final, usable state. For plasticized ceramic or metal powder formulations, the treatment typically requires drying (e.g. to remove the water plasticizer) and relatively high temperature firing to debind and to sinter or reaction-sinter the component powders into the desired consolidated material. Residual filler material present in the channel structure in these cases will ordinarily be completely removed by volatilization and/or oxidation in the course of the process. Obviously, more attention will be paid to the presence and/or behavior of residual filler material in the case of very high cell density parts, since capillary effects complicate filler removal as the channels become smaller in diameter. The finishing step may also include trimming the curved honeycomb body (e.g. by grinding, cutting, or machining) into the desired shape of the curved honeycomb article. In converting a substantially flat honeycomb body to a highly curved (e.g. spherical shell segment shaped) honeycomb body, there are likely to be strains and distortions that will have to be accommodated to avoid fracture. These may become especially important for large bodies and for small radii of curvature, and are expected to be found predominantly around the perimeter of the body. This may be thought of as analogous to the problem of projecting a spherical surface (e.g. the Earth) onto a flat surface (e.g. a map) without distortion. As shown in FIG. 5, it may be desirable for the substantially flat honeycomb body 70 to have a series of strain relief features 72 (e.g. notches around its perimeter. In the molding process, the notches will close, allowing a continuous curved honeycomb body 74 to be formed without undue strain or distortion due to compression of the larger diameter flat honeycomb body along its perimeter to a smaller diameter curved honeycomb body. Another embodiment of the present invention relates to a curved honeycomb article useful as a soot filter in an EUV lithography device. As shown in cross-sectional view in FIG. 6, curved honeycomb article 100 has a first face 102, a second face 104, and a honeycomb structure 105 formed from a plurality of channels 106 formed from the first face to the second face. Each channel 106 has a channel axis 108. The channel axes should not be mutually parallel. Desirably, most of the channel axes are arranged in a substantially radial fashion in at least one dimension. In certain desirable embodiments of the invention, the channel axes are arranged in a substantially radial fashion in two dimensions. Curved honeycomb article 100 desirably has a width of at least about 15 cm in at least one direction in the plane normal to a channel axis at its geometrical center. More desirably, the curved honeycomb article has a width of at least about 15 cm in two mutually perpendicular directions in the plane normal to a channel axis at its geometrical center. In certain desirable embodiments of the present invention, the curved honeycomb article has a width of at least about 20 cm in at least one direction in the plane normal to a channel axis at its geometrical center. The methods described hereinabove may be used for making the curved honeycomb article according to this embodiment of the invention. The honeycomb structure of the curved honeycomb article has an occluded area. For a general curved honeycomb article, the occluded area may be defined as the projection of the honeycomb structure onto a plane perpendicular to a channel axis at its geometrical center. For a curved honeycomb article installed as a soot filter in an EUV apparatus, the occluded area may be defined as the projection of the honeycomb structure onto a plane perpendicular to the optical axis of the EUV apparatus in the neighborhood of the soot filter. Each channel of the honeycomb structure has a virtual channel extension along the concave side of the curved honeycomb article. The virtual channel extension can be thought of as continuing the shape of the channel. As shown in FIG. 7, a channel 120 shaped as a rectangular solid will have a rectangular solid-shaped virtual channel extension 122. A channel 124 having a taper (e.g., the frustum of a square pyramid) will have a virtual channel extension 126 that continues to taper, converges to a point, then becomes larger. The virtual channel extensions of a substantial fraction of the channels of the honeycomb structure define a convergence area. The convergence area is defined as the area in a plane on the concave side of the curved honeycomb article where a substantial fraction (e.g. greater than about 50%) of the virtual channel extensions converge in a relatively small (e.g., the smallest) area. In FIG. 6, curved honeycomb article 100 has a convergence area 112. Desirably, the convergence area is substantially smaller than the occluded area of the honeycomb structure of the curved honeycomb article. For example, the convergence area may have at least one dimension that is less than 50% of the corresponding dimension of the occluded area of the honeycomb structure. More desirably, the convergence area has at least one dimension that is less than 20% of the corresponding dimension of the occluded area of the honeycomb structure. In certain embodiments of the invention, the convergence area has at least one dimension that is less than 5%, or even less than 1% of the occluded area of the honeycomb structure. The convergence area may have at least two mutually perpendicular dimensions that are each less than 50% of the corresponding dimensions of the occluded area of the honeycomb structure. More desirably, the convergence area has at least two mutually perpendicular dimensions that are each less than 20% of the corresponding dimensions of the occluded area of the honeycomb structure. In certain embodiments of the invention, the convergence area has at least two mutually perpendicular dimensions that are each less than 5%, or even less than 1% of the occluded area of the honeycomb structure. The convergence area desirably has at least one dimension that is less than about 2 cm in length. More desirably, the convergence area desirably has at least one dimension that is less than about 5 mm in length. In certain embodiments of the invention, the convergence area has at least one dimension that is less than about 1 mm in length. In order to be useful in EUV apparati of reasonable size, it is desirable for the convergence area to be less than about 60 cm from a surface on the concave side of the curved honeycomb article. More desirably, the convergence area is within about 30 cm from a surface on the concave side of the curved honeycomb article. In one envisioned use of the curved honeycomb article, a source of EUV radiation is positioned at or near the convergence area in an atmosphere of an inert buffer gas such as krypton. Any radiation that is given off by the source along one of the virtual channel extensions will propagate unimpeded through the honeycomb structure of the curved honeycomb article. Any soot given off by the source will randomy diffuse in the buffer gas, and will be trapped by the walls of the honeycomb structure. In order to increase the efficiency of the trapping of soot, the honeycomb structure desirably has a channel density of at least about 1.5 channels per square centimeter. More desirably, the honeycomb structure has a channel density of at least about 6 channels per square centimeter. In certain especially desirable embodiments of the invention, the honeycomb structure has a channel density of at least about 13 channels per square centimeter. In order to maintain adequate transparency, the honeycomb structure desirably has a geometrical transparency (total area of channels/area of honeycomb structure) of at least about 50% as viewed from the convergence area. More desirably, the honeycomb structure desirably has a geometrical transparency of at least about 70% as viewed from the convergence area. In certain desirable embodiments of the invention, the honeycomb structure has a geometrical transparency of at least about 90%. The curved honeycomb articles of the present invention may be constructed to work with a variety of EUV sources. For example, the convergence area may be very small (e.g. 5 mm in dimension, or even 2 mm in dimension), and, for example, roughly square or circular in shape. Alternatively in order to match an elongated shaped source, the convergence area may have an elongated shape, with one dimension substantially larger than the other. For example, to match an elongated shaped source, the curved honeycomb article may be formed as a cylindrical shell segment, with a convergence area as long as the device along the cylindrical axis, and a few millimeters in width. A single curved honeycomb article may have a plurality of honeycomb structures, each having its own convergence area. The skilled artisan will form the curved honeycomb articles having convergence areas to match a desired EUV source. The curved honeycomb article may have a variety of overall curvatures. For example, the curved honeycomb article may be formed as an essentially spherical shell segment, an essentially oblate shell segment, or as an essentially cylindrical shell segment. In cases where the curved honeycomb article is formed as an essentially spherical shell segment or as an essentially cylindrical shell segment, the curved honeycomb article desirably subtends an angle of at least about 50 degrees in at least one dimension. More desirably, the curved honeycomb article desirably subtends an angle of at least about 70 degrees in at least one dimension. In certain desirable embodiments of the invention, the curved honeycomb article subtends an angle of at least about 85 degrees in at least one dimension. The curved honeycomb article desirably has a concave radius of curvature of less than about 38 cm. More desirably, the curved honeycomb article has a concave radius of curvature of less than about 25 cm. Alternatively, the curved honeycomb article may have a substantially aspherical curvature, such as a parabolic or elliptical curvature. The curved honeycomb article may have a plurality of curved areas, each with an individual honeycomb structure and associated convergence area. The curved honeycomb article desirably has a thickness of at least about 3 times the minimum cross-sectional dimension of the channels of the honeycomb structure. For example, depending on the channel density of the honeycomb structure, the curved honeycomb article may desirably have a thickness of at least about 3 mm. In alternative embodiments of the invention, the curved honeycomb article may desirably have a thickness of at least about 8 mm. In certain embodiments of the invention, the curved honeycomb article has a thickness of at least about 16 mm. Another embodiment of the invention relates to an EUV apparatus including a curved honeycomb body. As shown in schematic cross-sectional view in FIG. 8, an EUV apparatus 200 according to this embodiment of the invention includes an EUV source 202, an optical system 204 coupled to the EUV source, and a curved honeycomb article 206 substantially as described hereinabove. Curved honeycomb article 206 is operatively positioned between the source and the optical system, such that a substantial fraction of the EUV radiation generated by the EUV source propagates through the channels of the curved honeycomb article. The curved honeycomb element may have the form of any of the embodiments and variations described hereinabove. It will be apparent to those skilled in the art that various modifications and variations can be made to the present invention without departing from the spirit and scope of the invention. Thus, it is intended that the present invention cover the modifications and variations of this invention provided they come within the scope of the appended claims and their equivalents. |
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abstract | A method of packaging a nuclear reactor vessel for decommissioning and removal, wherein closure plates are installed onto the vessel, concrete is injected into the vessel, shielding material is installed around the exterior of the vessel and the main nozzles of the vessel, the installed shielding materials are welded to themselves, the vessel is placed on shipping cradles and attached to longitudinal restraint mechanisms for transport. |
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abstract | Damper systems selectively reduce coolant fluid flow in nuclear reactor passive cooling systems, including related RVACS. Systems include a damper that blocks the flow in a coolant conduit and is moveable to open, closed, and intermediate positions. The damper blocks the coolant flow when closed to prevent heat loss, vibration, and development of large temperature gradients, and the damper passively opens, to allow full coolant flow, at failure and in transient scenarios. The damper may be moveable by an attachment extending into the coolant channel that holds the damper in a closed position. When a transient occurs, the resulting loss of power and/or overheat causes the attachment to stop holding the damper, which may be driven by gravity, pressure, a spring, or other passive structure into the open position for full coolant flow. A power source and temperature-dependent switch may detect and stop holding the damper closed in such scenarios. |
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claims | 1. A method of chemical decontamination for a carbon steel member of a nuclear power plant, comprising steps of:bringing a reduction decontaminating solution including a malonic acid and an oxalic acid within a range from 50 to 400 ppm into contact with a surface of a carbon steel member of a nuclear power plant; andexecuting reduction decontamination for the surface of the carbon steel member by the reduction decontaminating solution. 2. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 1, comprising step of:removing cations eluted from the carbon steel member into the reduction decontaminating solution by the reduction decontamination, from the reduction decontaminating solution. 3. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 1, comprising step of:injecting oxygen gas into the reduction decontaminating solution including the malonic acid and the oxalic acid within the range from 50 to 400 ppm,wherein the reduction decontamination for the surface of the carbon steel member is performed by using the reduction decontaminating solution including the malonic acid and the oxalic acid within the range from 50 to 400 ppm with the injected oxygen gas. 4. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 3, wherein the oxygen gas is micro bubbles generated by a micro-bubble generation apparatus. 5. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 1, wherein a malonic acid concentration of the reduction decontaminating solution is within a range from 2100 to 19000 ppm. 6. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 5, wherein the malonic acid concentration is within a range from 2100 to 7800 ppm. 7. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 1, comprising steps of:putting the carbon steel member detached from the nuclear power plant in a decontamination vessel; andsupplying the reduction decontaminating solution into the decontamination vessel,wherein the reduction decontamination for the surface of the carbon steel member is performed by bringing the reduction decontaminating solution into contact with the carbon steel member in the decontamination vessel. 8. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 7, wherein a concentration of the malonic acid of the reduction decontaminating solution is within a range from 12300 to 19000 ppm. 9. A method of chemical decontamination for a carbon steel member of a nuclear power plant, comprising steps of:connecting a second pipe to a first pipe, which is made of carbon steel, of the nuclear power plant; andsupplying a reduction decontaminating solution including a malonic acid and an oxalic acid within a range from 50 to 400 ppm to the first pipe through the second pipe,wherein reduction decontamination for an inner surface of the first pipe is performed by bringing the reduction decontaminating solution into contact with the inner surface. 10. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 9, comprising step of:removing cations eluted from the first pipe into the reduction decontaminating solution by the reduction decontamination, from the reduction decontaminating solution. 11. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 9, comprising steps of:forming a closed loop including a first pipe, a second pipe, and a third pipe by connecting one end portion the second pipe to the first pipe and by connecting another end portion of the second pipe to the third pipe made of stainless steel and connected to the first pipe;supplying an oxidation decontaminating solution injected from an oxidation decontaminating solution injection apparatus connected to the second pipe into the third pipe through the second pipe;performing oxidation decontamination for an inner surface of the third pipe by the oxidation decontaminating solution; andperforming the reduction decontamination for the inner surface of the third pipe by the reduction decontaminating solution supplied to the third pipe through the second pipe as well as performing reduction decontamination for an inner surface of the first pipe. 12. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 11, wherein the reduction decontaminating solution is generated by the malonic acid injected from a malonic acid injection apparatus connected to the second pipe into the second pipe and the oxalic acid injected from an oxalic acid injection apparatus connected to the second pipe into the second pipe. 13. A method of chemical decontamination for a carbon steel member of a nuclear power plant, comprising steps of:injecting oxygen gas into a reduction decontaminating solution including a malonic acid and an oxalic acid;bringing the reduction decontaminating solution including the malonic acid and the oxalic acid with the injected oxygen gas into contact with a surface of the carbon steel member of the nuclear power plant; andperforming reduction decontamination for the surface of the carbon steel member by the reduction decontaminating solution brought into contact with the surface of the carbon steel member. 14. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to claim 13, comprising steps of:putting the carbon steel member detached from the nuclear power plant in a decontamination vessel;supplying the reduction decontaminating solution including the malonic acid and the oxalic acid into the decontamination vessel; andinjecting the oxygen gas into the reduction decontaminating solution in the decontamination vessel,wherein the reduction decontamination for the surface of the carbon steel member is performed in the decontamination vessel by bringing the reduction decontaminating solution including the malonic acid and the oxalic acid with the injected oxygen gas into contact with the carbon steel member. |
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054266750 | description | DETAILED DESCRIPTION OF THE INVENTION Referring to the drawing, FIG. 1 in particular, a nuclear fission reactor plant 10 comprises a reactor pressure vessel 12 enclosing a core 14 of fissionable nuclear fuel which is surrounded by a fuel core shroud 16. Coolant water 18 covers the heat producing fuel core 14 and fills a substantial portion of the reactor pressure vessel 12. The coolant water 18 is circulated through the reactor plant to carry away heat produced by a fission reaction within the fuel core which forms steam from a portion of the coolant water to perform work such as drive a turbine for generating electrical power. The circulating coolant water 18 follows a repeating circuit of downward in an annular path in a downcomer between the pressure vessel 12 inside wall and the outside of the shroud 16 surrounding the fuel core 14, then around the bottom of shroud 16 through the shroud support 24 and up through the heat producing fuel core 14 where a portion of the coolant water is converted to steam that continues upward with liquid coolant water whereupon the steam is dispensed from the reactor pressure vessel to perform work. Coolant water not vaporized into steam on passing through the fuel core 14 continues upward with the vaporized steam through the shroud head and separators, then it reverses direction and again flows downward in the annular downcomer along with added condensate water from work expended and condensed steam. One category of boiling water nuclear fission reactor plants 10 employ coolant water circulating pumps comprising a multiplicity of impellers positioned in a generally circular arrangement around the inside bottom of the reactor pressure vessel 12 and connected by means of drive shafts passing through the pressure vessel to externally mounted drive motors, as shown in FIG. 1 and 2. Typically with such a coolant circulating pump arrangement, bore openings 20 are provided in the lower portion of the reactor pressure vessel 12 having a tube-like section 22, or hollow cylindrical stub, extending inward and upward around the bore openings. The bore openings 20 and the adjoining inward tube sections 22 are generally located in the curved bottom of the pressure vessel 12 in an annular peripheral area between the pressure vessel side wall and a core shroud support 24. In conjunction with an arrangement for coolant water circulating pumps, a hollow cylindrical housing 26 is located concentrically upward into each pressure vessel bore opening 20 and the adjoining inward tube section 22 and thus positioned fixed by a weld 28 around its upper annular end to the surrounding tube section 22. A portion of the hollow cylindrical housing 26 projecting outward and downward from the reactor pressure vessel 12 supports a drive motor casing 29, which provides an enclosure for the electrical motor 30. Enclosed within the hollow cylindrical housing 26 is a drive shaft 32 extending from the motor 30 through the reactor pressure vessel bore opening 20 to a pump impeller 34 positioned within the diffuser 35 in an opening in the pump deck 36 within the lower portion of the reactor pressure vessel 12. Thus arranged, the motor driven pump impellers positioned in the opening 38 through pump deck 36 and spaced around the lower portion of the reactor pressure vessel 12 draw coolant water downward within the annular downcomer area outside of the fuel core shroud 16 and force the coolant water below the core shroud 16 then up through the heat producing fuel core 14. This circulation of coolant water is continuously repeated with the loss due to evaporation into steam of a portion of the liquid coolant water being made up by recycled steam condensation returned from performing work such as driving a turbine. The substantially continuous operation of such nuclear reactor coolant water circulation pumps over prolonged periods of time within an environment of radiation, high temperatures and pressures, and vibrations, creates a need for providing routine periodic maintenance inspections and/or service of the pump components. For example, it may be necessary to repair weld the annular end of the hollow cylindrical housing 26 and the surrounding tube-like section 22 or stub extending from the bore opening 20 in the pressure vessel 12 or to replace or recondition the weld 28 which may be fractured or fatigued. Since the nuclear reactor pressure vessel 12 is substantially filled with coolant water 18 for submerging the heat producing fuel core 14 and carrying away the produced heat by routine circulation of the coolant water, maintenance service or repairs for such coolant water circulating pump components must be carried out without removing or substantially reducing the water contents of the reactor pressure vessel 12. In accordance with this invention a unique self-aligning sealing system 40 is provided which can be installed from overhead extending down through the substantial depth of the pool of coolant water contained within the reactor pressure vessel 12, following removal of the pump impeller 34 from the drive shaft 32 and the diffuser 35 from overhead. Thus with the novel sealing means of the invention, maintenance personnel, after removal of the reactor pressure vessel top head and steam dryer, operating from above with remote underwater tools, lift away the pump impeller 34 and the diffuser 35 and then apply the self-aligning seal of system 40 to the upper end of the tube-like section 22 or stub projecting inward and up from the reactor pressure vessel 12 bore opening 20. The self-aligning sealing system 40 for boiling water, nuclear fission reactor plants 10 of this invention utilized in carrying out maintenance service of submerged coolant water circulating pumps, as shown in FIGS. 3 and 4, includes the following components. This system 40 comprises a cylindrical seal unit 42 having an elongated side wall 44 and open bottom with a tapered inward projecting flange 46 containing a seal for abutting with and closing off the upper circular open end of the tube-like section 22 or stub projecting inward and up from the reactor pressure vessel 12 bore opening 20. The cylindrical seal unit 42 has a closed top 48 provided with a central upward projecting shaft 50 having a knob 52 on the upper end for gripping and handing under water by means of a remote handling device such as a pole or other device with an end grip for grasping and releasing the knob 50. Preferably a collar flange 54 acting as a stop member is provided on the shaft 50 between the cylinder closed top 48 and the knob 52 as illustrated. Additionally the cylindrical seal unit, 42 is provided with a multiplicity of generally equally spaced, or symmetrically arranged vertical grooves 56, providing guide members, in the external side wall of the cylindrical unit extending downward from the closed top 48. For example, at least three equally spaced symmetrical grooves 56 extending down the side wall 44 of the cylindrical seal unit 42 is preferred. This self-aligning seal system 40 additionally comprises a seal guide member 58, as shown in FIG. 5, having a generally horizontal disk top 60 of slightly large diameter than the closed top 48 of the cylindrical seal unit 42, and is provided with a central opening 62. Opening 62 is of adequate diameter to receive passing therethrough the shaft 84 and knob 52 projecting upward from the closed top 48 of the cylindrical seal unit 42, but of a smaller diameter than the collar flange 54 thereby providing a stop to prevent traversal therethrough after the seal guide member 58 has been mounted on cylindrical seal unit 42 and the collar flange 54 installed on shaft 50. Disk top 60 of the seal guide member 58 is provided with a multiplicity of depending legs 64 generally equally spaced, or symmetrically arranged and extending down the top disk 60. The depending legs 64 of the seal guide member 58 are spaced around the top disk 60 in an alignment corresponding in position with the vertical grooves 56 in the side wall 44 of cylindrical seal unit 42, and are of a configuration which will closely mate with the vertical grooves 56. The legs 64 are tapered 66 on their outermost surface inward from the top disk 60 down towards their lower end. The seal guide member 58 is designed and constructed to fit down over the cylindrical seal unit 42 with the central opening 62 of the disk top 60 of the guide member 58 passing around the shaft 84 and knob 52 of seal unit 42 and resting superimposed upon the closed top 48 of the seal unit 42. Additionally, the legs 64 depending down from the disk top 60 of guide member 58 with their downward and inward tapered outer face 66, which are in corresponding number, position and cross-sectional configuration with the grooves 56 in the side wall 44 of the cylindrical seal unit 42, mate within and slide down within the grooves 56 to form a composite self-aligning seal system 40. The collar flange stop 54 is installed on the shaft 84 above the top disk 60 of the seal guide member 58 following assembly of the member over and embracing the cylindrical seal unit 42 within the top disk 60 and depending legs 64. The tapered legs 64 depending from the top disk 60 of the seal guide member 58 are of a greater circumference around their outer periphery adjacent to the disk top 60 than the circumference of the cylindrical seal unit 42 whereby they each project radially outward from its respective mating groove 56 adjacent to the upper portion of the cylindrical seal unit 42, and its closed top. From their upper projections the legs 64 fit into the vertical grooves 56 taper downward and inward at least to the outer surface of cylindrical side wall 44 and its circumference. This arrangement provides a wedge shaped configuration which is self-aligning upon moving downward into an opening such as the opening 38 in the pump deck 36. Thus arranged or assembled, the composite seal system 40, can be gripped by means of the knob handle 52 with a remote operating machine and manipulated by personnel from overhead deep down within a reactor pressure vessel 12 under water to a position just above a diffuser 35 opening 38 in the pump deck 36 and the open annular end of the tube-like section 22, or stub, extending up around a bore opening 20 in the bottom of the reactor pressure vessel 12. With the pump impeller 34 and diffuser 35 removed, the tapered legs 64 guide and center the cylindrical seal unit 42 down within an opening 38 in the pump deck 36 and in turn the concentrically underlying annular end of the tube-like section 22 or stub until the inward projecting flange 46 of the cylindrical seal unit 42 adjoins the annular upper end of the tube-like section 22 or stub, sealing off same against coolant water loss therethrough when components such as the drive motor 30 and drive shaft 32 are to be removed from below for any need maintenance service in a dry field or environment. |
054066018 | abstract | A transport and storage cask for spent nuclear fuel. A cask body having one open end receives a basket in such a manner so as to maintain a steady state gap between the cask body and basket. Centering keys on the cask body and basket provide the centering function. For ease of manufacture, the basket is formed from multiple layers of rowed carbon steel plates that have complementary grooves in their mating surfaces to form fuel cell channels when the plates are assembled together. Bands are attached together around the plates at narrowed diameter portions on the plates to hold the plates in their assembled position. The gap between the basket and cask body allows the cask to withstand a fire transient without transmitting heat from the fire into the basket. |
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description | The United States Government has rights in this invention pursuant to Contract No. W-7405-ENG-36 between the United States Department of Energy and the University of California for the operation of the Los Alamos National Laboratory. The present invention is generally related to nuclear fuels. More particularly, the present invention is related to methods and apparatuses for developing microstructured nuclear fuels for advanced nuclear power cycles. Nuclear fuel structures that will allow most fission products to separate from the fissile material during burn-up are also described. Like coal, oil and natural gas, uranium is an energy resource that must be processed through a series of steps to produce an efficient fuel for use in the generation of electricity. Each fuel has its own distinctive fuel cycle; however, the uranium or ‘nuclear fuel cycle’ is more complex than the others. To prepare uranium for use in a nuclear reactor, it undergoes the steps of mining and milling, conversion, enrichment and fuel fabrication. These steps make up the ‘front end’ of the nuclear fuel cycle. After uranium has been used in a reactor to produce electricity it is known as ‘spent fuel’ and may undergo a further series of steps including temporary storage, reprocessing, and recycling before eventual disposal as waste. Collectively these steps are known as the ‘back end’ of the fuel cycle. Current processing methods require uranium to be in the form of a gas before it can be enriched, the natural uranium from the processed ore is converted into the gas uranium hexafluoride (UF6). Enriched UF6 is transported to a fuel fabrication plant where it is converted to enriched uranium dioxide (UO2) powder (typically 3-4% U-235 with the remaining uranium mostly U-238) and pressed into small pellets. These pellets are inserted into thin tubes, usually of a zirconium alloy (zircalloy) or stainless steel, to form fuel rods. The rods are then sealed and assembled in clusters to form fuel elements or assemblies for use in the core of the nuclear reactor. Some 25 tonnes of fresh fuel is required each year by a 1000 MWe reactor. Spent fuel assemblies taken from the reactor core are highly radioactive and give off a lot of heat. They are therefore stored in special ponds, which are usually located at the reactor site, to allow both their heat and radioactivity to decrease. The water in the ponds serves the dual purpose of acting as a barrier against radiation and dispersing the heat from the spent fuel. Spent fuel can be stored safely in these ponds for long periods. It can also be dry stored in engineered facilities. However, both kinds of storage are intended only as an interim step before the spent fuel is either reprocessed or sent to final disposal. The longer it is stored, the easier it is to handle, due to decay of radioactivity. There are two alternatives for spent fuel: 1) reprocessing to recover the usable portion of it, and 2) long-term storage and final disposal without reprocessing. The present inventors have determined that microstructured fuels contain fissile material structures with micrometer-scale dimensions dispersed in a matrix material. Most fission products escape from the fissile material structures and come to rest in the matrix material. This can allow a much cheaper separation of the fission products and fissile material, after the fuel is removed from the power system and allowed to cool for a number of years. The present inventors have also found that microstructured nuclear fuels can be manufactured to contain micrometer-sized fissile material structures (e.g., typically in the shape of spheres) spaced uniformly in matrix materials. The advantage of such structures is that they will allow most fission product elements to escape from the fissile material and come to rest in the matrix material. According to a feature of the present invention, microstructured nuclear fuel is provided that is adapted to a nuclear power system. The fuel includes fissile material structures of micrometer-scale dimension dispersed in matrix material. In accordance with another feature of the present invention, microstructured nuclear fuel is produced in a fluidized-bed chemical vapor deposition (CVD) reactor including a gas inlet for providing controlled gas flow into a particle coating chamber, a lower bed hot zone region to contain powder, and an upper bed region to enable powder expansion. In accordance with another feature of the present invention, at least one pneumatic and electric vibrator is operationally coupled to the particle coating chamber for causing vibration of the particle coater to promote uniform powder coating within the particle coater during fuel processing. In accordance with another feature of the present invention, an exhaust associated with the particle coating chamber can provide a port for placement and removal of particles and powder. In accordance with another feature of the present invention, during use of the fuel in a nuclear power reactor, fission products escape from the fissile material structures and comes to rest in the matrix material. The escape of fission products in the fuel occurs because the fission process generates two fission product nuclei with 167 million electron volts of kinetic energy that travel about 2-10 micrometers through the surrounding material before coming to rest. The average distance traveled depends on the mass and charge of the fission product nucleus and the stopping power of the surrounding material. The relatively uncommon fission events (˜0.2%) that generate three product nuclei will behave in a similar manner. After a period of use in a nuclear power reactor and subsequent cooling, separation of the fissile material from the matrix containing the embedded fission products will provide an efficient partitioning of the bulk of the fissile material from the fission products. This partitioning process involves processes such as size reduction and separation by density or selective dissolution of the matrix material, e.g., oxidation of a carbon matrix to carbon dioxide or dissolving of a magnesium oxide matrix in alkaline aqueous solution. The fissile material can be reused by incorporating it into new microstructured fuel. The fission products and matrix material can be incorporated into a waste form for disposal or processed to separate valuable components from the fission product mixture. In accordance with yet another aspect of the present invention, carbon is used as the matrix material as an example of one type of microstructured nuclear fuel. The carbon can be used as the matrix between uranium dioxide fuel particles that are 0.5-5.0 micrometers in average diameter. The goal would ideally be to coat these <5.0 micrometer UO2 particles with a 1-5 micrometer thickness of carbon. The coated particles can then be packed into a fuel pellet and the thickness of the carbon coating will determine the spacing between the UO2 particles. In accordance with yet another aspect of the present invention, a fluidized-bed chemical vapor deposition (CVD) reactor is provided for carrying out methods of the present invention. The reactor includes a gas inlet (e.g., for propylene and an inert gas such as helium), and a particle coating chamber including a lower bed hot zone region to contain the powder and upper bed region to allow bed expansion. In accordance with a feature of the chemical vapor deposition (CVD) reactor, pneumatic and electric vibrators can be provided to promote uniform powder movement and an exhaust. Efficient separation of the fissile materials from the fission products using the present invention can greatly reduce costs for an advanced fuel cycle relative to presently used aqueous or proposed pyrochemical separation methods. Referring to FIGS. 1A and 1B, microstructured fuels contain fissile material structures with micrometer-scale dimensions dispersed in a matrix material. Most fission products escape from the fissile material structures and come to rest in the matrix material. This can allow a much cheaper separation of the fission products and fissile material, after the fuel is removed from the power system and allowed to cool for a number of years. FIGS. 1A and 1B is a schematic drawing that illustrates the tracks of fission products (arrows) as they escape from an array of fissile material structures. In FIG. 1A an array of spherical particles of fissile material (e.g., enriched UO2) is pictured with pairs of arrows generally representing the tracks of fission products as they travel out of the fissile particle and eventually come to rest in the surrounding matrix material that separates the spherical particles. In FIG. 1B a cylindrical array of channels filled with fissile material is shown with arrows generally representing the tracks of fission products as they travel out of the fissile material and come to rest in the matrix material separating the channels. The diameter of the spherical particles or channels of fissile material must be on the order of 2-3 micrometers or smaller to allow most of the fission products to escape the matrix. The spherical array of particles will provide the fuel structure with the most complete separation of fission products from fission material, but the cylindrical array can also be quite good and provides additional methods for production of microstructured fuels. Based on previous experience, coating enriched UO2 fuel particles within a fluidized-bed was determined by the present inventors to be a very practical approach to preparing a microstructured nuclear fuel that will allow separation of the fission products and fissile material during use in the reactor. According to features of the present invention, several materials including tungsten boride (WB) and hafnium dioxide (HfO2) were evaluated as surrogate materials for the 0.5-5.0 micrometer UO2 powder to evaluate fluidization characteristics and required coating reactor design. HfO2 powder (nominal 5.0 micrometers average diameter) was eventually selected as a good substitute for UO2 particle size and density. Minimum fluidization velocity was calculated for the HfO2 and numerous experiments were conducted in a glass reactor similar to the actual high temperature coating reactor geometry to evaluate gas velocity, powder motion, and influence of external vibration. Use of the HfO2 surrogate permitted validation of the present invention and analysis of new challenges without concern for addressing the radiological and pyrophoric issues related to using finely divided UO2 powder. Referring to FIG. 2, illustrated is a cross-section diagram of a fluidized-bed coater system design. A chemical vapor deposition (CVD) reactor 100 is constructed incorporating a gas inlet 110 (e.g., for propylene and an inert gas such as helium), a particle coating chamber 120 including a lower bed hot zone region 123 to contain the powder and an upper bed region 127 to allow expansion, and an exhaust 150. The gas inlet 110 can be water cooled 115. The lower bed “hot zone” region 123 and upper bed “free board” region 127 can be formed from graphite, the combined zones representing a UO2 particle coating chamber 120. An inductance coil 190 surrounding the particle coating chamber 120 can provide heat during particle processing. Numerous experiments were conducted using this reactor 100 to coat the HfO2 surrogate with pyrocarbon. Some difficulty, however, was encountered in keeping the powder circulating within the desired lower bed hot zone region 120 because it is easily entrained in the incoming gas stream and carried downstream (leaving through the exhaust 150). This appears to be due to chances in particle adhesion properties during the coating process and are addressed with adjustments in reactor design and experimental technique. The HfO2 powder was successfully coated with pyrocarbon, but less than 0.5 micrometers thick because of the short residence time in the lower bed hot zone. It was anticipated that the pyrocarbon coating can be grown to 1.0-2.0 micrometers by retaining the powder within the lower bed hot zone 123. Referring to FIG. 3, an improved reactor 200 was developed that incorporated changes permitting better control of the fine UO2 powder. These improvements included vibratory means 210 and/or 220, such as pneumatic vibrators, integrated with, coupled to, or in otherwise mechanical communication with, the particle coating chamber 120. One vibrator 210 will preferably be associated with the lower bed 123, while a second vibrator 220 can also be incorporated in the system in association with the upper bed region 127. It is preferred for one type of microstructured nuclear fuel that carbon be used as the matrix between uranium dioxide fuel particles that are 0.5-5.0 micrometers in average diameter. For example, utilizing the present invention, coating of <5.0 micrometer UO2 particles can be accomplished with a 1-5 micrometer thickness of carbon. The coated particles can then be packed into a fuel pellet and the thickness of the carbon coating will determine the spacing between the UO2 particles. Fluidized-bed chemical vapor deposition (CVD) technique is a viable approach for coating UO2 with pyrocarbon or graphite with well-controlled thickness. The CVD methods can also be used to deposit many other coatings such as carbides, nitrides and oxides of the elements Mg, Al, Si, Zr, Y, Ce, Nb and Ta (such as MgO, Al2O3, SiC, ZrN, Y2O3, CeO2, NbO2 and TaO2). Additionally, the CVD methods can be used to deposit coatings such as silicates, phosphates and aluminates of the elements Mg, Zr, Y, Ce, Nb and Ta (such as MgAl2O4, ZrSiO4, YPO4, CePO4, NbSiO4 and TaSiO4). These other coating materials provide similar properties to the fuel system by inhibiting chemical reactions between the matrix and fuel materials. These coatings, in any combination, can be applied to give the fuel material more than one matrix layer. Before coating the UO2 particles, surrogate powders in the size and density range of interest were used to determine coating conditions and the most effective coater geometry. Hafnium dioxide (aka, Hafnia)(HfO2 density 9.7 g/cc) and tungsten boride (WB density 10.77 g/cc) are good matches to UO2 (10.96 g/cc), available commercially in particle sizes <10.0 micrometers at relatively low cost, and stable at the temperatures required for coating (1000-1200 C). A preliminary investigation of the surrogate powders, WB (CERAC) and HfO2 (Wah Chang) was undertaken. Laser scattering particle size analysis for the WB yielded a mean diameter of 1.558 micrometers with a standard deviation of 1.439 micrometers. Analysis of the HfO2 powder shows a mean diameter of 5.0 micrometers. A test system similar in design to the CVD system 100 of FIG. 2 was first used for coating of particles. A quartz-glass fluidized-bed coater was used at low-temperature to visualize gas flow and bed behavior of the fine particles within the area representing the particle coating chamber 120 of the CVD system 100. The powders were evaluated in a glass fluid bed suitable for observation and of the same geometry as the high temperature graphite reactor. An estimate of the minimum fluidization velocity was obtained. FIG. 4 illustrates the results for particles with an 11 g/cc (WB) density fluidized in helium gas at one atmosphere and assuming porosity at fluidization of 0.75. For the particle size range of interest (0.5-5.0 micrometers), the fluidization velocity is very low. For the test reactor geometry, flow rates were on the order of 1 sccm (standard cubic centimeter per minute). The WB particles could not be fluidized at flow rates ranging from 1 sccm to 10,000 sccm. This is not surprising; at low flows there is insufficient energy to overcome the cohesive forces between particles. This material is a type C powder by the Geldart classification and is normally not fluidizable. However, type C powders can often be fluidized as agglomerates. In this case the bed may rise as a plug and breakup behaving as large particles. This was not observed for the WB powder. Conditioners are often added to type C powders to reduce agglomeration. These are very fine submicron powders that adhere to the primary particle and effectively screen the attractive Van der Waals forces between particles. Carbosil is often used for this purpose. The addition of 2% by weight of carbosil to the WB had no effect. If a conditioner is used the submicron particles attached to the primary particles would necessarily be incorporated into any coating, for this reason a carbon black was also tested as a conditioner. Again no effect was observed on the WB. In accordance with a feature of the present invention, vibration of the particle coating chamber 120, at one or both of the upper 123 and/or lower 127 beds, can also assist in the fluidization of fine particles by breaking up agglomerates. Again this was ineffective for the fluidization the WB material; however, vibration was capable of producing what appeared to be very fine or possible individual particles in the upper bed region 127 of the system 200. Apparently, vibration was able to break up the bed producing cracks and bubbles. This allows gas to pass through the dense bed at high velocity without fluidizing. The gas was found to entrain fine particles and form a low-density phase above the upper bed 127. Therefore, it seemed likely to the inventors that particles can be coated in this low-density phase. With the HfO2 powder (density 9.8 g/cc) the fluidization of individual particles also required low velocities similar to the results of FIG. 4. Fluidization was not observed at the low gas velocities but it was observed at higher velocity as agglomerated powder. A plot of pressure drop across the HfO2 powder bed versus flow rate is shown in FIG. 5. From this data a minimum fluidization velocity of approximately 30 cm/s was estimated. The agglomerates visually appeared to be on the order of 1 micrometers in diameter. From the measured velocity, a particle diameter was calculated. Assuming an agglomerate density of 50%, a particle diameter of 4 micrometers was calculated in reasonable agreement with observation. The HfO2 fluidization data was taken with bed agitation (tapping on the side of the bed). The application of vibration was also effective. To insure absorbed moisture was not contributing to cohesive interparticle forces thermo-gravimetric analysis of the HfO2 powder was preformed. No weight loss was observed up to 400 C. The effect of conditioners has not been evaluated. Referring again to FIG. 3, a graphite sleeve 215 machined to slip over the high temperature lower bed 123 was fabricated. The vibrator 210 is coupled to the sleeve 215 through a vacuum chamber 250 and used to agitate the lower bed 123 during deposition. Two modes of operation can be envisioned and the choice will depend on the behavior of the UO2 powder. If the powder behaves like the HfO2 then it can be fluidized as agglomerates. The agglomerates will be continually broken and reformed due to the vibration. This will allow for even coating of individual particles. For this process the system 200 can be filled with powder reaching into the upper bed region 127 shown in FIG. 3. The particles will be fluidized as agglomerates in the upper bed region 127 where the system 200 will be heated in excess of 1000 C in a helium/hydrocarbon atmosphere produce pyrocarbon coatings. Vibration can also be applied in the lower bed region 123 outside the heated zone. With the improved system's setup, a second possible mode of operation suitable for the nonfluidizable WB particles is also possible. In this case powder will be filled only in the lower bed region 123, which will be vibrated. The entrained powder resulting from vibration will be carried into the heated upper bed region 127, which will now function as the freeboard region. In this region particles are continuously rising out of and falling back into the lower bed region 123 with some of the finer particles being carried out of the particle coating chamber through the exhaust 150. In this design a third larger diameter region 129 can be added as shown in FIG. 3. The large diameter region 129 will reduce the gas velocity allowing for some of the very fine particles that have been carried out of the upper bed region 127 to fall back down into the lower bed region 123. The application of vibration to this large diameter region 129, or the upper bed region 127, should assist in removing any of those particles that may stick to the inside walls of the particle coating chamber 120. In accordance with a method of carrying out the invention, reference is now made to the flow diagram 500 of FIG. 6. As shown in FIG. 6, the method for producing microstructured nuclear fuel is started as shown by block 510 by providing fissile material particles (e.g., UO2) within a particle coating chamber associated with a CVD reactor. The fissile material particles are heated, as shown in block 520. As shown in block 530, matrix material precursors are provided to the fluidized-bed CVD reactor through a gas inlet along with an inert gas, such as helium. As shown in block 540, the matrix material is deposited on the fissile materials particles to give a uniform coating of 1-5 micrometers in thickness. As shown in block 550, coated particles may be removed from the coating chamber through the exhaust or the inlet orifice. As shown in optional block 535, the particle coating chamber is vibrated during processing of the fissile material in steps 530 through 550. The mode of operation depends on the behavior of the powder. The preferred method may be the fluidization of agglomerates like the HfO2 powder. This provides a dense bed and a more efficient process. Alternatively the freeboard coating method, which does not rely on agglomerates, may have the advantage of better quality coatings. However, the “fluidized phase” is not as dense and will probably require longer processing times. Experiments were performed on hafnium oxide powder using the setup described above without the second vibrator. Approximately 40 grams of powder were loaded in the graphite tube. The volume of the powder extended above the lower bed region. Helium and propylene gas mixtures were introduced through the bottom of the bed in known amounts through mass flow controllers. The experiments were run at a nominal exhaust pressure of 590 Torr. The gas composition was maintained at 30 volume percent propylene. The first run was performed at a total gas flow of 40 slpm (standard liters per minute) and a temperature of 900 C. The initial pressure drop across the bed was 6.0 Torr but steadily decreased over approximately one hour to roughly 1.0 Torr. No powder was left in the reactor. A second run was performed at a total flow of 28 slpm. Again, most of the powder was carried out of the reactor. Powder was retrieved from the exhaust section of the system and appeared to be partly coated. The white powder now appeared gray and observation under a low magnification microscope revealed a mixture of coated and non-coated agglomerates. The removal of the powder out of the reactor through the exhaust under the coating conditions was unexpected. The room temperature fluidization experiments did not indicate such rapid removal of powder by entrainment in the exhaust gas at the chosen gas flows. While the gas within in the hot zone at the operating temperature for coating may have a higher velocity than the room temperature experiments, this would not account for the rapid removal of the powder. A more likely explanation is that once some degree of coating occurs the cohesive behavior of the powder is altered and the fluidization conditions are now different. The partly coated powder does not appear to be agglomerated to the same degree as the pristine powder. The tap density of the partly coated powder was measured at 2.86 g/cc while that of the non-coated powder is 2.2 g/cc. This represents a 60% increase in density and is consistent with the idea that the coated powder may be less cohesive. This suggests that if the gas flow remains constant during the process as particles are coated the agglomerate size decreases and more powder will be entrained. The initial coating appears to produce a more fluidized powder (the addition of carbon black and carbosil to the powder was another approach to accomplishing this as described above). With this in mind, additional experiments were performed with the aim of producing a sufficient quantity of “conditioned” powder to explore its fluidization properties. The system was modified such that the powder bed was lifted into the wider heated zone of the graphite tube by inserting graphite felt into the lower bed region. The graphite felt then served to hold the powder bed up and function as a porous gas distributor. In this arrangement the majority of the powder bed will be in the hot zone at all times. The total gas flow was set at 0.714 slpm. This low flow rate combined with the wider diameter graphite tube resulted in a low gas velocity and no fluidization. The coating process was preformed in a vibrating bed only. This was sufficient to develop thin coatings on the powder and thereby condition them. The process was run for 100 minutes at 1010 C. All the powder remained in the graphite tube. The top third of the bed was thoroughly coated and appeared black. The lower two thirds remained white. The coated and uncoated regions are likely due to a combination of position in the hot zone and a residence time effect. The lower part of the bed is not as hot as the upper portion. The coated powder produced by this modification compared with uncoated powder is shown in a photograph provided in FIG. 7. The HfO2 powder was successfully coated with pyrocarbon, but scanning electron microscope images indicated the coating was less than 0.5 micrometers thick because of the short residence time in the lower bed hot zone. It is anticipated that the pyrocarbon coating can be grown to 1.0-2.0 micrometers by retaining the powder within the hot zone. Additional experiments have indeed yielded thicker pyrocarbon coatings, as anticipated. Additionally, it should be noted that multiple coatings are possible. Also, each coat does not need to be the same compound. However, if multiple coatings are utilized, the total coating should preferably not exceed 5 micrometers in thickness. The use of the HfO2 surrogate permitted validation of previous experience fluidizing fine powders and analysis of new challenges without concern for addressing the radiological and pyrophoric safety issues related to UO2 powders containing particles in the micrometer size range. A fine-divided UO2 powder synthesized from a uranium(VI) oxalate precipitation process was obtained from the Y-12 Plant in Tennessee for use in UO2 coating tests. While the preferred embodiments of the present invention have been illustrated and described, it will be apparent to those of ordinary skill in the art that various changes, modifications or variations may be easily made without deviating from the scope of the invention. |
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053655540 | summary | The present invention relates to an instrumentation probe for use in the on-line measurement and recording of physical parameters within a fluid duct without connection to external instrumentation. A particular application of the instrumentation probe is to the measurement and recording of structural, thermal and hydraulic data within a fuel channel of a fluid-cooled nuclear reactor, and such data may include fuel channel coolant pressure, in-channel flow magnitudes and variations, fluid and pressure tube temperatures, pressure tube vibration, fuel bundle displacements, and the like. However, the invention is not exclusively concerned with nuclear applications, but is applicable to analogous systems in which physical parameters within a fluid duct are to be measured and recorded. In the case of fluid-cooled nuclear reactors, problems associated with fuel bundle damage have underlined the need to have a capability to obtain accurate measurement data directly from within any fuel channel during any phase of reactor operation. The inability to obtain structural integrity measurement data directly from locations within fuel channels during reactor operation has resulted in uncertainty as to quantifying activity within the fuel channels, such activity relating more particularly to the pressure tubes, coolant flow and fuel bundles. This uncertainty has in turn resulted in severe information constraints regarding structural and thermal-hydraulic concerns with fuel channels. In order to deal with these concerns, accurate on-line measurements within the fuel channels are required. According to the present invention, there is provided an instrumentation probe for measuring and recording one or more physical parameters within a fluid duct, which may be a fuel channel of a nuclear reactor, although the invention in its broadest aspect is not limited to nuclear applications. The instrumentation probe basically comprises an elongate support frame adapted to be located within the fluid duct, and self-contained measuring and recording means mounted on the frame. The self-contained measuring and recording means comprises one or more sensors each responsive to a physical parameter to be measured and recorded, and a scribe coupled to the or each sensor, which scribe cooperates with a volume driven recording chart, such as a rotary drum, disk or tape for recording variations in the respective physical parameter over time. The recording chart is driven by a longitudinally extending rotary shaft carrying an impeller which is responsive to fluid flow for rotating the shaft at a constant speed, the shaft being coupled to the recording chart through a speed reducing mechanism. In the case of a nuclear reactor application the fluid flow is itself maintained constant and so the constant speed of the recording chart is thereby maintained. In applications where the fluid flow is not constant it is necessary to govern the shaft rotation by any suitable speed governing device, or alternatively to apply signal reference markings to the recording chart at regular intervals of time during monitoring periods so as to provide a time scale. It will be appreciated that, in the latter case, the chart will be driven at a measurable speed since the speed can be later deduced from the known time periods between the applied reference markings. |
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abstract | A process for the treatment of radioactive graphite which includes the following steps: (i) reacting the radioactive graphite at a temperature in the range of from 250xc2x0 C. to 900xc2x0 C. with superheated steam or gases containing water vapor to form hydrogen and carbon monoxide; (ii) reacting the hydrogen and carbon monoxide from step (i) to form water and carbon dioxide; and (iii) reacting the carbon dioxide of step (ii) with metal oxides to for carbonate salts. The process enables radioactive graphite, such as graphite moderator, to be treated either in-situ or externally of a decommissioned nuclear reactor. |
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abstract | A storage apparatus with radiation shielding for spent nuclear fuel includes a fuel basket comprising elongated fuel storage tubes each defining an open cell configured to hold a nuclear fuel assembly. Gamma radiation attenuation inserts are nested inside at least some of the storage tubes. The inserts each comprise elongated open-ended tubular bodies which may have a rectangular cuboid configuration with square cross section. The inserts are composed of a dense metallic material selected for blocking gamma radiation and may have high thermal conductivity for effective heat dissipation from the decaying nuclear fuel. Attenuation inserts can occupy some or all perimeter tubes to provide shielding against gamma radiation emanating in a lateral direction from the fuel basket. The inserts may include upper and lower securement features for detachable fixation to the storage tubes, and air/gas flow cutouts. |
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abstract | A power generation system includes an inert gas power source, a thermal/electrical power converter and a power plant. The thermal/electrical power converter includes a compressor with an output coupled to an input of the inert gas power source. The power plant has an input coupled in series with an output of the thermal/electrical power converter. The thermal/electrical power converter and the power plant are configured to serially convert thermal power produced at an output of the inert gas power source into electricity. The thermal/electrical power converter includes an inert gas reservoir tank coupled to an input of the compressor via a reservoir tank control valve and to the output of the compressor via another reservoir tank control valve. The reservoir tank control valve and the another reservoir tank control valve are configured to regulate a temperature of the output of the thermal/electrical power converter. |
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description | This application is the U.S. National Phase application under 35 U.S.C. §371 of International Application No. PCT/EP2014/069644, filed on Sep. 16, 2014, which claims the benefit of European Patent Application Nos. 13186594.1, filed on Sep. 30, 2013 and 14177486.9, filed on Jul. 17, 2014. These applications are hereby incorporated by reference herein. The invention relates to the field of differential phase contrast imaging (dPCI). In particular, the invention relates to a device and a method for X-ray differential phase contrast imaging and X-ray attenuation imaging. X-ray differential phase contrast imaging (dPCI) visualizes the phase information of coherent X-rays passing through a scanned object. For example, the coherent X-rays may be generated by a source grating between an incoherent X-ray source and the scanned object. In addition to classical X-ray attenuation imaging, dPCI may determine not only the absorption properties of the object along a projection line, but also the phase-shift of the transmitted X-rays. After the object, a phase-shifting grating (also known as phase grating) is placed, which generates an interference pattern that contains the required information about the beam phase-shift in the relative position of its minima and maxima, typically in the order of several micrometers. Since a common X-ray detector may not be able to resolve such fine structures, the interference pattern is sampled with an analyzer grating (also known as absorber grating), which features a periodic pattern of transmitting an absorbing strip with a periodicity similar to that of the interference pattern. The similar periodicity produces a Moiré pattern behind the grating with a much larger periodicity, which is detectable by a common X-ray detector. For performing dPCI there exist essentially two different system geometries: planar 2D detection and slit-scanning systems. In planar 2D detection, a 2D detector array takes an entire projection image in a single X-ray exposure and the phase acquisition has to be realized by a process called “phase-stepping” with, for example, 4, 8 or 16 exposures, in which one of the source grating, the phase grating and the absorber grating is moved relative to the other two gratings. In the slit scanning approach, the woman's breast is scanned by a scan arm or gantry movement below the breast. The redundancy of the data acquisition by means of a typical arrangement of a number of parallel detector lines can be exploited to eliminate the need for phase-stepping and the gratings need not be moved with respect to each other. Hence, the phase-acquisition can be implemented in the ordinary scanning motion, as, for example described in WO2013/004574A1. However, in both cases, the differential phase contrast technique, characterized by its use of three gratings, has the disadvantage that approximately 50% of the X-rays that have passed through the breast tissue are actually absorbed by the absorber grating and thus are lost for imaging or are not used for imaging. Sometimes it may be desirable to perform not only dPCI imaging but instead or in addition attenuation imaging without gratings. For example, W2012/0099702 A1 shows a differential phase contrast imaging device with a movable grating. Calibration of the detector and dPCI system may be difficult, if the gratings cannot be removed, and calibration e.g. for drift or other detector performance changes may be required more than just at the factory/initial installation but on a regular basis, and at least after every maintenance. WO 2013/111050 A1 discloses an x-ray system provided with grating structures, such grating structures comprising sets of slits, which sets have mutually different directions. U.S. 2013/0202081 A1 discloses a detector arrangement for phase contrast imaging comprising movable gratings. US 2013/0230135 A1 discloses a joint imaging apparatus comprising gratings. WO 2011/070488 A1 discloses a device for phase contrast imaging comprising a grating being movable out of the X-ray beam. In a 2D detection system, the absorber grating covers the entire detection area and has to be removed from the entire field-of-view (up to about 30 cm×40 cm), i.e. over a rather long distance. This may result in a bad alignment of the absorber grating and the phase grating after the movement and may degrade the image quality. Additionally, the preferred movement direction in a 2D mammographic system would be the anterior-posterior direction, shifting the absorber grating out of the X-ray imaging path away from the breast. There may be a need to provide an X-ray imaging device that is able to easily switch between pure attenuation imaging and dPCI imaging (while providing attenuation, differential phase and scatter imaging) There also may be need to provide a device that is fast, simply, accurately and reliably switched between these two operation modes. These needs may be met by the subject-matter of the independent claims. Further exemplary embodiments are evident from the dependent claims and the following description. An aspect of the invention relates to an X-ray differential phase contrast imaging device, for example to a mammography device. According to an embodiment of the invention, the imaging device comprises an X-ray source for generating an X-ray beam; a source grating (G0) for generating a coherent X-ray beam from a non-coherent X-ray source (20); a collimator comprising slits for splitting the coherent X-ray beam into a plurality of fan-shaped X-ray beams for passing through an object; a phase grating and an absorber grating arranged after the object; and a line detector comprising detector lines for detecting a Moiré pattern generated by the phase grating and the absorber grating from the fan-shaped X-ray beams passing through the object. The X-ray source, source grating, collimator, phase grating, absorber grating and detector are fixed to a common gantry and are movable with respect to the object, such that a number of interference patterns from different positions of the gantry are detectable for reconstructing a differential phase image of the object. At least one grating of the source grating, phase grating and the absorber grating comprises groups of grating lines and transparent areas between the groups of grating lines. Herein, the groups of grating lines and the transparent areas alternate with respect to each other in a direction perpendicular to the direction of the detector lines. At least one grating of the source grating, phase grating and the absorber grating is movable with respect to the gantry, such that: in a first (dPCI) position of the source grating the X-ray beams pass through the grating lines and subsequently pass through the slits of the collimator, and in a second (attenuation imaging) position of the source grating, the X-ray beams pass through the transparent areas and subsequently pass through the slits of the collimator, or, in a first position at least one of the phase grating or the absorber grating the fan-shaped X-ray beams pass through the grating lines, and in a second position at least one of the phase grating or the absorber grating the fan-shaped X-ray beams pass through the transparent areas. Therefore, since the groups of grating lines and transparent areas alternate with respect to each other in a direction perpendicular to the detector lines, in the first position the groups of grating lines project onto the spaces between neighboring detector lines (i.e. the spaces between neighboring detector lines in a direction perpendicular to the direction of the detector lines), whereas in the second position the transparent areas project onto such spaces. Consequently, the X-ray differential phase contrast imaging device according to the invention effectively utilizes the spaces between neighboring detector lines thereby enabling switching from dPCI imaging to attenuation imaging (and from attenuation imaging to dPCI imaging) while circumventing the need to move one or more of the grating(s) out of the X-ray field of view entirely (and the need to move said one or more of the grating(s) into the X-ray field of view, respectively). In a slit scanning device, only the areas of the phase grating and/or absorber grating, which are exposed to X-rays need to have grating lines. The other areas may be made or left transparent to X-rays. In the configuration with all gratings in a position such that the slits are aligned with the grating lines of the gratings, the imaging device is adapted to generate dPCI data, which usually contains information about attenuation, differential phase and scatter of X-rays in the object. When it is desired to make a (pure) attenuation image, the phase grating and/or the absorber grating may be moved into a position, such that they are exposed to X-rays at areas transparent to X-rays without grating lines. In such a way, the attenuation of the gratings during the (pure) attenuation imaging and the X-ray dose delivered to the object (which may be a patient) may be reduced. The phase grating and the absorber grating may be seen as an interferometer of the imaging device, which may be seen as a slit scanning device with a retractable interferometer that allows switching between phase contrast and conventional imaging (mammography). This is achieved by removing (and introducing) the areas with grating lines from the setup. For example, the absorber grating may have a silicon wafer as substrate comprising (50%) absorbing portions (trenches in the wafer filled with gold for forming the grating lines) and transparent areas (plain silicon). An on-off mechanism may be realized by a lateral shift of the absorber grating over a relatively small distance. Contrary to this, in planar 2D detection such as in full-field digital mammography (FFDM), the phase grating and/or absorber grating have to be completely removed from/reinserted into the X-ray field-of-view entirely, over a larger distance. A further aspect of the invention relates to a method for acquiring differential phase image data and attenuation image data with the same device, which may be an image device as described in the above and in the following. According to an embodiment of the invention, the method comprises: moving a grating selected from a phase grating and an absorber grating of the device in a first position, such that the fan-shaped X-ray beams generated by a collimator can pass through groups of grating lines on the grating; acquiring differential phase image data by moving a gantry with the phase grating and absorber grating and a line detector with respect to an object and by detecting X-rays passing through the object, the phase grating and the absorber grating at a plurality of positions of the gantry; moving the grating in a second position, such that the fan-shaped X-ray beams can pass through transparent areas on the grating; and acquiring (pure) attenuation image data by moving the gantry with respect to the object and by detecting X-rays passing through the object at a plurality of positions of the gantry. For example, by simply moving the absorber grating and/or the phase grating by a relative small distance, differential phase imaging may be switched to (pure) attenuation imaging. In mammography, depending on the clinical application, certain parts of the mammography workload will benefit from the differential phase imaging. However, part of the workload may still be carried out in conventional mode (attenuation imaging only). A clinician may desire to choose between modes on a per-patient (or even per-view) basis. These and other aspects of the invention will be apparent from and elucidated with reference to the embodiments described hereinafter. The reference symbols used in the drawings, and their meanings, are listed in summary form in the list of reference symbols. In principle, identical parts are provided with the same reference symbols in the figures. FIG. 1 shows an X-ray imaging device 10 with a movable gantry 12 that is movable in an angular range with respect to an object 14. The X-ray imaging device 10 may be a mammography device and the object 14 may be a woman's breast that is supported on a support platform 18 fixed to the device 10, for example to a frame 16 also carrying the gantry 10. The gantry 10 carries an X-ray source 20 (for example an X-ray tube), a source grating G0, a collimator 22, a phase grating G1, an absorber grating G2, and a line detector 24. The gratings G0, G1, G2, the collimator 18 and the line detector 24 are attached to the gantry 10 at fixed distances with respect to the X-ray source 16. For acquiring image data, a controller 26 of the device 10 moves the gantry 12 in an angular range around the object 14 and acquires a number of projection images with the line detector 24 at different positions along the angular range. A coherent X-ray beam generated by the source grating G0 from the (non-coherent) X-ray beam of the X-ray source 20 is split into fan-shaped beams 28 (one of which is indicated in FIG. 1) by the collimator 22. The collimator 22 is shown in FIG. 2. It comprises a substantially rectangular substrate with a plurality of parallel equidistant slits 30. Usually, the longer sides of the collimator 22 are substantially parallel to the axis of movement of the gantry 12. Returning to FIG. 1, the fan-shaped beams 28 pass through the object 14 and the gratings G1, G2 and fall onto the line detector 24, which is shown in FIG. 3. Also the detector 24 comprises a substantially rectangular shape (seen from the X-ray source 20), wherein the longer sides are usually substantially orthogonal to the axis of movement of the gantry. The detector 24 comprises the same number of detector lines 32 as the collimator comprises slits 30. The detector lines 32, each of which comprises a line of pixel detector elements, are aligned with the slits 30, such that each fan-shaped X-ray beam 28 passing through the slits 30 falls on the respective line detector 32. The slit arrangement and/or arrangement of detector lines 32 may be more complicated than shown in the figures. Slits 30 and/or detector lines 32 may not go through in one piece (otherwise there would not be the need for several slits) and they might be slightly offset. The controller 26 may evaluate the acquired image data and the interference pattern encoded therein, and may determine/reconstruct an image indicating the phase shifts of the object 16. A possible embodiment of the gratings G0, G1 and G2 is shown in FIG. 4. FIG. 4 shows a section of one of the gratings G0, G1, G2. Since only the components of the X-ray beam that in the end fall onto the detector need to be diffracted by the gratings, only the areas of the gratings G0, G1 and G2 need to have grating lines 34 that are in the way of the fan-shaped beams 28. Thus, the grating lines 34 may be grouped in groups 36 of grating lines 34 separated by areas without grating lines between them. However, it is also possible that the only one or only two of the gratings are designed like shown in FIG. 4 and that the other gratings have grating lines over its whole area, which in particular may be the case for the source grating G0. Since the diffraction gratings G0, G1 and G2 attenuate the X-ray beam more or less (at least by the area of the grating lines 34), it may be beneficial to remove them from the X-ray beam, when acquiring attenuation imaging data with the line detector 24. In the embodiment shown in FIG. 1, this is realized with areas 38 (see FIG. 4) transparent to X-rays provided in the gratings G1 and G2. By laterally moving the gratings G1 and/or G2, the gratings may be moved from an “on”-position, in which the X-ray beams 28 run through the groups 36 into an “off”-position, in which the X-ray beams 28 run through the transparent areas 38. According to an embodiment of the invention, at least one grating of the phase grating G1 and the absorber grating G2 comprises groups 36 of grating lines 34 and transparent areas 38 between the groups of grating lines and is movable with respect to the gantry 12, such that in a first “on”-position of the grating, the fan-shaped X-ray beams 28 pass through the grating lines 34, and in a second “off”-position of the grating, the fan-shaped X-ray beams 28 pass through the transparent areas 38. Note that the direction of the grating lines 34 may be substantially normal to the plane of FIG. 1, i.e. that the grating lines may run substantially orthogonal to the movement direction of the gratings G0 and G1. The transparent areas 38 are usually substantially rectangular. The smaller side of the rectangle may be much longer than the distance between two neighboring grating lines 34. Due to the equidistant slits 30 of the collimator 22, the groups 36 and the areas 38 may alternate and may all have the same distances from each other. According to an embodiment of the invention, the groups 36 of grating lines 34 and the transparent areas 38 alternate with respect to each other. According to an embodiment of the invention, the groups 36 of grating lines 34 are equidistant and the transparent areas 38 are equidistant. Usually, the gratings G0, G1 and G2 are manufactured by etching a wafer (substrate) and/or putting a structured metallization on at least one side of the wafer. For example, the wafer may be a thin silicon wafer (300 mu-500 mu thick), which is substantially transparent to X-rays. The grating lines 34 of the gratings G0, G2 may be gold filled silicon trenches that absorb substantially 50% of the radiation impinging orthogonal to the grating G0, G2. The grating lines 34 of the phase grating G1 may be (empty) trenches in the substrate. According to an embodiment of the invention, the grating lines may be metal lines on the substrate and/or may be or may comprise trenches in the substrate. The transparent areas 38 may be provided with unprocessed parts of the substrate/silicon wafer, for example parts with no gold and no trenches at all. According to an embodiment of the invention, at least one of the gratings G0, G1, G2 comprises a substrate transparent for X-rays and the transparent areas 38 are areas on the substrate without metallization. However, it is also possible that the transparent areas 38 are provided by holes in the substrate. According to an embodiment of the invention, the transparent areas 38 may be or may comprise holes in a substrate of the grating. As further indicated in FIG. 1, grating G0 may be removed from the X-ray beam 28 with another method. The grating G0 may be attached to a hinge 48 and may be flapped out of the optical path of the X-rays 28, for example with a motor 40 controlled by the controller 26. However, it also may be possible that trenches of G0 are filled with liquid metal that may be emptied. According to an embodiment of the invention, the imaging device 10 further comprises a hinge 48 for removing the source grating G0 from the X-ray beam. The other two gratings G1 and G2 may be moved linearly by a (further) motor 40 controlled by the controller 26. According to an embodiment of the invention, the device 10 comprises a motor 40 for moving at least one of the gratings G1, G2 between the first position and the second position, wherein a controller 26 controls the movement. The motion may be implemented by a stepper motor, a belt-drive, or similar, provided the “on” end-point is mechanically robust and well-defined. Since the “in-line” setting requires a very accurate alignment of the gratings G1, G2, the distance of mechanical movements may have to be limited in order to preserve accuracy. Similarly, G2 may have to be moved out or in within seconds, in order to not disrupt the workflow. The positioning of the gratings G1 and G2 particularly in the first “on”-position (dPCI mode) usually must be well-defined as alignment of the gratings G1, G2 when employing dPCI is critical. Therefore, means may be required for ensuring a high mechanical precision in the “on”-position. The device 10 may comprise a two-state switching mechanism 40, 44, with a geometrically very well-defined, lockable “on”-position and a geometrically more laxly defined “off”-position. The former would take the much higher precision into account with which the gratings G2 would have to be placed relative to G1 once phase contrast is required. The required accuracy may be ensured for example by a rigid mechanical stopper, or by an optical detection of the position of grating G2 at the “on”-position with a sensor 46. The accuracy needs to be sufficient to eliminate the need for a new calibration. According to an embodiment of the invention, the first position of the grating G1, G2 is determined by a mechanical stopper 44. According to an embodiment of the invention, the first position of the grating G1, G2 is determined by a position sensor 46. If G1 stays fixed, and only G2 is mobile, the accurate locked position and/or motion end point for the dPCI mode may be ensured by measuring the Moiré pattern, i.e. an air image, while shifting G2 back, until it is again in the calibrated position and then stopping the motor. Teflon rails 42 or similar may be provided as guides for the gratings G1, G2 in the other spatial directions vertical and other in-plane axis, without creating a risk of dust/residue that would accumulate on the detector or on the gratings G1, G2. According to an embodiment of the invention, the imaging device 10 further comprises Teflon rails 42 for guiding the grating. Usually, only the absorber grating G2 substantially influences the effective detective quantum efficiency of the device 10, i.e. the absorption of X-rays. The source grating G0, which may be located at the exit window of the X-ray source 20 and hence in front of the object, 14 influences only the available X-ray flux, not the dose. The phase grating G1, despite being behind the breast exerts only a small influence on the dose, as G1 is usually by design a phase-grating with small attenuation of only about 5-10%. Hence, switching from phase contrast to conventional imaging may be realized by removing only the absorber grating G2 from the setup, i.e. from the post-object X-ray imaging path. According to an embodiment of the invention, only the absorber grating G2 has the transparent areas 38 and/or may be moved between the first position and the second position. However, it is also possible to provide the phase grating G1 with transparent areas and to shift it for moving the transparent areas 38 in the optical path of the fan-shaped X-ray beams 38. According to an embodiment of the invention, the phase grating G1 and the absorber grating G2 have the transparent areas 38 and/or may be moved between the first position and the second position. As shown in FIG. 1, both gratings are moved by a respective motor 40 on a respective rail 42, i.e. they are movable independently from each other. According to an embodiment of the invention, the phase grating G1 and the absorber grating G2 are movable independently from each other between the first position and the second position. There may be a first motor 40 for moving the phase grating G1 and a second motor for moving the absorber grating G2. FIG. 5 shows further possible embodiments, how the gratings G0, G1 and G2 may be moved. In FIG. 5, the gratings G1 and G2 are connected with each other, for example with a common frame and cannot move relative with each other, which may enhance the accurate alignment of the two gratings G1, G2. According to an embodiment of the invention, the phase grating G1 and the absorber grating G2 are fixedly connected with each other and are movable together between the first position and the second position. The whole interferometer for dPCI, which comprises the two gratings G1, G2 may be retractable. The movement may be performed with a common motor 40 controlled by the controller 26. Furthermore, FIG. 5 shows that the grating G0 (when movable) may be moved like the gratings G1 and G2 shown in FIG. 5. In particular, the grating G0 may comprise groups 36 of grating lines 34 and may be laterally shifted by a motor 40 on a rail 42. According to an embodiment of the invention, the source grating G0 comprises groups 36 of grating lines 34 and a transparent area 38 between the groups of grating lines and is movable with respect to the gantry 12, such that in a first position of the source grating G0, X-ray beams from the groups of grating lines pass through slits 30 in the collimator 22, and in a second position of the source grating G0, X-ray beams from transparent areas in the source grating pass through the slits 30 in the collimator 22. According to an embodiment of the invention, the device 10 comprises a motor for moving the source grating G0 and/or a Teflon rail 42 for guiding the source grating G0. The movement of the grating lines of the source grating G0 out of the optical path may also enhance the speed of acquisition. When the source grating is not attenuating the X-ray beams, there may be 3 to 4 times more X-ray flux and the scan time may be much shorter for the same image quality. FIG. 6 shows a method for acquiring differential phase image data and attenuation image data with the device 10 shown in FIG. 1 or 5. In the beginning, it is assumed that the device 10 and the gratings G0, G1, G2 are in the “off”-position, i.e. the second position, in which the X-ray beams 28 may pass through areas 38 in one or more of the gratings G0, G1, G2. Furthermore, grating G0 may be flapped out of the optical path of the X-ray beams 28 by the hinge 48 (see FIG. 1). In step 60, it is decided that dPCI data should be acquired. For example, a clinician using the device 10 may input a corresponding command into the controller 26. The controller 26 moves the gratings G0, G1, G2 into the “on”-position (first position), when they are movable, for example by a lateral shift of the grating by a distance corresponding to a distance less than half the distance between two detector lines 32. As the distance between two adjacent slits 30 of the collimator 22 is typically many multiples of a detector width of a single pixel in a device 10 with one pixel per line-width, it is easy to fabricate gratings G0 and/or G2 with Au filled silicon trenches only where needed while leaving adjacent areas metal-free. For example, the gratings (or one or two of the gratings) may be moved by a distance equal to the distance between a group 36 of grating lines 34 and a neighboring transparent area 38. According to an embodiment of the invention, at least one grating, two of the gratings or all of the gratings G0, G1, G2 of the device 10 are moved in a first position; such that fan-shaped X-ray beams 28 generated by a collimator 22 can pass through groups 36 of grating lines 34 on the grating. The correct positioning of the gratings G0, G1, G2 after the movement may be controlled by mechanical stoppers 44 and/or by a sensor 46, which data is evaluated by the controller 26. Furthermore, the gratings G1, G2 may be moved independently from each other (FIG. 1) or may be moved with a common frame (FIG. 2). In step 62, differential phase image data is acquired by moving the gantry 12 with the phase grating G1 and absorber grating G2 and the line detector 24 with respect to the object 14 and by detecting X-rays passing through the object 14, the phase grating G1 and the absorber grating G2 at a plurality of positions of the gantry 12. The controller 26 may control the movement of the gantry 12 and furthermore may evaluate the acquired data from the line detectors 24 to generate the differential phase contrast image data. In step 62, it is decided that pure attenuation image data should be acquired with the device 10. Also this command may be input by a clinician into the controller 26. The gratings G0, G1, G2 (if movable) are moved back into the “off”-position, i.e. the second position. According to an embodiment of the invention, at least one of the gratings G0, G1, G2 are moved in the second position, such that the fan-shaped X-ray beams 28 can pass through transparent areas 38 on the grating G0, G1, G2. In step 64, attenuation image data is acquired by moving the gantry 12 with respect to the object 14 and by detecting X-rays passing through the object 14 at a plurality of positions of the gantry 12. The controller 26 may control the movement of the gantry 12 and furthermore may evaluate the acquired data from the line detectors 24 to generate the attenuation image data. In addition, the device 10 allows the calibration of the dPCI interferometer G1, G2 to be performed in two separate steps: The common calibration of the image detector (e.g. for dark current, sensitivity) without the absorbing “analyzer” grating(s) G1 and G2, in which the source grating is preferable in the second “off” position, and the subsequent additional calibration of the dPCI interferometer G1, G2 with both gratings G1 and G2 in the first “on” position and the source grating G0 is in the second “off” position. A system with the interferometer locking into the imaging path requires a much more complex, time-consuming, and potentially less accurate combined calibration. In step 66, the gratings G1 and G2 and optionally the source grating G0 are moved in the second “off” position and the line detector 24 is calibrated. In step 68, the gratings G1 and G2 are moved in the first “on” position and the source grating G0 is moved in second “off” position. After that the interferometer comprising the phase grating G1 and the absorber grating G2 is calibrated. An imaging device 10 able for both differential phase contrast imaging and attenuation imaging may have many advantages. On the one hand, differential phase contrast mammography (dPCM) is currently being explored very intensively in view of its potential benefits like increased visibility and sharpness of micro-calcifications, better tumor delineation versus healthy tissue and increased general image sharpness and quality. At the current state of research, though, it is yet unclear, whether the improved image quality will allow a dose reduction that compensates for the X-ray quanta that are lost in the absorbing part of the interferometer behind the patient. On the other hand, depending on the clinical application (screening with manual reading or with computer-assisted-diagnosis, diagnostic workup, image-guided biopsy, etc.), clinical protocols and local guidelines will require part of the mammography workload to be carried out in “conventional” setup without the post-patient interferometer (i.e. gratings G1 and G2), while other parts will benefit from the “dPCI” setup with the interferometer. While immediately obvious for the clinical approval studies themselves and the initial years after introduction of dPCI systems to the market, the need for inserting/removing the phase contrast interferometer on a scan-by-scan basis is expected to remain forever. Furthermore, in clinical settings with a wide mix e.g. of dense vs. fatty breast or Caucasian vs. African-American vs. Asian women, a clinician will require the ability to choose on a per-patient or even per-view basis. Finally, market uptake will be faster if a clinician is not forced to choose a priori between a conventional system without dPCI or a system that can only perform dPCI, but can buy a conventional system with dPCI as an option that he can turn on or off per his clinical judgment. As pointed out above and depending on the nature of the mammographic scan being acquired (screening or diagnostic or interventional), clinical protocols and guidelines and clinician preferences on a per-scan basis will usually require the ability to switch between conventional and phase contrast imaging, as is possible with the imaging device 10 as described above. While the invention has been illustrated and described in detail in the drawings and foregoing description, such illustration and description are to be considered illustrative or exemplary and not restrictive; the invention is not limited to the disclosed embodiments. Other variations to the disclosed embodiments can be understood and effected by those skilled in the art and practising the claimed invention, from a study of the drawings, the disclosure, and the appended claims. In the claims, the word “comprising” does not exclude other elements or steps, and the indefinite article “a” or “an” does not exclude a plurality. A single processor or controller or other unit may fulfil the functions of several items recited in the claims. The mere fact that certain measures are recited in mutually different dependent claims does not indicate that a combination of these measures cannot be used to advantage. Any reference signs in the claims should not be construed as limiting the scope. G0 source grating G1 phase grating G2 absorber grating 10 X-ray imaging device 12 gantry 14 object 16 frame 18 support platform 20 X-ray source 22 collimator 24 line detector 26 controller 28 fan-shaped beam 30 slit 32 detector line 34 grating line 36 group of grating lines 38 transparent areas 40 motor 42 rail 44 stopper 46 position sensor 48 hinge |
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description | This is a U.S. National Phase Application under 35 USC 371 of International Application PCT/JP2016/081222 filed on Oct. 21 , 2016. This application claims the priority of Japanese application no. 2016-043490 filed Mar. 7, 2016, the entire content of which is hereby incorporated by reference. The present invention relates to a method of manufacturing a laminated scintillator panel used in a Talbot system. Currently, in an X-ray image diagnosis, an absorption image in which X-ray attenuation is imaged after an X-ray has passed through an object is used. On the other hand, an X-ray is a kind of electromagnetic waves, and attention is paid to the wave nature of X-rays, and attempts to image phase change of the X-rays after passing through an object have been recently made. These are called absorption contrast and phase contrast, respectively. The imaging technique using this phase contrast is considered to be more sensitive to soft tissues of a human body, which contains a lot of light elements, because the technique has higher sensitivity to light elements than a conventional absorption contrast. However, since conventional phase contrast imaging techniques have required use of a synchrotron X-ray source or a minute focus X-ray tube, it has been thought that practical use in general medical facilities is difficult because the former requires a huge facility and the latter is unable to secure sufficient X-ray dose to photograph a human body. In order to solve this problem, an X-ray image diagnosis (Talbot system) using an X-ray Talbot-Lau interferometer capable of acquiring a phase contrast image using an X-ray source used in medical practice has been conventionally expected. In the Talbot-Lau interferometer, as shown in FIG. 3, a G0 lattice, a G1 lattice, and a G2 lattice are arranged between a medical X-ray tube and an FPD, respectively, and refraction of X-rays by a subject is visualized as moire fringes. X-rays are irradiated in the longitudinal direction from the X-ray source arranged in an upper portion, and reach an image detector through G0, a subject, G1, and G2. As a method of manufacturing a lattice, for example, a method in which a silicon wafer having high X-ray transparency is etched to provide lattice-shaped recesses and a heavy metal having high X-ray shielding properties is filled therein is known. However, with the above-described method, it is difficult to increase the area due to the size of an available silicon wafer, restrictions on an etching apparatus, or the like, and an object to be photographed is limited to a small part. It is not easy to form a deep recess in a silicon wafer by etching, and it is also difficult to evenly fill a metal up to the depth of the recess, and therefore, it is difficult to fabricate a lattice having a thickness enough to sufficiently shield X-rays. For this reason, particularly under high-voltage photographing conditions, X-rays pass through such a lattice, resulting in failure to obtain a favorable image. On the other hand, it is also considered to adopt a lattice-shaped scintillator having a lattice function added to a scintillator constituting an image detector. For example, “Structured scintillator for x-ray grating interferometry” (Paul Scherrer Institute (PSI)), Applied Physics Letter 98, 171107 (2011) discloses a lattice-shaped scintillator in which a groove of a lattice fabricated by etching a silicon wafer is filled with a phosphor (CsI). However, in the above method, since a silicon wafer is used as in the above-described method of manufacturing a G2 lattice, problems caused by a silicon wafer such as constraints of the area of the wafer and difficulty in thickening the wafer are not improved. Furthermore, a new problem that emission of CsI attenuates due to repeated collisions on a wall of a silicon lattice, whereby the luminance decreases has been brought about. Non Patent Document 1: Applied Physics Letter 98, 171107 (2011) The present invention relates to a method of manufacturing a lattice-shaped laminated scintillator panel capable of enlarging the area and increasing the thickness with a means completely different from a conventional technique using a silicon wafer. In order to realize at least one of the above-described objects, a method of manufacturing a laminated scintillator panel reflecting one aspect of the present invention includes the following. A method of manufacturing a laminated scintillator panel having a structure in which a scintillator layer and a non-scintillator layer are repeatedly laminated in a direction substantially parallel to the direction of radiation incidence, the method including: a step of forming a laminate by repeatedly laminating the scintillator layer and the non-scintillator layer; and a joining step of pressurizing the laminate to join the scintillator layer and the non-scintillator layer integrally. According to the present invention, a lattice-shaped laminated scintillator panel can be provided by a simple method in which a scintillator layer and a non-scintillator layer are laminated and joined, instead of performing operations such as etching of a silicon wafer and filling of a phosphor into a groove. According to such a method, it is possible to increase the area and thickness, and the lattice pitch is arbitrarily adjustable. The laminated scintillator panel according to the present invention can be used as a scintillator having a function of a lattice for a Talbot-Lau interferometer. The laminated scintillator of the present invention has high luminance and is suitable for large area and thick film. This enables high-pressure imaging as well, enabling photography of thick subjects such as thoracoabdominal parts, thighs, elbows, knees, or hip joints. Conventionally, in diagnostic imaging of cartilage, MRI is the mainstream, there are drawbacks that the photographing cost is high and the photographing time is long because large-scale equipment is used. On the other hand, according to the present invention, it is possible to photograph soft tissue such as cartilage, muscle tendon, ligament, and visceral tissue with a faster x-ray image at lower cost. Therefore, it can be widely applied to orthopedic diseases such as rheumatoid arthritis or knee osteoarthritis and image diagnosis of soft tissue including breast cancer. A method of manufacturing the laminated scintillator panel of the present invention will be described. As shown in FIG. 2, a laminated scintillator panel has a structure in which a scintillator layer and a non-scintillator layer are repeatedly laminated in a direction substantially parallel to the direction of radiation incidence. By facing the radiation incidence surface or the opposite surface thereof the laminated scintillator panel to a photoelectric conversion panel, it is possible to convert an emission of the scintillator by radiation into an electric signal to acquire a digital image. Substantially parallel is almost parallel, and perfect parallel and some inclination also fall within the category of substantially parallel. The thickness (hereinafter referred to as lamination pitch) of a pair of a scintillator layer and a non-scintillator layer in the lamination direction and the ratio (hereinafter duty ratio) of the thickness of the scintillator layer to the thickness of the non-scintillator layer in the lamination direction are derived from Talbot interference conditions, and in general, the lamination pitch is from 0.5 to 50 μm and the duty ratio is preferably from 30/70 to 70/30. In order to obtain a diagnostic image with a sufficient area, it is preferable that the number of repeated lamination layers of the lamination pitch is from 1,000 to 500,000. The thickness of the laminated scintillator panel of the present invention in the radiation incidence direction is preferably from 10 to 1,000 μm, and more preferably from 100 to 500 μm. When the thickness in the radiation incidence direction is smaller than the lower limit value of the above range, the light emission intensity of the scintillator is weakened, and the image quality is deteriorated. When the thickness in the radiation incidence direction is larger than the upper limit of the above range, the distance of light emitted from the scintillator to a photoelectric conversion panel becomes long, and therefore, light easily diffuses and the sharpness deteriorates. The scintillator layer in the present invention is a layer containing a scintillator as a main component, and preferably contains scintillator particles. As the scintillator according to the present invention, substances capable of converting radiation such as X rays into radiation having different wavelengths such as visible light can be appropriately used. Specifically, scintillators and phosphors described in “Phosphor Handbook” (edited by Phosphor Research Society, Ohmsha Ltd., 1987) ranging from page 284 to page 299, substances listed in the web site “Scintillation Properties (http://scintillator.lbl.gov/)” of the US Lawrence Berkeley National Laboratory, or the like may be used, and substances not mentioned here can also be used as scintillators as long as they are “substances capable of converting radiation such as X-rays into radiation having different wavelengths such as visible light”. Specific examples of the composition of the scintillator include the following examples. First, a metal halide phosphor represented by Basic composition formula (I): MIX·aMIIX′2·bMIIIX″3: zA can be mentioned. In the basic composition formula (I), MI represents at least one element selected from the group consisting of elements capable of becoming monovalent cations, such as lithium (Li), sodium (Na), potassium (K), rubidium (Rb), cesium (Cs), thallium (Tl), and silver (Ag). MII represents at least one element selected from the group consisting of elements capable of becoming divalent cations, such as beryllium (Be), magnesium (Mg), calcium (Ca), strontium (Sr), barium (Ba), nickel (Ni), copper (Cu), zinc (Zn), and cadmium (Cd). MIII represents at least one element selected from the group consisting of scandium (Sc), yttrium (Y), aluminum (Al), gallium (Ga), indium (In), and elements belonging to the lanthanoid. X, X′, and X″ each represent a halogen element, and may be different or the same. 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 each independently represent a numerical value within the range of 0≤a<0.5, 0≤b<0.5, 0<z<1.0. Rare earth activated metal fluorohalide phosphors represented by Basic composition formula (II): MIIFX: zLn can also be mentioned. In the basic composition formula (II), MII represents at least one alkaline earth metal element, Ln represents at least one element belonging to the lanthanoid, and X represents at least one halogen element. z satisfies 0<z≤0.2. Rare earth oxysulfide phosphors represented by Basic composition formula (III): Ln2O2S: zA can also be mentioned. In the basic composition formula (III), Ln represents at least one element belonging to the 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 satisfies 0<z<1. In particular, Gd2O2S using gadolinium (Gd) as Ln is preferable since it is known to exhibit high emission characteristics in a wavelength region where the sensor panel is most likely to receive light by using terbium (Tb), dysprosium (Dy) or the like as the element type of A. Metal sulfide-based phosphors represented by Basic composition formula (IV): MIIS: zA can also be mentioned. In the basic composition formula (IV), MII represents at least one element selected from the group consisting of elements capable of becoming divalent cations, such as alkaline earth metals, Zn (zinc), Sr (strontium), and Ga (gallium), 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 satisfies 0<z<1. Metal oxoacid salt-based phosphors represented by Basic composition formula (V): MIIa(AG)b: zA can also be mentioned. In the basic composition formula (V), MII represents a metal element which can be a cation, (AG) represents at least one oxo acid group selected from the group consisting of phosphate, borate, silicate, sulfate, tungstate and 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 all possible values depending on the valence of the metal and oxo acid groups. z satisfies 0<z<1. A metal oxide-based phosphor represented by Basic composition formula (VI): MaOb: zA can be mentioned. In the 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 all possible values depending on the valence of the metal and oxo acid groups. z satisfies 0<z<1. Besides, a metal acid halide-based phosphor represented by Basic composition formula (VII): LnOX: zA can be mentioned. In the basic composition formula (VII), Ln represents at least one element belonging to the 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 satisfies 0<z<1. The average particle size of the scintillator particles is selected according to the thickness of a scintillator layer in the lamination direction, and is preferably 100% or less, more preferably 90% or less, with respect to the thickness in the lamination direction of the scintillator layer. When the average particle diameter of the scintillator particles exceeds the above range, disturbance of the lamination pitch becomes large and the Talbot interference function decreases. The content of the scintillator particles in the scintillator layer is preferably 30 vol % or more, more preferably 50 vol % or more, still more preferably 70 vol % or more in consideration of luminescent properties. The non-scintillator layer in the present invention does not contain a scintillator as a main component, and the content of the scintillator in the non-scintillator layer is less than 10 vol %, preferably less than 1 vol %, and most preferably 0 vol %. Preferably, the non-scintillator layer contains a variety of glasses, polymer materials, metals, and the like as main components. These may be used singly or in combination of a plurality of them. Specifically, plate glass such as quartz, borosilicate glass, or chemically tempered glass; ceramics such as sapphire, silicon nitride, or silicon carbide; semiconductor such as silicon, germanium, gallium arsenide, gallium phosphide, or gallium nitride; polymers such as: polyester such as polyethylene terephthalate (PET) or polyethylene naphthalate (PEN); aliphatic polyamide such as nylon; aromatic polyamide (aramid); polyimide; polyamide imide; polyether imide; polyethylene; polypropylene; polycarbonate; triacetate; cellulose acetate; epoxy; bismaleimide; polylactic acid; sulfur-containing polymers such as polyphenylene sulfide or polyether sulfone; polyether ether ketone; fluororesin; acrylic resin; or polyurethane; carbon fibers, glass fibers or the like (in particular, fiber reinforced resin sheets including such fibers); metal foil such as aluminum, iron, copper; bionanofibers including chitosan, cellulose, or the like can be used. As a non-scintillator layer, a polymer film is preferably used as a main component from the viewpoint of manufacturing handling, and a material transparent to an emission wavelength of a scintillator is particularly preferable. By making the non-scintillator layer transparent, the light emission of the scintillator propagates not only within the scintillator layer but also into the non-scintillator layer, and the amount of light reaching a sensor increases and the luminance improves. The transmittance of the non-scintillator layer is preferably 80% or more. The present invention includes the steps of: laminating a scintillator layer and a non-scintillator layer; and joining the scintillator layer and the non-scintillator layer. Joining in the present invention means bonding the scintillator layer and the non-scintillator layer to integrate them. As a joining method, both of them can be adhered via an adhesive layer, and from a viewpoint of process simplification, it is preferable to preliminarily contain an adhesive resin in the scintillator layer or the non-scintillator layer, and bring them into close contact with each other by pressure, thereby joining them without interposing an adhesive layer. Heating in a pressurized state is more preferable because a substance having adhesiveness is melted or cured to strengthen adhesion. It is also possible to coat the surface of the non-scintillator layer with a composition capable of forming a scintillator layer, or to join the scintillator layer and the non-scintillator layer by further removing the solvent, as needed. An adhesive resin may be contained in either a scintillator layer or a non-scintillator layer, and particularly preferably, a scintillator layer contains an adhesive resin as a binder for scintillator particles. The adhesive resin is preferably a material that is transparent to the emission wavelength of the scintillator so as not to inhibit the propagation of light emitted from the scintillator. The adhesive resin is not particularly limited as long as the object of the present invention is not impaired, and examples thereof include natural polymers such as proteins such as gelatin, polysaccharides such as dextran, or gum arabic; and synthetic polymeric substances such as polyvinyl butyral, polyvinyl acetate, nitrocellulose, ethylcellulose, vinylidene chloride·vinyl chloride copolymer, poly(meth)acrylate, vinyl chloride·vinyl acetate copolymer, polyurethane, cellulose acetate butyrate, polyvinyl alcohol, polyester, epoxy resin, polyolefin resin, and polyamide resin. These resins may be crosslinked with a crosslinking agent such as epoxy or isocyanate, and these adhesive resins may be used singly or in combination of two or more kinds. The adhesive resin may be either a thermoplastic resin or a thermosetting resin. The content of an adhesive resin contained in the scintillator layer is preferably from 1 to 70 vol %, more preferably from 5 to 50 vol %, and still more preferably from 10 to 30 vol %. When the content is lower than the lower limit of the above range, sufficient adhesiveness is not obtained, and conversely, when the content is higher than the upper limit of the above range, the content of the scintillator becomes insufficient and the amount of luminescence decreases. The scintillator layer may be formed by coating a composition in which the scintillator particles and an adhesive resin are dissolved or dispersed in a solvent, or may be formed by coating a composition prepared by heating and melting a mixture containing the scintillator particles and an adhesive resin. When coating the composition in which the scintillator particles and the adhesive resin are dissolved or dispersed in a solvent, examples of usable solvents include: lower alcohols such as methanol, ethanol, isopropanol, and n-butanol; ketones such as acetone, methyl ethyl ketone, methyl isobutyl ketone, and cyclohexanone; esters of lower fatty acids and lower alcohols such as methyl acetate, ethyl acetate, and n-butyl acetate; ethers such as dioxane, ethylene glycol monoethyl ether, ethylene glycol monomethyl ether; aromatic compounds such as toluol and xylol; halogenated hydrocarbons such as methylene chloride and ethylene chloride; and mixtures thereof. A variety of additives such as a dispersant for improving dispersibility of scintillator particles in the composition and a curing agent or a plasticizer for improving the bonding force between an adhesive resin and scintillator particles in a scintillator layer after formation may be mixed in the composition. Examples of the dispersant used for such purpose include phthalic acid, stearic acid, caproic acid, and lipophilic surfactant. Examples of the plasticizer include: phosphoric acid esters such as triphenyl phosphate, tricresyl phosphate, and diphenyl phosphate; phthalic acid esters such as diethyl phthalate and dimethoxyethyl phthalate; glycolic acid esters such as ethyl phthalyl ethyl glycolate and butyl phthalyl butyl glycolate; and polyesters of polyethylene glycol and aliphatic dibasic acids such as polyesters of triethylene glycol and adipic acid, and polyesters of diethylene glycol and succinic acid. As the curing agent, a known curing agent for a thermosetting resin can be used. When heating and melting the mixture containing the scintillator particles and the adhesive resin, it is preferable to use a hot-melt resin as the adhesive resin. As the hot-melt resin, for example, one mainly composed of a polyolefin-based, polyamide-based, polyester-based, polyurethane-based, or acrylic-based resin can be used. Among these, from viewpoints of light permeability, moisture resistance, and adhesiveness, those based on a polyolefin resin as a main component are preferable. As the polyolefin-based resin, for example, ethylene-vinyl acetate copolymer (EVA), ethylene-acrylic acid copolymer (EAA), an ethylene-acrylic acid ester copolymer (EMA), ethylene-methacrylic acid copolymer (EMAA), ethylene-methacrylic acid ester copolymer (EMMA), an ionomer resin or the like can be used. These resins may be used as a so-called polymer blend in which two or more kinds of resins are combined. There are no particular restrictions on means for coating a composition for forming a scintillator layer, and usual coating means such as a doctor blade, a roll coater, a knife coater, an extrusion coater, a die coater, a gravure coater, a lip coater, a capillary coater, or a bar coater can be used. In the present invention, there is a step of bonding the scintillator layer and the non-scintillator layer after repeatedly laminating the scintillator layer and the non-scintillator layer. There are no particular restrictions on the method of repeatedly laminating the scintillator layer and the non-scintillator layer, and an individually formed scintillator layer and non-scintillator layer may be divided into a plurality of sheets, and then the sheets may be alternately repeatedly laminated. In the present invention, it is preferable that a plurality of partial laminates in which the scintillator layer and the non-scintillator layer are bonded to each other are formed, and then the plurality of partial laminates are laminated to form the laminate since it is easy to adjust the number of layers and the thickness of the laminate. For example, a partial laminate composed of a pair of scintillator layer and non-scintillator layer may be formed in advance, the partial laminate may be divided into a plurality of sheets, and the sheets may be laminated repeatedly. When the partial laminate composed of the scintillator layer and the non-scintillator layer has a film shape that can be wound up, efficient lamination is possible by winding the film on a core. The winding core may be cylindrical or a plate. More efficiently, the repeated laminate of the scintillator layer and the non-scintillator layer fabricated by the above method may be bonded (integrated) by pressurization, heating, or the like, and then divided into a plurality of sheets, and the sheets may be repeatedly laminated. There is no particular restriction on the method of forming a partial laminate composed of a scintillator layer and a non-scintillator layer, and a scintillator layer may be formed by selecting a polymer film as a non-scintillator layer and coating a composition containing scintillator particles and an adhesive resin on one side thereof. A composition containing scintillator particles and an adhesive resin may be coated on both sides of a polymer film. As described above, when a partial laminate is formed by coating a composition containing scintillator particles and an adhesive resin on a polymer film, it is possible to simplify a process and to easily divide the partial laminate into a plurality of sheets. The dividing method is not particularly limited, and a usual cutting method is selected. A transfer substrate coated with a scintillator layer in advance may be transferred onto a film composed of a non-scintillator layer. As needed, the transfer substrate is removed by means such as peeling. In the present invention, the scintillator layer and the non-scintillator layer are integrally bonded by pressurizing the laminate in such a manner that the scintillator layer and the non-scintillator layer are in a substantially parallel direction to the direction of radiation incidence. By heating a repeated laminate of a plurality of scintillator layers and non-scintillator layers in a pressurized state so as to obtain a desired size, the lamination pitch can be adjusted to a desired value. There is no particular restriction on the method of pressurizing the repeated laminate of the plurality of scintillator layers and the non-scintillator layer to have a desired size, and it is preferable to apply pressure in a state in which a spacer such as a metal is provided in advance so that the laminate is not compressed less than a desired size. The pressure at that time is preferably from 1 MPa to 10 GPa. When the pressure is lower than the lower limit of the above range, there is a possibility that a resin component contained in the laminate may be not deformed to a predetermined size. When the pressure is higher than the upper limit of the above range, a spacer may be deformed, and the laminate may be compressed less than a desired size. By heating the laminate in a pressurized state, bonding can be made more robust. Depending on the kind of a resin, it is preferable to heat a repeated laminate of a plurality of scintillator layers and non-scintillator layers for about from 0.5 to 24 hours at a temperature equal to or higher than the glass transition point for a thermoplastic resin and at a temperature equal to or higher than the curing temperature for a thermosetting resin. The heating temperature is preferably from 40° C. to 250° C. in general. When the temperature is lower than the lower limit of the above range, the fusion or curing reaction of the resin may be insufficient, and there is a possibility of poor bonding or returning to the original size when releasing compression. When the temperature is higher than the upper limit of the above range, there is a possibility that the resin deteriorates and the optical characteristics are impaired. There are no particular restrictions on the method of heating the laminate under pressure, and a press equipped with a heating element may be used, the laminate may be oven-heated in a state of being enclosed in a box-shaped jig so as to have a predetermined size, or a heating element may be mounted on a box-shaped jig. As a state before a repeated laminate of a plurality of scintillator layers and non-scintillator layers is pressurized, it is preferable that voids exist inside the scintillator layer, inside the non-scintillator layer, or in the interface between the scintillator layer and the non-scintillator layer. When pressure is applied in the absence of any voids, a part of a constituent material flows out from an end face of the laminate to cause disorder in the lamination pitch or return to the original size when releasing the pressure. When a void exists, the void functions as a cushion even when pressurized, and the laminate can be adjusted to an arbitrary size in the range until the void becomes zero, and in other words, the lamination pitch can be adjusted to an arbitrary value. The porosity is calculated from the following formula using a measured volume (area×thickness) of the laminate and the theoretical volume (weight÷density) of the laminate.(measured volume of laminate−theoretical volume of laminate)÷theoretical volume of laminate×100 When the area of the laminate is constant, the porosity is calculated from the measured thickness of the laminate and the theoretical thickness (weight÷density÷area) of the laminate according to the following formula.(measured thickness of laminate−theoretical thickness of laminate)÷theoretical thickness of laminate×100 The porosity of the scintillator layer after pressurizing is preferably 30 vol % or less. When the porosity exceeds the above range, the packing ratio of the scintillator decreases and the luminance decreases. As means for providing voids in the scintillator layer or the non-scintillator layer, for example, bubbles may be contained in the layer in the process of manufacturing the scintillator layer or the non-scintillator layer, or hollow polymer particles may be added. On the other hand, even when irregularities are present on the surface of the scintillator layer or the non-scintillator layer, the same effect can be obtained since a void is formed at the contact interface between the scintillator layer and the non-scintillator layer. As means for providing irregularities on the surfaces of the scintillator layer or the non-scintillator layer, for example, an irregularity-forming treatment such as a blast treatment or an emboss treatment may be applied to the surface of the layer, or irregularities may be formed on the surface by incorporating a filler in the layer. When a scintillator layer is formed by coating a composition containing scintillator particles and an adhesive resin on a polymer film, irregularities are formed on the surface of the scintillator layer, and voids can be formed at the contact interface with the polymer film. The size of the irregularities can be arbitrarily adjusted by controlling the particle size and dispersibility of the filler. Since a radiation source emitting radiation such as X-rays is generally a point wave source, when individual scintillator layers and non-scintillator layers are formed completely in parallel, X-rays obliquely enter the peripheral region of a laminated scintillator. As a result, in the peripheral region, so-called vignetting, in which radiation is not sufficiently transmitted, occurs. Vignetting becomes a serious problem as the scintillator becomes larger in area. This problem can be improved by making individual scintillator layers and non-scintillator layers parallel to the radiation by bending the laminated scintillator panel or by forming the laminated scintillator panel to have an inclination structure. In order for the laminated scintillator panel to have an inclination structure, for example, in a step of pressurizing the repeated laminate of the plurality of scintillator layers and non-scintillator layers, the pressurizing direction is made oblique, thereby forming an inclination structure having a trapezoidal cross section as shown in FIG. 4. The inclination angle is the maximum at an end side of the laminated scintillator panel, and it becomes continuously close to parallel toward the center. The maximum inclination angle is determined by the size of the laminated scintillator panel or the distance between the laminated scintillator panel and the radiation source, and is usually from 0 to 10°. As a pressurizing method for forming the inclination structure, for example, a pressurizing jig having a predetermined inclination as shown in FIG. 4 is used. The inclination angle 0° means parallel, and the above range is included in the concept of “substantially parallel” in the specification of the present application. At the interface between the scintillator layer and the non-scintillator layer of the laminated scintillator panel, for the purpose of improving sharpness, a light-shielding layer for suppressing diffusion of light emitted from the scintillator may be provided. The light-shielding layer is not particularly limited as long as it has a function of suppressing propagation of light emitted from the scintillator, and may have, for example, a light-reflecting function, or may have a light-absorbing function. In order not to significantly impair the luminance, the light-shielding layer is preferably inserted at a ratio of one layer to a lamination pitch of from five layers to 500 layers, and more preferably is inserted at a ratio of one layer to a lamination pitch of from 10 layers to 100 layers. When the lamination pitch falls below the lower limit of a specified value, the sharpness improves but the luminance greatly decreases, and when the lamination pitch exceeds the upper limit, the effect of improving the sharpness is impaired. Means for providing the light-shielding layer is not particularly restricted, and dye or pigment, or ink containing metal nanoparticles may be applied, or a metal thin film may be provided by a gas phase method such as vapor deposition or sputtering. In the present invention, it is preferable that a plurality of scintillator layers and non-scintillator layers are bonded and then a bonding end face is planarized. In particular, scattering of a scintillator light at the bonding end face can be suppressed by planarizing the face on the radiation incidence side, the side opposite thereto, or both sides, thereby improving the sharpness. The planarizing method is not particularly limited, and energy such as ions, plasma, electron beam, or the like may be irradiated in addition to machining such as cutting, grinding, and polishing. In the case of machining, it is preferable to work in a direction parallel to a laminated structure so as not to damage the lamination structure of scintillator layers and non-scintillator layers. Since the thickness of the laminated scintillator panel in the present invention in the direction of incidence of radiation is as thin as several mm or less, in order to maintain the lamination structure, it is preferable that the surface on the radiation incidence side, the side opposite thereto, or both surfaces are bonded and held on a support. As the support, a variety of glasses, polymer materials, metals, or the like which can transmit radiation such as X-rays can be used, and examples thereof include: glass sheets such as quartz, borosilicate glass, and chemically tempered glass; ceramic substrates such as sapphire, silicon nitride, and silicon carbide; semiconductor substrates (photoelectric conversion panels) such as silicon, germanium, gallium arsenide, gallium phosphorus, and gallium nitrogen; polymer films (plastic films) such as cellulose acetate films, polyester films, polyethylene terephthalate films, polyamide films, polyimide films, triacetate films, and polycarbonate films; metal sheets such as aluminum sheets, iron sheets, and copper sheets; metal sheets having a coating layer of the metal oxide; carbon fiber reinforced resin (CFRP) sheets; and amorphous carbon sheets. The thickness of the support is preferably from 50 μm to 2,000 μm, and more preferably from 50 to 1,000 μm. A method of laminating a laminated scintillator panel and a support is not particularly specified, and for example, an adhesive, a double-sided tape, a hot-melt sheet, or the like can be used. After laminating the laminated scintillator panel and the support, the surface opposite to the bonding surface may be planarized. Between the laminated scintillator panel and the support may be provided a layer that reflects light emitted from the scintillator or a layer that absorbs light emitted from the scintillator depending on an intended use. The luminance is improved by providing a layer that reflects light emitted from the scintillator, and the sharpness is improved by providing a layer that absorbs light emitted from the scintillator. The support itself may have a function of reflecting or absorbing light emitted from the scintillator. By facing the laminated scintillator panel of the present invention to a photoelectric conversion panel, it is possible to convert light emitted from the scintillator caused by radiation into an electric signal to acquire a digital image. Although the laminated scintillator panel and the photoelectric conversion panel may be faced to each other in a non-contact manner, in order to reduce the optical loss at the interface between the laminated scintillator panel and the photoelectric conversion panel, it is preferable that they are bonded with a transparent material having a refractive index exceeding 1.0 (air). The bonding method of the laminated scintillator panel and the photoelectric conversion panel is not particularly specified, and for example, an adhesive, a double-sided tape, a hot-melt sheet or the like can be used. The facing laminated scintillator panel and the photoelectric conversion panel may be curved so as to prevent the aforementioned vignetting. In this case, the photoelectric conversion panel is preferably a flexible material. According to the present invention as described above, a laminated scintillator panel capable of enlarging the area and increasing the thickness with a concept totally different from the use of a silicon wafer can be provided. According to the present invention, it is possible to provide a laminated scintillator panel by a simple method of going through a process of joining a scintillator layer and a non-scintillator layer, it is also possible to enlarge the area of the panel or to increase the thickness of the panel, which has been conventionally difficult, and it is also possible to arbitrarily adjust the lamination pitch. Therefore, the laminated scintillator panel according to the present invention can be used as a scintillator for a Talbot system. By changing the scintillator particles of the present invention to high-X-ray-absorbing particles such as heavy metals, the present invention can also be applied to manufacturing methods of a variety of lattices for Talbot such as G0 lattice, G1 lattice, and G2 lattice. Hereinafter, the present invention will be described by way of Examples, but the present invention is not limited to such Examples in any manner. Gd2O2S: Tb particles having an average particle diameter of 2 μm and an ethylene-vinyl acetate based hot melt resin (Evaflex EV150, melting point=61° C., manufactured by Du Pont-Mitsui Polychemicals Co., Ltd.) were mixed in such a manner that the solid content ratio (volume fraction) was 75/25, thereby obtaining a composition for forming a scintillator layer. This composition was melted at 200° C. and coated on a PET film (non scintillator layer) having the theoretical film thickness of 3 μm (calculated from the weight) with a die coater in such a manner that the theoretical film thickness was 3 μm (calculated from the weight), thereby preparing a partial laminate composed of a scintillator layer and a non-scintillator layer. Thereafter, 20,000 sheets of the partial laminates cut into 120 mm×3 mm were laminated. The actual film thickness of this laminate was 140 mm. Since the theoretical film thickness of this laminate is 120 mm, the porosity was 17%. Subsequently, a pressure of 0.2 GPa was applied in parallel to the lamination surface using a metal jig in such a manner that the film thickness of the laminate was 120 mm and the laminate in this state was heated at 100° C. for 1 hour to prepare a laminated block (120 mm×120 mm×3 mm) composed of 20,000 layers of partial laminates. The porosity of the laminate after pressurization was 0%. One side (120 mm×120 mm surface) of the laminated block was flattened by lathe machining, then an epoxy adhesive was applied thereto, and the laminated block was bonded to a CFRP plate having a thickness of 0.5 mm Thereafter, the laminated scintillator panel (120 mm×120 mm×0.3 mm) was obtained by lathe cutting until the thickness of the laminated block became 0.3 mm. As a result of observing the surface (120 mm×120 mm surface) of the fabricated laminated scintillator panel with a microscope, it was confirmed that the partial laminate was precisely aligned with a lamination pitch of 6 μm. Gd2O2S: Tb particles having an average particle diameter of 2 μm and a polyester resin (VYLON 200 manufactured by Toyobo Co., Ltd., Tg=67° C.) were mixed in a MEK solvent in such a manner that the solid content ratio (volume fraction) was 75/25, thereby obtaining a composition for forming a scintillator layer. This composition was melted and coated on a PET film (non scintillator layer) having the theoretical film thickness of 3 μm (calculated from the weight) with a die coater in such a manner that the theoretical film thickness was 3 μm (calculated from the weight), thereby preparing a partial laminate composed of a scintillator layer and a non-scintillator layer. Thereafter, 20,000 sheets of the partial laminates cut into 120 mm×3 mm were laminated. The actual film thickness of this laminate was 160 mm. Since the theoretical film thickness of this laminate is 120 mm, the porosity was 33%. Subsequently, a pressure of 0.2 GPa was applied in parallel to the lamination surface using a metal jig in such a manner that the film thickness of the laminate was 120 mm and the laminate in this state was heated at 100° C. for 1 hour to prepare a laminated block (120 mm×120 mm×3 mm) composed of 20,000 layers of partial laminates. The porosity of the laminate after pressurization was 0%. The above-described laminated film was processed in a similar manner to Example 1 to obtain a laminated scintillator panel (120 mm×120 mm×0.3 mm). As a result of observing the surface (120 mm×120 mm surface) of the fabricated laminated scintillator panel with a microscope, it was confirmed that the partial laminate was precisely aligned with a lamination pitch of 6 μm. A laminated scintillator panel was fabricated in a similar manner to Example 2 except that 20,000 sheets of partial laminates were laminated and then the laminate was pressed in such a manner that the film thickness of the laminate was 140 mm. It was confirmed that the porosity of the present laminate after pressurization was 17%, and the partial laminates were accurately arranged with a lamination pitch of 7 μm. In Example 1, after laminating 20,000 sheets of partial laminates cut into 120 mm×3 mm, the laminate was pressed in such a manner that the laminate had an inclination structure having a trapezoidal cross section as shown in FIG. 4 using a metal inclination jig with an inclination angle of 2°, and the laminate in this state was heated at 100° C. for 1 hour to prepare a laminated block (120 mm×120 mm (average value because of a slope)×3 mm) composed of 20,000 layers of minimum lamination unit). The surface on the radiation incidence side (see FIG. 4) of the laminated block was flattened by lathe machining, then an epoxy adhesive was applied thereto, and the laminated block was bonded to a CFRP plate having a thickness of 0.5 mm Thereafter, the laminated scintillator panel (120 mm×120 mm×0.3 mm) was obtained by lathe cutting until the thickness of the laminated block became 0.3 mm. The above-described laminated scintillator panel was bonded to a photoelectric conversion panel with an optical double-sided tape (CS9861US manufactured by Nitto Denko Corporation) on the surface of the radiation emission side (see FIG. 4), and an X-ray image was acquired by X-ray irradiation under a condition of a tube voltage of 40 kV. It was confirmed that by adjusting the distance between the laminated scintillator panel and the X-ray source to 172 cm, the angle of incidence of the X-ray coincides with the angle of the laminated structure, and that uniform light emission (luminance) can be obtained in the panel surface. This indicates that vignetting was prevented by making the laminated scintillator panel an inclination structure. |
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abstract | A charged particle beam writing apparatus according to an embodiment, includes a first dose calculating unit configured to calculate a first dose map for each set of a proximity effect correction coefficient map and a base dose map of a beam; a dimension map creation unit configured to create a dimension map of a pattern by using the first dose map calculated for each set; an adder configured to add dimensions of all sets for each position of the dimension map by using the dimension map of each set; a set map creation unit configured to create a set of a proximity effect correction coefficient map and a base dose map by using an added dimension map after addition; and a second dose calculating unit configured to calculate a second dose map by using a created set of the proximity effect correction coefficient map and the base dose map. |
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