JP2010032558A - Reactor core of light-water reactor, and fuel assembly - Google Patents

Abstract

<P>PROBLEM TO BE SOLVED: To provide a reactor core of a light-water reactor and a fuel assembly capable of further improving the degree of combustion by observing limiting conditions for safety and increasing nuclear non-proliferation resistance properties, and capable of performing multiple-recycle and effectively utilizing a trans-uranium nuclide. <P>SOLUTION: The reactor core flow rate of BWR is adjusted to a set reactor core flow rate determined from characteristics shown in Fig.2 according to the ratio of Pu-239 in TRU included in a new fuel assembly loaded to the reactor core. The reactor core flow rate is held at a set reactor core flow rate in the reactor core through each operation cycle of BWR. As a result, the ratio of a plurality of isotopes of TRU included by the fuel assembly when the fuel assembly is taken out of the BWR reactor core as a spent fuel assembly becomes essentially the same as the ratio of the isotopes of TRU included by the fuel assembly in the state of the new fuel assembly loaded to the BWR reactor core. The ratio of Pu-239 in TRU included in the new fuel assembly is not less than 40% and not more than 50%. <P>COPYRIGHT: (C)2010,JPO&INPIT

Description

  The present invention relates to a light water reactor core and a fuel assembly, and more particularly to a light water reactor core and a fuel assembly suitable for application to a boiling water reactor core.

  Actinide nuclides with many isotopes, when burned in the core in the nuclear fuel material loaded in the core of a light water reactor, undergo nuclear transmutation such as fission and neutron absorption, Transition between isotopes sequentially. As actinide nuclides, there are so-called odd nuclei that have a large fission cross section for the resonance region and thermal neutrons, and so-called even nuclei that fission only for fast neutrons. The abundance of each isotope of the nuclide changes greatly. This change in the existence ratio is known to depend on the neutron energy spectrum at the loading position of the fuel assembly in the core.

  Current light water reactors use low enriched uranium as nuclear fuel. However, since natural uranium resources are finite, fuel assemblies used in light water reactors are converted into depleted uranium, which is a residue during uranium enrichment, and transuranium nuclides (hereinafter referred to as TRU) extracted from spent fuel assemblies in light water reactors. Need to be replaced with a recyclable fuel assembly that uses enriched nuclear fuel material. The TRU is recycled as a useful resource for a fairly long period of time when a commercial furnace is expected to be required, during which time the amount of TRU needs to increase or remain almost constant. Patent Document 1 describes a technology for realizing a breeder reactor in which the amount of fissile Pu is increased or substantially constant through the burning of nuclear fuel in a light water reactor that currently occupies most of commercial reactors. A light water reactor that realizes the breeder reactor described in Patent Document 1 and Non-Patent Document 1 includes a plurality of fuel assemblies in which a plurality of fuel rods are densely arranged in a triangular lattice and the cross section has a hexagonal shape. Arranged in the core. The core of this light water reactor is a dense arrangement of fuel rods that reduces the amount of water around the fuel rods and increases the proportion of resonance and fast energy neutrons, while lowering the height of the mixed oxide fuel section of the TRU and The blanket area of depleted uranium is placed in the area, and the negative void coefficient, which is a safety standard, is maintained. Further, the core is applied with the concept of a parfait-type core described in Non-Patent Document 2, and the growth ratio is set to 1 or more while ensuring economic efficiency by stacking the core portion in two stages.

  In order to recycle TRU, reprocessing of spent fuel is indispensable. However, the demand for nuclear non-proliferation has become increasingly severe due to the fear that civil TRUs will be diverted to weapons of mass destruction, and restrictions on TRU recycling are becoming stricter.

  Also, sometime in the future, there will always be a better power system to replace the nuclear fission reactor, at which time TRU will turn from a very useful nuclear fuel equivalent to enriched uranium to troublesome long-lived waste. It will go down. Therefore, preparing a TRU disposal method is the most important issue for nuclear development.

Japanese Patent No. 3428150

R. TAKEDA et al., Proc. Of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems.GLOBAL '95 Versailles, France, September, 1995, P.938 G. A. Ducat et al., "EVALUATION OF THE PARFAIT BLANKET CONCEPT FOR FAST BREEDER REACTORS", MITNE-157, January, 1974

  At present, there is a track record of burning only Pu among the TRUs contained in the spent fuel of the light water reactor, but it is considered impossible to recycle Pu and TRU. Considering that fast neutron fields are effective for TRU combustion, combining subcritical systems and large accelerators (ADS) Even in systems with positive reactivity coefficient, a method to ensure safety by stopping the beam from the accelerator Development is underway with two-branch system. However, this development remains in the TRU partial weight loss scenario.

  In the light water reactor that recycles TRUs described in Patent Document 1 and Published Document 1, the thermal limitations of the light water reactor include the maximum linear power density (MLHGR) that defines the center temperature of the fuel pellet and the burnout of the fuel rod cladding tube. There is a minimum critical power ratio (MCPR) to prevent, and the improvement of core performance has been hindered by the limitation of MCPR. In the process leading to the recycling age, TRUs with different isotope ratios are supplied from the core of the light water reactor loaded with uranium fuel. For this reason, various reactivity coefficients, which are important safety constraints, are deteriorated and the tolerance to the constraints is reduced. Therefore, there is a possibility that the recycling must be interrupted and multiple recycling cannot be realized. .

  Recently, interest in non-proliferation has increased worldwide, making it difficult to use TRUs that could be diverted to weapons of mass destruction. For this reason, there is a need for a system that can recycle TRUs that have a high proliferation resistance and a small proportion of Pu-239 among the TRUs to be recycled.

  If recycling is repeated when it is desired to extinguish TRU, only the fissionable odd-numbered nuclides disappear first, and the proportion of even-numbered nuclides that only fission with fast energy neutrons increases. As a result, the criticality cannot be maintained and the fission chain reaction cannot be continued, or the reactivity coefficient, which is a safety constraint, becomes positive, and the TRU disappearance work is abandoned halfway. There is a problem to be solved in realizing the multiple recycling, such as unavoidable.

  An object of the present invention is to increase nuclear non-proliferation resistance while maintaining safety constraints, to further increase the burnup, and to enable multiple recycling and to make effective use of transuranium nuclides. The object is to provide a core and a fuel assembly for a light water reactor.

  The feature of the present invention that can achieve the above-described object is that the ratio of Pu-239 in all of the above-mentioned transuranium nuclides contained in the fuel assembly of zero burnup loaded in the core is in the range of 40% to 45%. There is in.

  According to the present invention, it is possible to increase nuclear non-proliferation resistance while maintaining safety constraints, to further increase the burnup, and to enable multiple recycling and to effectively use transuranium nuclides. it can.

It is explanatory drawing which shows the production | generation decay chain of an actinide nuclide. It is a characteristic view which shows the relationship between a relative core flow volume and the ratio (existence ratio) of Pu239 in TRU. It is a characteristic view which shows the relationship of the water to fuel material ratio with respect to the abundance ratio of Pu239 in TRU. It is a characteristic view which shows the relationship between the abundance ratio of Pu239 in TRU, combustion efficiency, and a void coefficient. BRIEF DESCRIPTION OF THE DRAWINGS It is a block diagram of the light water reactor of Example 1 which is one suitable Example of this invention. FIG. 6 is a transverse sectional view of the core shown in FIG. 5. FIG. 7 is an explanatory view showing a fuel assembly lattice shown in FIG. 6. FIG. 7 is an explanatory diagram showing the arrangement of fuel assemblies in the equilibrium core state of the core shown in FIG. 6. It is explanatory drawing which shows the opening distribution of the orifice in the equilibrium core shown in FIG. FIG. 7 is an explanatory diagram showing the enrichment distribution of fissile Pu in the axial direction of a new fuel assembly loaded on the equilibrium core of the core shown in FIG. 6. It is explanatory drawing which showed arrangement | positioning of each fuel rod from which the enrichment level of fissile Pu in the cross section of the fuel assembly shown in FIG. 7 differs. It is explanatory drawing which shows the power distribution and void ratio distribution in the axial direction of the core shown in FIG. It is explanatory drawing which shows arrangement | positioning of the fuel assembly in the equilibrium core of the light water reactor of Example 2 which is another Example of this invention. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core of the core shown in FIG. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 3 which is another Example of this invention. FIG. 6 is an explanatory diagram showing the arrangement of fuel assemblies in the equilibrium core of a light water reactor according to a third embodiment. FIG. 17 is an explanatory diagram showing the enrichment distribution of fissile Pu in the axial direction of a new fuel assembly loaded on the equilibrium core of the core shown in FIG. 16. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 4 which is another Example of this invention. FIG. 10 is an explanatory diagram showing the arrangement of fuel assemblies in the equilibrium core of a light water reactor according to a fourth embodiment. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core of the core shown in FIG. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 5 which is another Example of this invention. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the core of the light water reactor of Example 5. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 6 which is another Example of this invention. FIG. 10 is an explanatory diagram showing the arrangement of fuel assemblies in an equilibrium core of a light water reactor according to a sixth embodiment. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core of the core shown in FIG. It is a transverse cross section of the core in the light water reactor of Example 7 which is other examples of the present invention. It is explanatory drawing which shows the fuel assembly lattice in the core shown in FIG. FIG. 27 is an explanatory diagram showing the arrangement of fuel assemblies in the equilibrium core state of the core shown in FIG. 26. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core shown in FIG. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 8 which is another Example of this invention. FIG. 10 is an explanatory diagram showing the arrangement of fuel assemblies in an equilibrium core of a light water reactor according to an eighth embodiment. FIG. 32 is an explanatory diagram showing an enrichment distribution of fissile Pu in the axial direction of a new fuel assembly loaded on the equilibrium core shown in FIG. 31. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 9 which is another Example of this invention. FIG. 10 is an explanatory diagram showing the arrangement of fuel assemblies in an equilibrium core of a light water reactor according to a ninth embodiment. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core shown in FIG. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 10 which is another Example of this invention. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded in the core of the light water reactor of Example 10. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 11 which is another Example of this invention. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded in the core of the light water reactor of Example 11. It is explanatory drawing which shows the fuel assembly grating | lattice in the light water reactor of Example 12 which is another Example of this invention. FIG. 20 is an explanatory diagram showing the enrichment distribution of fissile Pu in the axial direction of a new fuel assembly loaded on the core of a light water reactor according to Example 12. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded into the core of the light water reactor of Example 13 which is another embodiment of the present invention.

  The Na-cooled fast reactor aiming at breeding TRU has as high a neutron flux as possible in the fast neutron region with a large ν value representing the number of neutrons generated per fission in order to increase the breeding ratio. Designed to be The fast reactor was designed with attention paid only to fissile Pu, that is, Pu-239 and Pu-241, which are important for maintaining criticality. In light water breeder reactors, this concept was followed and the amount of water that slows down the neutrons to minimize the energy required to cool the fuel rods to increase the neutron energy in the field. However, light water used as a coolant in light water reactors classified as thermal neutron reactors has the following two major features compared to coolants such as heavy water of other reactor types, graphite, Na, and Pb. Have. The first feature is that the hydrogen atom of light water, which is a moderator for neutrons, has a mass comparable to that of neutrons and has a high moderation capability. It is possible to supply many neutrons in the resonance and thermal energy regions that occupy part. The second feature is that the scattering cross section of the hydrogen atom is a large value of about 20 burns from the thermal energy to about 10 keV, but the scattering cross section starts to decrease rapidly from around 10 keV and is 10 burns at 200 keV, 4 MeV or more. Since it is less than 2 burns and smaller than the total cross section of Na and decreases to 1 burn at 10 MeV, the fast neutron flux of 0.1 MeV is higher than other systems, and it contributes to fast fission of even nuclides. Many neutrons can be supplied to the energy region.

  The inventors have fully considered the first and second features described above, and paid attention to not only fissile Pu but all TRU nuclides, and a core that is a feature of a boiling water reactor (BWR), which is a kind of light water reactor. Since the cooling water is separated by the channel box of each fuel assembly, it is possible to load fuel fuel assemblies with different structures inside the fuel assembly into one core, and to install fuel assemblies with different isotope ratios. A function to change the neutron energy spectrum by changing the water-to-fuel volume ratio of the assembly when it is necessary to load the core, and to change the neutron energy spectrum by controlling the core flow rate. Is used to adjust the isotope ratio and recycle the TRU isotope ratio substantially constant between cycles in each cycle so that Pu-2 in the TRU Newly found that if the abundance ratio of 9 is kept below a certain value, it will be possible to provide a light water reactor that can increase, maintain, or quickly decrease TRU with sufficient thermal margin while maintaining a negative void coefficient. It was.

  The present invention aims at function expansion and performance improvement of a recycle type light water reactor. The present invention uses TRU as a nuclear fuel to improve the performance as a breeder reactor in the light water reactor shown in Patent Document 1 and to dispose of it as a long-lived radioactive waste when it becomes unnecessary. However, in the case where all TRUs other than one core TRU are fissioned at the end, the burnup of the fuel assembly including the TRU is further increased, and multiple recycling of the TRU is made possible.

  Here, an outline of the parfait-type core will be described. The parfait-type core is a new fuel assembly to be loaded (with a burnup of zero). The lower blanket area, lower fuel area, internal blanket area, upper fuel area, and upper blanket area are arranged in this order from the lower end to the upper end. The fuel assembly arranged in is used. For this reason, the lower core area, the lower fuel area, the inner blanket area, the upper fuel area, and the upper blanket area are also formed in the parfait core from the lower end portion toward the upper end portion. The lower fuel region and the upper fuel region contain TRU oxide fuel (or a mixed oxide fuel of TRU oxide and uranium oxide). A core in which there is no internal blanket region and one fuel region between the upper blanket region and the lower blanket region is referred to as a one-region core. The fuel region of this one-region core also includes TRU oxide fuel (or a mixed oxide fuel of TRU oxide and uranium oxide).

  The present invention is directed to the above-described recycle type light water reactor and the core of the light water reactor. In the following, the results of investigations by the inventors will be described. An example of a BWR core having an electrical output of 1350 MW and a growth ratio of 1.01 having 720 fuel assemblies loaded in the core and 271 fuel rods per fuel assembly will be described.

  In this BWR core, a fuel assembly having a burnup of 45 GWd / t in the core region including the upper and lower fuel regions and the internal blanket region excluding the upper and lower blanket regions described in Patent Document 1 and Non-Patent Document 1 When further increasing the burnup of the body, if the burnup is increased with the conventional fuel assembly, the following problems occur. This problem is due to the lack of reactivity to maintain criticality, deterioration of the isotope ratio of Pu, reduction of the growth ratio, and the void coefficient, which is a safety indicator, becoming positive. It will be forced to stop at. That is, multiple recycling becomes impossible.

  In order to continue the TRU recycling while safely operating the BWR having the above BWR core, it is necessary to keep the void coefficient within a predetermined range. As a result of the study, the inventors set the core flow rate, which is a function unique to the BWR, to a predetermined value to adjust the void ratio of the core, and as a result, adjust the neutron energy spectrum to obtain the burnup degree of the fuel assembly. It has been newly found that TRU can be further increased and multiple recycling of TRUs can be realized. By setting the core flow rate found by the inventors, the ratio of the plurality of TRU isotopes present in the BWR core at the end of the operation of the BWR in a certain operation cycle can be determined by the BWR in that operation cycle. The inventors have a new finding that the operation can be started, for example, the ratio of those isotopes of the TRU existing in the BWR core immediately before the start of the operation in the operation cycle can be substantially the same. It was found by. The void coefficient can also be maintained within a predetermined range (substantially constant) during the operation cycle. The BWR core is loaded with a new fuel assembly (a fuel assembly with a burnup of zero) and a fuel assembly that has stayed in the core for at least one operating cycle immediately before the start of the operation. Yes. Paying attention to a certain fuel assembly loaded in the BWR core, this fuel assembly has experienced operation in, for example, four operation cycles in the core before being taken out of the BWR core as spent fuel. Will do. By adjusting the above-described core flow rate found by the inventors, the ratio of a plurality of TRU isotopes contained in the fuel assembly at the time when the fuel assembly is taken out from the BWR core is determined as the fuel assembly. It can be substantially the same as the ratio of those isotopes of the TRU that the body contains in the state of a new fuel assembly loaded into the BWR core. A new fuel assembly is a fuel assembly with zero burnup that has not been operated in a nuclear reactor.

  The ratio of multiple isotopes of TRUs existing in the BWR core at the end of the BWR operation in a certain operation cycle described above is present in the BWR core in a state where the operation of the BWR in the operation cycle can be started. For the sake of convenience, it is referred to as maintaining the ratio of the TRU isotopes. Also, the ratio of the multiple TRU isotopes contained in the fuel assembly at the time of removal from the BWR core as spent fuel is included in the state of the new fuel assembly loaded in the BWR core. Substantially the same as the ratio of those isotopes of TRUs that are still another view of maintaining the ratio of TRU isotopes.

The above-described core flow rate adjustment in the BWR is performed so as to obtain a relative flow rate determined by the characteristics shown in FIG. FIG. 2 shows the relationship between the ratio of Pu-239 in TRU contained in the new fuel assembly loaded in the core and the relative core flow rate. FIG. 2 shows the core flow rate at which the ratio of TRU isotopes with different ratios of Pu-239 in the TRU contained in the new fuel assembly can be maintained. As a result of investigating the composition of each isotope of TRU contained in the spent fuel assembly generated in the light water reactor, the inventors have set the core flow rate by paying attention to the ratio of Pu-239 in the TRU. This is a new finding that the proportion of elements can be maintained.
The above core flow rate is set in order to keep the void coefficient in each operation cycle within a predetermined range, and depends on the ratio of Pu-239 in the TRU contained in the new fuel assembly loaded in the core. This is performed based on the relative core flow rate (referred to as the set core flow rate) determined by FIG. In each operation cycle, at the time when the reactor power reaches at least the rated power, the core flow rate is at least adjusted to the above set core flow rate. The core flow rate is maintained at the set core flow rate until the operation cycle ends. For this reason, the control of the reactor power is performed using control rods.

  As another measure for solving the lack of reactivity, it is conceivable to increase the ratio of Pu-239 in the TRU in each fuel rod. In this other measure, the ratio of each isotope of TRU in the core at the end of operation of one operation cycle can be made substantially the same as the ratio at the start of operation of that operation cycle. Disappear. If they are made substantially the same, the core flow rate needs to be reduced from the set core flow rate, so that the MCPR criterion, which is one of the thermal limiting conditions, cannot be satisfied. As a result of the study by the inventors, as shown in FIG. 2, the ratio of Pu-239 in all TRUs contained in the new fuel assembly is lowered to 45% or less, so that all the constraints are satisfied and the growth ratio is 1.01. While maintaining the above, it was possible to improve the core performance, such as increasing the burnup of the fuel assembly, and to achieve multiple recycling of TRUs. In order to increase the burnup of the fuel assembly and to achieve multiple TRU recycling by effectively using TRU, the ratio of Pu-239 in all TRUs included in the new fuel assembly is set to 40%. It is desirable to make it the range below 45% above. By making the ratio within that range, the amount of TRU in the BWR core can be kept constant until the end of the operating cycle without reducing it compared to the start of the operating cycle, and in some cases increases at the end of the operating cycle. Can be made.

  Next, another BWR core having an electric output of 1350 MW and loaded with 720 fuel assemblies having 331 fuel rods per fuel assembly will be described as an example. This BWR core has a function of extinguishing the TRU.

  In the case of repeating TRU recycling for the purpose of reducing TRU, there is usually a problem that only odd-numbered nuclides burn first and become subcritical in the middle, and TRU remains unburned. This problem can be solved by burning the TRU while keeping the ratio of each isotope of the TRU substantially constant by maintaining the ratio of the TRU isotopes found by the inventors. For this reason, the burnup of the fuel assembly can be further increased, and multiple recycling of TRUs can be realized. However, in order to reduce TRU, it is necessary to reduce the ratio of Pu-239 in all TRUs in the TRU, and to reduce the amount of Pu-239 newly supplied from U-238 every time it is recycled.

  FIG. 3 shows the ratio of Pu-239 in TRU contained in the new fuel assembly to be loaded and the water-to-nuclear fuel material ratio (fuel pellets occupying the cross-sectional area of the unit fuel rod lattice) in the parfait-type core and the one-region core. The ratio of the cross-sectional area). These relationships shown in FIG. 3 are obtained by the inventors on the assumption that the TRU isotope ratio retention is realized in each core. Characteristic 41 is for a parfait core, and characteristic 42 is for a one-zone core. As shown in FIG. 3, by reducing the diameter of the fuel rods in the new fuel assembly, the ratio of water to TRU increases (the ratio of water to nuclear fuel material decreases), so each neutron in the resonance region and thermal energy region increases. To do. The increase of these neutrons promotes neutron capture of even-numbered nuclides, improves the transmutation efficiency from even-numbered nuclides to odd-numbered nuclides, and increases the TRU combustion efficiency. Therefore, TRU can be reduced more quickly. The TRU combustion efficiency is defined as the net decrease in TRU with respect to the total fission amount during the life of the fuel assembly.

  FIG. 4 shows the relationship between the ratio of Pu-239 in TRU contained in the new fuel assembly to be loaded, the combustion efficiency, and the void coefficient in the parfait-type core and the one-region core. The relationship between the ratio of Pu-239 in the TRU and the combustion efficiency is represented by a characteristic 10 for a parfait type and a characteristic 43 for a one-region core. The relationship between the ratio of Pu-239 in the TRU and the void coefficient is as follows. The characteristic 11 of the core extraction burnup 47 GWd / t, the characteristic 12 of the core extraction burnup 65 GWd / t, and the one-region core for the parfait core. Is represented by the characteristic 44 of the core take-off burnup 75 GWd / t. As shown in FIG. 4, the void coefficient in the system with less Pu-239 is less Pu-239 as long as the ratio of each isotope of TRU during combustion is kept substantially constant. The higher the fast energy component having a positive void reactivity component, the less it will be relative to the growth ratio 1 reactor core, so it remains negative. Further, since the amount of water in the core is larger than that in the core having a growth ratio of 1, there is no problem with MCPR. On the other hand, in the parfait-type core, when the proportion of Pu-239 in the total TRU is less than 8%, the proportion of even-numbered nuclides that fission only in the high-speed energy region increases, so the core height increases to maintain criticality and voids. Since the reaction rate becomes positive, the safety standards for light water reactors cannot be met. A one-region core whose core height is lower than the puffer type core and the void reaction rate is more negative is also a light water reactor because the void reaction rate becomes positive when the percentage of Pu-239 in the total TRU is less than 3%. The safety standards cannot be met. In order to satisfy the safety standards for light water reactors, it is necessary that the ratio of Pu-239 in all TRUs be 8% or more in the parfait core and 3% or more in the one-region core. In order to reduce the TRU in the core, the parfait-type core must make the ratio of Pu-239 in the total TRU 8% or more and less than 40%, and the one-region core must make the ratio 3% or more and less than 40%. . However, since the reactor power of the one-region core is small, when the ratio of Pu-239 in the total TRU is larger than 8%, even if the combustion efficiency of the TRU is slightly high, the net decrease amount of the TRU is lowered. By making the ratio 3% or more and 8% or less, the combustion efficiency of TRU can be further increased, and the net reduction amount of TRU can be increased. Note that when the ratio of Pu-239 in the total TRU is 15% or less within a range of 3% or more and less than 40%, the TRU combustion efficiency is remarkably increased, so that TRU can be rapidly reduced.

  In both reactor cores, if the ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice in the channel box exceeds 55%, the gap between the fuel rods becomes less than 1 mm, making assembly of the fuel assembly extremely difficult. become. For this reason, the ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice needs to be 55% or less. If the area ratio is less than 30%, the fuel rods become too thin and the amount of nuclear fuel material in the cross section decreases, so the length of the fuel rods needs to be increased and the void coefficient becomes positive. Therefore, the area ratio must be 30% or more.

  It is also possible to load a new fuel assembly manufactured using nuclear fuel material from which TRU is reprocessed to remove minor actinides into a core (for example, a parfait-type core). Even in such a core, by adjusting the core flow rate so as to be a set core flow rate determined in accordance with the ratio of Pu-239 in the total Pu contained in the new fuel assembly loaded in the core, The above TRU isotope ratio retention can be realized. In the case of using a nuclear fuel material from which minor actinides are removed as a nuclear fuel material, in order to further increase the burnup of the fuel assembly and realize multiple recycling of TRU, Pu− in all Pu contained in the new fuel assembly It is necessary to set the ratio of 239 to 3% to 50% and the ratio of Pu-240 in the total Pu to 35% to 45%. When the ratio of Pu-239 exceeds 50%, the heat removal capability decreases, so it is necessary to lower the reactor power from the rated power. Therefore, the power generation capacity possessed by the BWR cannot be fully utilized. For the above reasons, the Pu-239 ratio needs to be 50% or less. Further, since the void coefficient becomes positive when the ratio of Pu-239 in all Pu is less than 3%, the ratio of Pu-239 must be 3% or more. When the ratio of Pu-240 in the total Pu exceeds 45%, the void coefficient becomes positive. Therefore, the ratio of Pu-240 must be 45% or less. Further, when the ratio of Pu-240 in the total Pu is less than 35%, the heat removal capability is lowered, so that the power generation capability possessed by the BWR cannot be fully utilized. Therefore, the ratio of Pu-240 needs to be 35% or more.

  Hereinafter, embodiments of the present invention will be described in detail with reference to the drawings.

  A light water reactor according to embodiment 1, which is a preferred embodiment of the present invention, will be described in detail below with reference to FIGS. The light water reactor of this example is provided with a core for an electric output of 1350 MW.

However, the output scale is not limited to this. By changing the number of fuel assemblies loaded in the core, it is possible to realize a light water reactor having another power scale to which the present embodiment can be applied.

  An outline of a BWR for an electric output of 1350 MW, which is a light water reactor of the present embodiment, will be described with reference to FIG. In the BWR 19, a reactor core 20, a steam separator 21, and a steam dryer 22 are disposed in a reactor pressure vessel 27. The core 20 is a parfait-type core and is surrounded by a core shroud 25 in the reactor pressure vessel 27. The steam / water separator 21 is disposed above the core 20, and the steam dryer 22 is disposed above the steam / water separator 21. A plurality of internal pumps 26 are installed at the bottom of the reactor pressure vessel 27, and an impeller of the internal pump 26 is disposed in a downcomer formed between the reactor pressure vessel 27 and the core shroud 25. A main steam pipe 23 and a water supply pipe 24 are connected to the reactor pressure vessel 27. As shown in FIG. 6, 720 fuel assemblies 1 are loaded on the core 20. Three fuel assemblies 1 are provided with Y-shaped control rods 2 at a ratio of one, and 223 control rods 2 are arranged so as to be inserted into the core 20. Each control rod 2 is connected to a separate control rod drive 29 provided at the bottom of the reactor pressure vessel 27. The control rod drive device 29 is motor-driven and can finely adjust the movement of the control rod 2 in the axial direction. The control rod drive device 29 performs operations of pulling out the control rod 2 from the core 20 and inserting the control rod 2 into the core 20. A plurality of local output region detection devices (LPRM) 32 which are neutron detection devices are arranged in the core 20. These LPRMs 32 are connected to an average output area monitor (APRM) 31, and the APRM 31 is connected to a control rod drive control device 30.

FIG. 7 shows a cross section of the fuel assembly lattice. In the fuel assembly 1, 271 fuel rods 3 having a diameter of 10.1 mm are arranged in an equilateral triangular lattice within a channel box 4 which is a hexagonal cylinder. The cross section of the fuel assembly 1 has a hexagonal shape, and the gap between the fuel rods 3 is 1.3 mm. A plurality of fuel pellets (not shown) made of nuclear fuel material are arranged in a cladding tube (not shown) of the fuel rod 3 so as to be aligned in the axial direction. Nine fuel rods 3 are arranged in the outermost fuel rod row. The ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice in the channel box 4 is 54%. The Y-shaped control rod 2 has three wings extending outward from a tie rod located at the center. Each wing is provided with a plurality of neutron absorbing rods filled with B ₄ C, and is arranged around the tie rod with an interval of 120 degrees. The control rod 2 is provided with a follower portion made of carbon, which is a substance having a lower deceleration ability than light water, at an insertion end portion that is first inserted into the reactor core 20.

  When the BWR 19 is in operation, the coolant in the downcomer is pressurized by the rotation of the internal pump (coolant supply device) 26 and supplied to the core 20. The coolant supplied into the core 20 is guided into each fuel assembly 1 and heated by heat generated by the fission of the fissile material, and a part thereof becomes steam. The coolant in the gas-liquid two-phase flow state is led from the core 20 to the steam / water separator 21 to separate the steam. The water is further removed from the separated steam by the steam dryer 22. The steam from which moisture has been removed is supplied to the turbine (not shown) through the main steam pipe 23, and the turbine is rotated. A generator (not shown) connected to the turbine is rotated to generate electric power. The steam discharged from the turbine is condensed in a condenser (not shown) to become condensed water. The condensed water (feed water) is guided into the reactor pressure vessel 27 through the feed water pipe 24. The liquid coolant separated by the steam separator 22 is mixed with the above-mentioned water supply in the downcomer and is pressurized by the internal pump 26 again.

The arrangement of the fuel assembly 1 in the core 20 in the state of an equilibrium core will be described with reference to FIG.
The fuel assembly 1D having the longest in-reactor period and the fourth operation cycle is arranged in the outermost peripheral region of the core having a low neutron importance. A fuel assembly 1A having the highest neutron infinite multiplication factor and a stay in the reactor in the first cycle is loaded in the outer region located inside thereof, and the power distribution in the radial direction of the core is flattened. I am trying. The fuel assemblies 1B, 1C, and 1D having a residence time in the reactor of the second to fourth cycles are distributed in the core inner region. By such a distributed arrangement, the power distribution in the core inner region is flattened. The fuel assemblies 1A, 1B, 1C, and 1D are the fuel assemblies 1 shown in FIG. 7 and FIGS. 10 and 11 described later, respectively. Lower tie plates (not shown) of these fuel assemblies are supported by a plurality of fuel support fittings (not shown) provided on a core support plate (not shown) arranged at the lower end of the core 20. The A coolant passage for guiding the coolant to the fuel assembly is formed in the fuel support fitting, and an orifice (not shown) installed in the fuel support fitting is disposed at the inlet of the coolant passage. The core 20 is formed with two regions of the outermost peripheral region 6 and the inner region 1 located inside thereof in the radial direction (see FIG. 9). The diameter of the orifice located in the outermost peripheral area 2 where the output of the fuel assembly 1 is small is smaller than the diameter of the orifice located in the inner area 1.

  As shown in FIG. 10, the fuel assembly 1 has an active fuel length portion from the upper end toward the lower end, and as shown in FIG. 10, the upper blanket region 5, the upper fuel region 6, the inner blanket region 7, the lower fuel region 8 and the lower blanket. Five regions of region 9 are sequentially formed. The height of each region is as follows. The height of the upper blanket region 5 is 120 mm, the height of the upper fuel region 6 is 227 mm, the height of the inner blanket region 7 is 450 mm, the height of the lower fuel region 8 is 225 mm, and the lower blanket The height of the region 9 is 180 mm. When the fuel assembly 1 is a new fuel assembly with zero burnup, all the fuel rods 3 of the fuel assembly 1 are filled with deteriorated uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. Is filled with a mixed oxide fuel in which deteriorated uranium is mixed at a weight ratio of 172 when the weight of TRU is 100. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket region is not filled with the mixed oxide fuel. In each blanket region, natural uranium or depleted uranium recovered from the spent fuel assembly may be used instead of depleted uranium.

  The fuel assembly 1 includes five types of fuel rods 3 shown in FIG. These fuel rods 3 are fuel rods 3A to 3E. The fuel rods 3A to 3E are arranged in the fuel assembly 1 as shown in FIG. The mixed oxide fuel filled in each of the upper fuel region 6 and the lower fuel region 8 of each of the fuel rods 3A to 3E has a fissile Pu enrichment of 10.3 in the fuel rod 3A in the state of the new fuel assembly. 7 wt%, fuel rod 3B has an enrichment of 13.5 wt%, fuel rod 3C has an enrichment of 16.8 wt%, fuel rod 3D has an enrichment of 18.2 wt%, and fuel rod 3E has The enrichment is 19.5 wt%. Although there is no TRU in each blanket region of each fuel rod 3, each mixed oxide fuel in the upper fuel region 6 and lower fuel region 8 of each fuel rod 3 contains TRUs having the composition shown in Table 1. . The fuel assembly 1 is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 44 wt%. Table 1 also shows that after the fuel assembly 1 is removed from the core 20, it stays outside the reactor for a total of 3 years, 2 years in the fuel storage pool and fuel reprocessing facility, and 1 year in the fuel production facility, and again in the core. It is also the TRU composition when loaded as new fuel.

  After stopping the operation of the BWR 19 in a certain operation cycle, among the fuel assemblies 1 arranged in the core 20 which is an equilibrium core, for example, a fuel assembly in which a quarter of the fuel assemblies 1 has a burnup of zero (New fuel assembly) 1 is replaced. After the new fuel assembly 1 is loaded on the core 20, the operation of the BWR 19 in the next operation cycle is started. The new fuel assembly 1 will be the first fuel assembly in the next operation cycle. The internal pump 26 is driven, and the coolant is supplied to the core 20 as described above. The coolant flow rate (core flow rate) supplied to the core 20 is a minimum flow rate. The rotational speed of the internal pump 26 is controlled by a core flow rate control device (coolant flow rate control device) 33. Based on the control signal from the control rod drive control device 30, the control rod drive device 29 is driven, and the control rod 2 is pulled out of the core 20. After the temperature rise and pressure increase of the BWR 19 that has reached a critical state, the control rod 2 is further pulled out, so that the reactor power increases. The increase in reactor power due to the withdrawal of the control rod 2 is temporarily stopped.

  A storage device (not shown) of the core flow rate control device 33 stores the characteristics shown in FIG. The operator uses an input device (not shown) to obtain data on the ratio of Pu-239 in all TRUs included in the new fuel assembly 1 loaded in the core 20 at the time of the fuel change (referred to as ratio information), that is, Enter 44 wt%. The core flow rate control device 33 sets a set core flow rate determined based on the input ratio information and the characteristics shown in FIG. According to the characteristics shown in FIG. 2, the set core flow rate corresponding to 44 wt% that is the ratio information is a relative core flow rate of 1.00. The core flow rate control device 33 increases the rotational speed of the internal pump 26 and increases the core flow rate until the set core flow rate is reached. When the core flow rate reaches the set core flow rate, the core flow rate control device 33 stops the increase in the rotational speed of the internal pump 26 and stops the increase in the core flow rate. Thereafter, the core flow rate is maintained at the set core flow rate by the core flow rate control device 33 until the operation of the BWR 19 in the operation cycle is stopped. As the core flow rate is increased to the set core flow rate, the reactor power also increases. After the increase in the core flow rate is stopped, the withdrawal of the control rod 2 is resumed, and the reactor power is increased to 100%, which is the set power. At the end of the operation cycle, the control rod 2 is inserted into the core 20 and the operation of the BWR 19 is stopped.

  FIG. 12 shows the power distribution and void ratio distribution in the axial direction of the core 20 when the BWR 19 is operated at a reactor power of 100%. The average void ratio of the core is 67%, and the steam weight ratio at the core outlet is 41 wt%. When the BWR 19 is operated at the set output, each LPRM 32 outputs a detection signal by detecting neutrons generated by fission. The APRM 31 averages these input detection signals to obtain the reactor power, and this reactor power is input to the control rod drive control device 30. The control rod drive control device 30 operates the control rod drive device 29 and pulls out the control rod 2 from the core 20 so that the input reactor power becomes the rated output. In this way, the reactor power is maintained at the rated power during the operating cycle.

  In the present embodiment in which the ratio of Pu-239 in all TRUs included in the new fuel assembly 1 is 44 wt% and the set core flow rate, that is, the relative core flow rate is 1.00, the TRU isotope ratio retention is realized. The reason why this can be done will be specifically described below using the actinide nuclide production decay chain shown in FIG.

  The absolute amounts of the plurality of TRU isotopes shown in Table 1 included in the new fuel assembly 1 are the reactors until the new fuel assembly 1 is removed from the core 20 as a spent fuel assembly. Decreases during internal stay period (4 driving cycles). However, since the transmutation shown in the chain of actinide nuclide generation occurs, it is included in the fuel assembly 1 when it is taken out from the core 20 as a spent fuel assembly and loaded again as a new fuel assembly in the core. The ratio of each isotope of TRU is substantially the same as that of the new fuel assembly 1 described above. As typical isotopes of the above TRU, Pu-239, Pu-240, Pu-241 and Am-243 shown in Table 1 will be described as examples. The amount of Pu-239 contained in the upper fuel region 6 and the lower fuel region 8 of the new fuel assembly 1 decreases when the new fuel assembly 1 is removed from the core 20 as a spent fuel assembly. . However, during the four operating cycles, U-238 present in each blanket region is converted to Pu-239 by neutron capture reaction and subsequent β decay, and new Pu-239 is generated. Pu-240 also decreases in the upper fuel region 6 and the lower fuel region 8 when the fuel assembly 1 is taken out from the core 20, but is newly generated from U-238 in each blanket region. In Am-243, the ratio of newly generated TRU isotopes in the upper fuel region 6 and the lower fuel region 8 is balanced with the decrease rate of Am-243 due to neutron capture. The amount of Pu-241 generated in each blanket region is larger than the amount of decrease due to fission in the upper fuel region 6 and the lower fuel region 8. Pu-241 gains about 20% weight in the spent fuel assembly over the state of the new fuel assembly. However, Pu-241 has a relatively short half-life of 14.4 years, so it is removed as a spent fuel assembly from the core 20 and reduced by collapse while it is loaded as new fuel in the core again. To do. For this reason, the ratio of each isotope of the TRU contained in the fuel assembly when it is taken out as a spent fuel assembly and loaded again as a new fuel assembly in the core is the same as that in the new fuel assembly 1. It becomes substantially the same as the ratio of elements. In addition, the ratio of the plurality of TRU isotopes present in the BWR core at the end of the operation of the BWR in the operation cycle is also present in the BWR core in a state where the operation of the BWR in the operation cycle can be started. The ratio of those isotopes of TRU is substantially the same.

  According to the present embodiment, the core flow rate control device 33 adjusts the core flow rate so that it becomes a set core flow rate determined by 44 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1. As a result of this adjustment, the neutron energy spectrum is also adjusted. As described above, the corresponding isotopes are reduced in the TRU contained in the upper fuel region 6 and the lower fuel region 8 and the isotopes are generated in each blanket region. In the TRU isotopes that are hardly generated in the blanket region, the ratio of the TRU isotopes described above is due to the disappearance of this isotope and the generation of TRUs from other isotopes contained in the upper fuel region 6 and the lower fuel region 8. Retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In this embodiment, the nuclear non-proliferation resistance can be further increased while keeping safety constraints. In this embodiment, since the ratio of Pu-239 in TRU is 44 wt%, the amount of TRU in the fuel assembly 1 taken out from the core 20 can be increased as compared with the new fuel assembly 1.

Specifically, according to the present embodiment, in the BWR 19 that generates the same electrical output 1350 MW as the ABWR in the reactor pressure vessel 27 having the same size as the current ABWR, the upper and lower portions excluding the upper and lower blanket regions. The core region including the fuel region and the internal blanket region has a take-off burnup of 45 GWd / t, and the core region take-off burnup of 54 GWd / t, which is higher than that of the light water reactor breeding reactor described in Patent Document 1, upper part In addition, the core 20 including the lower blanket region has a take-off burnup of 47 GWd / t. Further, in this example, the void coefficient is −2 × 10 ⁻⁶ Δk / k /% void, the MCPR is 1.3, and the ratio of each isotope of TRU is kept substantially constant as described above. In this state, a growth ratio of 1.01 can be realized.

  In this embodiment, the control of the reactor power when the reactor power is reduced from the set reactor power (for example, the rated power) is not controlled by the core flow control device 33 but by the control rod drive control device 30. This is performed by operating (inserting) the control rod 2 by controlling 29. Therefore, the present embodiment can achieve both the retention of the TRU isotope ratio and the control of the reactor power.

  A light water reactor according to embodiment 2, which is another embodiment of the present invention, will be described in detail below with reference to FIGS. 13 and 14 and Table 2. FIG. The light water reactor of the present embodiment is the core shown in FIG.

20A has a configuration in which the fuel assembly 1 is replaced with the fuel assembly 1H shown in FIG. 14, and the other configurations are the same as those of the first embodiment. The configuration of the present embodiment will be described with respect to differences from the first embodiment, and the description of the same configuration as the first embodiment will be omitted. The core 20A is a parfait type core.

  The fuel assembly 1H arranged in the core 20A has the same configuration as the fuel assembly 1 used in Example 1 except for the dimensions shown in FIG. 14 and the composition of the TRU shown in Table 2. Also in the fuel assembly 1H, the ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is 54%, which is the same as in the first embodiment. In the core 20A, fuel assemblies 1A to 1E are arranged as shown in FIG. The fuel assembly 1E is the fuel assembly having the longest residence time in the reactor and the fifth operation cycle, and is arranged in the outermost peripheral region of the core 20A. The fuel assemblies 1A for the first cycle are arranged in the outer region of the core located inside the outermost peripheral region of the core, and the fuel assemblies 1B, 1C, 1D of the second to fourth cycles are dispersedly arranged in the inner region of the core. Has been. Note that three fuel assemblies 1D are arranged in the outermost peripheral region of the core. With such an arrangement of the fuel assemblies, the power distribution in the radial direction of the core 20A is flattened. Each of the fuel assemblies 1A to 1E used in the present embodiment is a fuel assembly 1H. High burnup is achieved by using mixed oxide fuel of plutonium, Am-241, and depleted uranium from which TRU is reprocessed and minor actinides are removed. Am-241 contained in the new fuel is produced by the collapse of Pu-241 in the plutonium before the plutonium, which has been reprocessed TRU and the minor actinides removed, is loaded into the core 20 as the new fuel. .

  Like the fuel assembly 1, the fuel assembly 1H forms five regions in the fuel effective length portion. As shown in FIG. 14, the upper blanket region 5 has a height of 200 mm, the upper fuel region 6 has a height of 211 mm, the inner blanket region 7 has a height of 310 mm, and the lower fuel region 8 has a height. Is 207 mm, and the height of the lower blanket region 9 is 220 mm. When the fuel assembly 1H is a new fuel assembly with zero burnup, all the fuel rods 3 of the fuel assembly 1H fill the three blanket regions with depleted uranium, and the upper fuel region 6 and the lower fuel region 8 Is a mixed oxidation in which depleted uranium is mixed at a ratio of 198 when the total weight of plutonium from which TRU is reprocessed to remove minor actinides and Am-241 generated by the decay of Pu-241 after reprocessing is 100 Filled with fuel. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket region is not filled with the mixed oxide fuel. Similarly to the first embodiment, the fuel assembly 1J also includes fuel rods 3A to 3E. These fuel rods 3 </ b> A to 3 </ b> E are fuel rods 3. Each mixed oxide fuel present in the upper fuel region 6 and the lower fuel region 8 has the composition shown in Table 2. In the fuel assembly 1H, the ratio of Pu-239 in all Pu and Am-241 is 48.6 wt%, and the ratio of Pu-240 in all Pu and Am-241 is 39.7 wt. %.

  The core flow control device 33 removes the minor actinides by reprocessing the TRU in the same manner as in FIG. 2 and the ratio information input from the input device (the ratio of Pu-239 in all Pu and Am-241 is 48.6 wt%). In a reactor core using Pu and Am-241 as a new fuel assembly, all Pu and Am-241 isotopes for each of a plurality of cores in which the ratio of Pu-239 in all Pu and Am-241 in the new fuel assembly is different. The set core flow rate determined based on the characteristics obtained by obtaining the core flow rate capable of maintaining the element ratio is set. As in the first embodiment, the core flow rate control device 33 increases the rotational speed of the internal pump 26 and increases the core flow rate until the set core flow rate is reached. The core flow rate control device 33 stops the increase in the rotation speed of the internal pump 26 when the core flow rate reaches the set core flow rate. Thereafter, the core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 48.6 wt%, which is the ratio of Pu-239 in the total Pu in the new fuel assembly 1H loaded in the core 20A, Similarly, the retention of the TRU isotope ratio can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. Also in this embodiment, the amount of TRU in the fuel assembly 1H taken out from the core 20 can be increased as compared with the new fuel assembly 1H.

More specifically, according to the present embodiment, the BWR 19 that generates the same electrical output 1350 MW as the ABWR in the reactor pressure vessel 27 having the same size as the current ABWR has a higher burnup than the first embodiment. The take-off burnup 51 GWd / t of the core 20A and the take-off burnup 68 GWd / t in the core region excluding the upper and lower blanket regions can be realized. In this example, the void coefficient is −3 × 10 ⁻⁵ Δk / k /% void, the MCPR is 1.3, and the ratio of each of the Pu and Am-241 isotopes is substantially the same as described in Example 1. Thus, a growth ratio of 1.01 can be realized while keeping the temperature constant.

  A light water reactor according to embodiment 3, which is another embodiment of the present invention, will be described in detail below with reference to FIGS. The light water reactor according to the present example is the core 2 shown in FIG.

0B has a configuration in which the fuel assembly 1 is replaced with the fuel assembly 1J shown in FIGS. 15 and 17, and the other configurations are the same as those of the first embodiment. The configuration of the present embodiment will be described with respect to differences from the first embodiment, and the description of the same configuration as the first embodiment will be omitted. The core 20B is a parfait type core.

The fuel assembly 1J arranged in the core 20B will be described with reference to FIG. The fuel assembly 1J has a hexagonal cross section, and 331 fuel rods 3J having a diameter of 9.2 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3J is 1.1 mm. The ratio of the cross sectional area of the fuel pellets to the cross sectional area of the unit fuel rod lattice is 53%. In the reactor core 20B, fuel assemblies 1A to 1D are arranged as shown in FIG. Similar to the core 20, the fuel assembly 1 </ b> D having the longest residence time in the reactor and the fourth operation cycle is disposed in the outermost peripheral region of the core 20 </ b> B. The fuel assemblies 1A for the first cycle are arranged in the outer region of the core located inside the outermost peripheral region of the core, and the fuel assemblies 1B, 1C, 1D of the second to fourth cycles are dispersedly arranged in the inner region of the core. Has been.
Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1D are annularly arranged. The power distribution in the radial direction of the core 20B is flattened. Each of the fuel assemblies 1A to 1E shown in FIG. 16 is a fuel assembly 1J.

The fuel assembly 1J, like the fuel assembly 1, forms five regions in the fuel effective length portion (see FIG. 17). The height of the upper blanket region 5 is 90 mm, the height of the upper fuel region 6 is 241 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 241 mm, and the lower blanket The height of the region 9 is 90 mm.
When the fuel assembly 1J is a new fuel assembly with zero burnup, all the fuel rods 3J of the fuel assembly 1J are filled with depleted uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. Is filled with a mixed oxide fuel in which deteriorated uranium is mixed at a weight ratio of 153 when the weight of TRU is 100. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket region is not filled with the mixed oxide fuel. Similarly to the first embodiment, the fuel assembly 1J also includes fuel rods 3A to 3E. These fuel rods 3A to 3E are fuel rods 3J. Each mixed oxide fuel present in the upper fuel region 6 and the lower fuel region 8 contains TRUs having the compositions shown in Table 3. The fuel assembly 1J is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 40.1 wt%. Each blanket region is not filled with the mixed oxide fuel.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (40.1 wt%) input from the input device and the characteristics shown in FIG. As in the first embodiment, the core flow rate control device 33 increases the rotational speed of the internal pump 26 and increases the core flow rate until the set core flow rate is reached. The core flow rate control device 33 stops the increase in the rotation speed of the internal pump 26 when the core flow rate reaches the set core flow rate. Thereafter, the core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 40.1 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1J loaded in the core 20B, it is the same as in the first embodiment. In addition, the TRU isotope ratio can be maintained. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. Also in this embodiment, the amount of TRU in the fuel assembly 1J taken out from the core 20B can be increased as compared with the new fuel assembly 1J.

More specifically, according to the present embodiment, in the reactor pressure vessel having the same size as that of the current ABWR, the BWR 19 that generates the same electric output 1350 MW as that of the ABWR, the extraction burnup 53 GWd / t of the core 20B is set. And a void coefficient of −3 × 10 ⁻⁶ Δk / k /% void can be realized. In this example, the MCPR is 1.3, and as described above, the TRU isotope ratio can be maintained and the growth ratio of 1.01 can be realized.

  The light water reactor of Example 4 which is another Example of this invention is demonstrated in detail below using FIGS. 18-20 and Table 4. FIG. The light water reactor of the present embodiment is the same as that of the first embodiment in which the core 20 is the core 2 shown in FIG.

The fuel assembly 1 is replaced by the fuel assembly 1K shown in FIGS. 18 and 20 at 0C, and other configurations are the same as those of the first embodiment. The configuration of the present embodiment will be described with respect to differences from the first embodiment, and the description of the same configuration as the first embodiment will be omitted. The core 20C is also a parfait type core.

In the fuel assembly 1K (see FIG. 18) arranged in the core 20C, 331 fuel rods 3K having a diameter of 7.7 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3K is 2.6 mm, and ten fuel rods 3K are arranged in the outermost fuel rod row. The ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is 36%.
In the reactor core 20C, as shown in FIG. 19 in the state of the equilibrium reactor, fuel assemblies 1A to 1D having different numbers of experienced operating cycles are arranged. The fuel assembly 1D in the fourth cycle is disposed in the outermost peripheral region of the core. The fuel assemblies 1A for the first cycle are arranged in the outer region of the core, and the fuel assemblies 1B, 1C, 1D of the second to fourth cycles are dispersedly arranged in the inner region of the core. Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1B are annularly arranged. In such a core 20C, the power distribution in the radial direction is further flattened. Each of the fuel assemblies 1A to 1E shown in FIG. 19 is a fuel assembly 1K.

As with the fuel assembly 1, the fuel assembly 1K forms five regions in the effective fuel length portion (see FIG. 20). The height of the upper blanket region 5 is 30 mm, the height of the upper fuel region 6 is 194 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 194 mm, and the lower blanket The height of the region 9 is 30 mm.
When the fuel assembly 1K is a new fuel assembly with zero burnup, all the fuel rods 3K of the fuel assembly 1K are filled with deteriorated uranium in the three blanket regions, Is filled with a mixed oxide fuel in which deteriorated uranium is mixed at a ratio of 7 when the weight of TRU is 100. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket region is not filled with the mixed oxide fuel. The fuel assembly 1K also includes fuel rods 3A to 3E which are fuel rods 3K. Each mixed oxide fuel present in the upper fuel region 6 and the lower fuel region 8 contains TRUs having the compositions shown in Table 4. The fuel assembly 1K is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 14.4 wt%.

  Similarly to the first embodiment, the core flow rate control device 33 sets a set core flow rate determined based on the ratio information (14.4 wt%) and the characteristics shown in FIG. The core flow rate control device 33 controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. When the core flow rate reaches the set core flow rate, the rotation of the internal pump 26 is stopped. Thereafter, the core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 14.4 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1K loaded in the core 20C, it is the same as in the first embodiment. In addition, the TRU isotope ratio can be maintained. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1K taken out from the core 20C can be reduced as compared with the new fuel assembly 1K.

More specifically, according to the present embodiment, in the BWR 19 that generates the same electrical output 1350 MW as the ABWR in a reactor pressure vessel having the same size as the current ABWR, the take-off burnup 65 GWd / t of the core 20C is Can be achieved. In this example, the TRU combustion efficiency is 44%, the void coefficient is −2 × 10 ⁻⁴ Δk / k /% void, the MCPR is 1.3, and the ratio retention of TRU isotopes can be realized. TRU can be reduced.

  A light water reactor according to embodiment 5, which is another embodiment of the present invention, will be described in detail below with reference to FIGS. The light water reactor of the present embodiment is a fuel assembly disposed in the core 20C in the fourth embodiment.

A configuration in which the body 1K is replaced with a fuel assembly 1L shown in FIGS. 21 and 22 is the same as that in the fourth embodiment. The configuration of the present embodiment will be described with respect to differences from the fourth embodiment. The core used in this embodiment is also a parfait-type core.

  The configuration of the fuel assembly 1L will be described with reference to FIGS. In the fuel assembly 1L, 331 fuel rods 3K having a diameter of 7.4 mm are arranged in an equilateral triangular lattice in the channel box 4. The gap between the fuel rods 3K is 2.9 mm, and ten fuel rods 3K are arranged in the outermost fuel rod row. The ratio of the cross sectional area of the fuel pellets to the cross sectional area of the unit fuel rod lattice is 31%. The arrangement of the fuel assemblies 1L in the radial direction of the core in the present embodiment is the same as the arrangement shown in FIG.

  As with the fuel assembly 1K, the fuel assembly 1L forms five regions in the fuel effective length portion (see FIG. 22). The height of the upper blanket region 5 is 20 mm, the height of the upper fuel region 6 is 237 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 237 mm, and the lower blanket The height of the region 9 is 20 mm. When the fuel assembly 1L is a new fuel assembly with zero burnup, all the fuel rods 3L of the fuel assembly 1L are filled with deteriorated uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. , Filled with TRU oxide fuel. The enrichment of this TRU fuel with fissile Pu is 13.3 wt%. The upper fuel region 6 and the lower fuel region 8 are not filled with a mixed oxide fuel of TRU and deteriorated uranium. Each blanket region does not contain TRU oxide fuel. The fuel assembly 1L also includes fuel rods 3A to 3E that are fuel rods 3L. Each TRU fuel existing in the upper fuel region 6 and the lower fuel region 8 includes TRUs having the composition shown in Table 5. The fuel assembly 1L is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 8.5 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (8.5 wt%) and the characteristics shown in FIG. The core flow rate control device 33 controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. When the core flow rate reaches the set core flow rate, the rotation of the internal pump 26 is stopped. Thereafter, the core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 8.5 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1L loaded in the core, as in the first embodiment. , TRU isotope ratio retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1L taken out from the core can be reduced as compared with the new fuel assembly 1L.

More specifically, according to the present embodiment, a core pressure burn-up of 65 GWd / t is realized in a BWR 19 that generates the same electrical output 1350 MW as the ABWR with a reactor pressure vessel having the same size as the current ABWR. it can. In this embodiment, the TRU combustion efficiency is 55%, the void coefficient is −3 × 10 ⁻⁵ Δk / k /% void, the MCPR is 1.3, and the ratio of TRU isotopes can be maintained. Can be reduced.

  A light water reactor according to embodiment 6, which is another embodiment of the present invention, will be described in detail below with reference to FIGS. The light water reactor according to the present embodiment is the core 2 shown in FIG.

0D has a configuration in which the fuel assembly 1 is replaced with the fuel assembly 1M shown in FIGS. 23 and 25, and the other configurations are the same as those of the first embodiment. The configuration of the present embodiment will be described with respect to the differences from the first embodiment. The light water reactor of the present embodiment has an electric output of 450 MW, and the core 20D is a one-zone core.

In the fuel assembly 1M (see FIG. 23) disposed in the core 20D, 331 fuel rods 3M having a diameter of 8.7 mm are disposed in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3M is 1.6 mm, and ten fuel rods 3M are arranged in the outermost fuel rod row. The ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is 46%.
FIG. 24 shows the state of the equilibrium core in the core 20D. The fuel assembly 1D in the fourth cycle is disposed in the outermost peripheral region of the core. The fuel assemblies 1A for the first cycle are arranged in the outer region of the core, and the fuel assemblies 1B, 1C, 1D of the second to fourth cycles are dispersedly arranged in the inner region of the core. Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1B are annularly arranged. In such a core 20D, the power distribution in the radial direction is further flattened. Each of the fuel assemblies 1A to 1E shown in FIG. 24 is a fuel assembly 1M.

The fuel assembly 1M forms three regions in the fuel effective length portion (see FIG. 25). The height of the upper blanket region 5 is 20 mm, the height of the lower blanket region 9 is 20 mm, and the height of the fuel region 15 formed between these blanket regions is 201 mm.
When the fuel assembly 1M is in the state of a new fuel assembly with zero burnup, the two blanket regions of all the fuel rods 3M are filled with deteriorated uranium, and the fuel regions 15 are filled with TRU oxide fuel . This TRU oxide fuel has a fissile Pu enrichment of 7.4 wt%. Each blanket area does not contain a TRU. The fuel assembly 1M also includes fuel rods 3A to 3E which are fuel rods 3M. The TRU oxide fuel present in the fuel region 15 includes TRUs having the compositions shown in Table 6. The fuel assembly 1M is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 4.0 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (4.0 wt%) and the characteristics shown in FIG. The core flow rate control device 33 controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. When the core flow rate reaches the set core flow rate, the rotation of the internal pump 26 is stopped. Thereafter, the core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted so as to have a set core flow rate determined by 4.0 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1K loaded in the core 20C, it is the same as in the first embodiment. In addition, the TRU isotope ratio can be maintained. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1K taken out from the core 20D can be reduced as compared with the new fuel assembly 1K.

More specifically, according to the present embodiment, in the BWR 19 that generates an electric output of 450 MW with a reactor pressure vessel having the same size as that of the current ABWR, it is possible to achieve a take-off burnup 75 GWd / t of the core 20D. In this embodiment, the TRU combustion efficiency is 80%, the void coefficient is −4 × 10 ⁻⁵ Δk / k /% void, the MCPR is 1.3, and the TRU isotope ratio can be maintained. Can be reduced.

  A light water reactor according to embodiment 7, which is another embodiment of the present invention, will be described in detail below with reference to FIGS. The light water reactor of the present embodiment is shown in FIGS.

The reactor core 20E has a configuration in which the fuel assembly 1 is replaced with the fuel assembly 1N shown in FIGS. 27 and 29, and other configurations are the same as those in the first embodiment. The configuration of the present embodiment will be described with respect to the differences from the first embodiment. The electric output of the light water reactor of the present embodiment is 830 MW, which is lower than that of the first embodiment, and the core 20E is a parfait-type core.

In the fuel assembly 1N having a square cross section arranged in the core 20E, 196 fuel rods 3N having a diameter of 8.1 mm are arranged in a square lattice in the channel box 4A. The pitch of the fuel rods 3N is 9.4 mm, and ten fuel rods 3M are arranged on the fuel rod 1N. The ratio of the cross sectional area of the fuel pellets to the cross sectional area of the unit fuel rod lattice is 41%. The reactor core 20E includes 872 bodies arranged in the reactor core 20E, and includes four cross-shaped control rods 2A in proportion to the four fuel assemblies 1N. Although not shown on the side of the gap area outside the channel box 4A in FIG. 27 where the cross-shaped control rod 2A is not inserted, a water drain plate having a function of draining water in the gap area outside the channel box 4A is a core upper lattice. It is suspended from the board.
The state of the equilibrium core in the core 20E is shown in FIG. The fuel assembly 1d of the fourth cycle and the fuel assembly 1e of the fifth cycle are arranged in the outermost peripheral region of the core. The first cycle fuel assemblies 1a are arranged in the core outer region, and the second to fourth cycle fuel assemblies 1b, 1c, 1d are distributed in the core inner region. Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1b are annularly arranged. In such a core 20E, the power distribution in the radial direction is further flattened.

  The fuel assembly 1N forms five regions in the fuel effective length portion (see FIG. 29). The height of the upper blanket region 5 is 40 mm, the height of the upper fuel region 6 is 180 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 174 mm, and the height of the lower blanket region 9 is 90 mm. is there. When the fuel assembly 1N is a new fuel assembly with zero burnup, all the fuel rods 3N of the fuel assembly 1N are filled with depleted uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. , Filled with TRU oxide fuel. The enrichment of this TRU fuel with fissile Pu is 17.8 wt%. Each blanket area does not contain a TRU. The TRU oxide fuel present in the upper fuel region 6 and the lower fuel region 8 contains TRUs having the compositions shown in Table 7. The fuel assembly 1N is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 12.9 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (12.9 wt%) and the characteristics shown in FIG. The core flow rate control device 33 controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. When the core flow rate reaches the set core flow rate, the rotation of the internal pump 26 is stopped. Thereafter, the core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 12.9 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1N loaded in the core, as in the first embodiment. , TRU isotope ratio retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1N taken out from the core can be reduced as compared with the new fuel assembly 1N.

More specifically, according to the present embodiment, in the current ABWR that can generate an electrical output of 848 MW, it is possible to realize a take-off burnup of 45 GWd / t of the core 20E. In this embodiment, the TRU combustion efficiency is 43%, the void coefficient is −2 × 10 ⁻⁵ Δk / k /% void, the MCPR is 1.3, and the ratio of TRU isotopes can be maintained. Can be reduced.

  The light water reactor of Example 8 which is another Example of this invention is demonstrated in detail below using FIGS. 30-32 and Table 8. FIG. The light water reactor of the present embodiment is the core 2 shown in FIG.

The fuel assembly 1 is replaced with the fuel assembly 1P shown in FIGS. 30 and 32 at 0F, and the other configurations are the same as those of the first embodiment. The configuration of the present embodiment will be described with respect to the differences from the first embodiment. The light water reactor of the present embodiment has an electrical output of 1350 MW, and the core 20F is a parfait type core.

  In the fuel assembly 1P arranged in the core 20F, 331 fuel rods 3P having a diameter of 8.7 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3P is 1.6 mm, and ten fuel rods 3P are arranged in the outermost fuel rod row. The ratio of the cross sectional area of the fuel pellets to the cross sectional area of the unit fuel rod lattice is 47%. The core 20F is in a state of a balanced core. The fuel assembly 1D in the fourth cycle is disposed in the outermost peripheral region of the core, and the fuel assembly 1A in the first cycle is disposed in the outer region of the core. The fuel assemblies 1B, 1C, and 1D in the second to fourth cycles are dispersedly arranged in the core inner region. Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1B are annularly arranged. In such a core 20F, the power distribution in the radial direction is further flattened. Each of the fuel assemblies 1A to 1E shown in FIG. 31 is a fuel assembly 1P.

  The fuel assembly 1P forms five regions in the fuel effective length portion (see FIG. 32). The height of the upper blanket region 5 is 90 mm, the height of the upper fuel region 6 is 240 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 240 mm, and the height of the lower blanket region 9 is 90 mm. is there. When the fuel assembly 1P is a new fuel assembly with zero burnup, all the fuel rods 3P of the fuel assembly 1P are filled with deteriorated uranium in the three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. Is filled with a mixed oxide fuel in which deteriorated uranium is mixed at a ratio of 108 when the weight of TRU is 100. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket area does not contain a TRU. The fuel assembly 1M also includes fuel rods 3A to 3E which are fuel rods 3M. The mixed oxide fuel contains TRU having the composition shown in Table 8. The fuel assembly 1P is a new fuel assembly, and the ratio of Pu-239 in all TRUs is 31.6 wt%.

  The core flow rate control device 33 sets the set core flow rate determined based on the ratio information (31.6 wt%) and the characteristics shown in FIG. 2, and controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. Let The core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 31.6 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1P loaded in the core, as in the first embodiment. , TRU isotope ratio retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1P taken out from the core can be reduced as compared with the new fuel assembly 1P.

According to the present embodiment, in the BWR 19 that generates the same electrical output 1350 MW as the ABWR in a reactor pressure vessel having the same size as that of the current ABWR, the core removal burnup of the core 20F can be increased to 57 Gwd / t. In this example, the void coefficient is −2 × 10 ⁻⁵ Δk / k /% void, the TRU combustion efficiency is 15%, the MCPR is 1.3, and the TRU isotope ratio can be maintained. Can be reduced.

  A light water reactor according to a ninth embodiment which is another embodiment of the present invention will be described below in detail with reference to FIGS. 33 to 35 and Table 9. FIG. The light water reactor according to the present embodiment is the core 2 shown in FIG.

0G has a configuration in which the fuel assembly 1 is replaced with the fuel assembly 1Q shown in FIGS. 33 and 35, and the other configurations are the same as those of the first embodiment. The configuration of the present embodiment will be described with respect to the differences from the first embodiment. The core 20G is a parfait type core.

  In the fuel assembly 1Q arranged in the core 20G, 331 fuel rods 3Q having a diameter of 8.5 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3Q is 1.8 mm, and ten fuel rods 3Q are arranged in the outermost fuel rod row. The ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is 45%. The core 20G is in a state of an equilibrium core. The fuel assembly 1D in the fourth cycle is disposed in the outermost peripheral region of the core, and the fuel assembly 1A in the first cycle is disposed in the outer region of the core. The fuel assemblies 1B, 1C, and 1D in the second to fourth cycles are dispersedly arranged in the core inner region. Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1B are annularly arranged. In such a core 20G, the power distribution in the radial direction is further flattened. Each of the fuel assemblies 1A to 1E shown in FIG. 34 is a fuel assembly 1Q.

  The fuel assembly 1Q forms five regions in the fuel effective length portion (see FIG. 35). The height of the upper blanket region 5 is 90 mm, the height of the upper fuel region 6 is 224 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 224 mm, and the height of the lower blanket region 9 is 90 mm. is there. When the fuel assembly 1Q is a new fuel assembly with zero burnup, all the fuel rods 3Q of the fuel assembly 1Q are filled with deteriorated uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. Is filled with a mixed oxide fuel in which deteriorated uranium is mixed at a weight ratio of 79 when the weight of TRU is 100. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket area does not contain a TRU. The fuel assembly 1Q also includes fuel rods 3A to 3E which are fuel rods 3Q. The mixed oxide fuel contains TRU having the composition shown in Table 9. The fuel assembly 1Q is a new fuel assembly, and the ratio of Pu-239 in all TRUs is 26.4 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (26.4 wt%) and the characteristics shown in FIG. 2, and controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. Let The core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 26.4 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1Q loaded in the core, as in the first embodiment. , TRU isotope ratio retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1Q taken out from the core can be reduced as compared with the new fuel assembly 1Q.

According to the present embodiment, in the BWR 19 that generates the same electric output 1350 MW as the ABWR in the reactor pressure vessel having the same size as that of the current ABWR, it is possible to increase the take-off burnup of the core 20G to 58 Gwd / t. In this embodiment, the void coefficient is −3 × 10 ⁻⁵ Δk / k /% void, the TRU combustion efficiency is 22%, the MCPR is 1.3, and the TRU isotope ratio can be maintained. Can be reduced.

  A light water reactor according to embodiment 10, which is another embodiment of the present invention, will be described below in detail with reference to FIGS. The light water reactor of the present embodiment is a fuel disposed in the core 20G in the ninth embodiment.

The fuel assembly 1Q is replaced with the fuel assembly 1R shown in FIGS. 36 and 37, and other configurations are the same as those of the ninth embodiment. The configuration of the present embodiment will be described with respect to differences from the ninth embodiment. The core used in this embodiment is also a parfait-type core.

  As shown in FIGS. 36 and 37, in the fuel assembly 1R, 331 fuel rods 3R having a diameter of 8.1 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3R is 2.2 mm, and ten fuel rods 3R are arranged in the outermost fuel rod row. The ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is 40%. The arrangement of the fuel assemblies 1R in the radial direction of the core in the present embodiment is the same as the arrangement shown in FIG.

  As with the fuel assembly 1Q, the fuel assembly 1R forms five regions in the effective fuel length portion (see FIG. 37). The height of the upper blanket region 5 is 40 mm, the height of the upper fuel region 6 is 212 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 212 mm, and the height of the lower blanket region 9 is 40 mm It is. When the fuel assembly 1R is a new fuel assembly with zero burnup, all the fuel rods 3R of the fuel assembly 1R are filled with depleted uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. Is filled with a mixed oxide fuel in which deteriorated uranium is mixed at a weight ratio of 39 when the weight of TRU is 100. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket area does not contain a TRU. The fuel assembly 1R also includes fuel rods 3A to 3E which are fuel rods 3R. Each mixed oxide fuel existing in the upper fuel region 6 and the lower fuel region 8 includes TRUs having the composition shown in Table 10. The fuel assembly 1R is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 19.7 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (19.7 wt%) and the characteristics shown in FIG. 2, and controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. Let The core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 19.7 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1R loaded in the core, as in the first embodiment. , TRU isotope ratio retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1R taken out from the core can be reduced as compared with the new fuel assembly 1R.

According to the present embodiment, in the BWR 19 that generates the same electric output 1350 MW as the ABWR in the reactor pressure vessel having the same size as that of the current ABWR, it is possible to increase the core burn-up burnup to 59 GWdt. In this example, the void coefficient is −4 × 10 ⁻⁵ Δk / k /% void, the TRU combustion efficiency is 34%, the MCPR is 1.3, and the ratio of TRU isotopes can be maintained. Can be reduced.

  A light water reactor according to an eleventh embodiment which is another embodiment of the present invention will be described below in detail with reference to FIGS. 38 and 39 and Table 11. FIG. The light water reactor of the present embodiment is a fuel disposed in the core 20C in the fourth embodiment.

The fuel assembly 1K is replaced with the fuel assembly 1S shown in FIGS. 38 and 39, and other configurations are the same as those in the fourth embodiment. The configuration of the present embodiment will be described with respect to differences from the fourth embodiment. The core used in this embodiment is also a parfait-type core.

  As shown in FIGS. 38 and 39, in the fuel assembly 1S, 331 fuel rods 3S having a diameter of 7.6 mm are arranged in an equilateral triangular lattice in the channel box 4. The gap between the fuel rods 3S is 2.7 mm, and ten fuel rods 3S are arranged in the outermost fuel rod row. The ratio of the cross sectional area of the fuel pellets to the cross sectional area of the unit fuel rod lattice is 35%. The arrangement of the fuel assemblies 1S in the radial direction of the core in the present embodiment is the same as the arrangement shown in FIG.

  As with the fuel assembly 1K, the fuel assembly 1S forms five regions in the fuel effective length portion (see FIG. 39). The height of the upper blanket region 5 is 35 mm, the height of the upper fuel region 6 is 189 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 189 mm, and the height of the lower blanket region 9 is 35 mm It is. When the fuel assembly 1S is a new fuel assembly having zero burnup, all the fuel rods 3S of the fuel assembly 1S are filled with deteriorated uranium in the three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. , Filled with TRU oxide fuel. This TRU oxide fuel has a fissile Pu enrichment of 18 wt%. Each blanket area does not contain a TRU. The fuel assembly 1S also includes fuel rods 3A to 3E which are fuel rods 3S, and each TRU oxide fuel present in the upper fuel region 6 and the lower fuel region 8 includes TRU having the composition shown in Table 10. The fuel assembly 1S is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 12.9 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (12.9 wt%) and the characteristics shown in FIG. 2, and controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. Let The core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 12.9 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1S loaded in the core, as in the first embodiment. , TRU isotope ratio retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1S taken out from the core can be reduced as compared with the new fuel assembly 1S.

According to the present embodiment, a core core burn-up burnup of 65 Gwd / t can be realized in the BWR 19 that generates the same electric output 1350 MW as the ABWR in a reactor pressure vessel having the same size as that of the current ABWR. In this example, the TRU combustion efficiency is 47%, the void coefficient is −3 × 10 ⁻⁴ Δk / k /% void, the MCPR is 1.3, and the ratio of TRU isotopes can be maintained. Can be reduced.

  A light water reactor according to embodiment 12, which is another embodiment of the present invention, will be described below in detail with reference to FIGS. The light water reactor of the present embodiment is a fuel disposed in the core 20C in the fourth embodiment.

The fuel assembly 1K is replaced with the fuel assembly 1T shown in FIGS. 38 and 39, and other configurations are the same as those of the fourth embodiment. The configuration of the present embodiment will be described with respect to differences from the fourth embodiment. The core used in this embodiment is also a parfait-type core.

  In the fuel assembly 1T, in the channel box 4, 331 fuel rods 3T having a diameter of 7.5 mm are arranged in an equilateral triangular lattice. The gap between the fuel rods 3T is 2.8 mm, and ten fuel rods 3T are arranged in the outermost fuel rod row. The ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is 34%. The arrangement of the fuel assemblies 1T in the radial direction of the core in the present embodiment is the same as the arrangement shown in FIG.

  As with the fuel assembly 1K, the fuel assembly 1T forms five regions in the fuel effective length portion (see FIG. 41). The height of the upper blanket region 5 is 30 mm, the height of the upper fuel region 6 is 204 mm, the height of the inner blanket region 7 is 560 mm, the height of the lower fuel region 8 is 204 mm, and the height of the lower blanket region 9 is 30 mm. It is. When the fuel assembly 1T is a new fuel assembly with zero burnup, all the fuel rods 3T of the fuel assembly 1T are filled with depleted uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 are filled. , Filled with TRU oxide fuel. This TRU oxide fuel has a fissile Pu enrichment of 16 wt%. Each blanket area does not contain a TRU. The fuel assembly 1T also includes fuel rods 3A to 3E, which are fuel rods 3T, and each TRU oxide fuel present in the upper fuel region 6 and the lower fuel region 8 includes TRUs having the composition shown in Table 12. The fuel assembly 1T is a new fuel assembly, and the ratio of Pu-239 in all TRUs is 11.0 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (11.0 wt%) and the characteristics shown in FIG. 2, and controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. Let The core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since this embodiment is also adjusted to have a set core flow rate determined by 11.0 wt%, which is the ratio of Pu-239 in TRU in the new fuel assembly 1T loaded in the core, as in the first embodiment. , TRU isotope ratio retention can be realized. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1T taken out from the core can be reduced as compared with the new fuel assembly 1T.

According to the present embodiment, a core take-off burnup of 65 Gwd / t can be realized in the BWR 19 that generates the same electric output 1350 MW as the ABWR in a reactor pressure vessel having the same size as the current ABWR. In this example, the TRU combustion efficiency is 50%, the void coefficient is −2 × 10 ⁻⁴ Δk / k /% void, the MCPR is 1.3, and the ratio of TRU isotopes can be maintained. Can be reduced.

  A light water reactor according to embodiment 13, which is another embodiment of the present invention, will be described in detail below with reference to FIG. The light water reactor of the present embodiment has a configuration in which the fuel assembly 1 arranged in the core 20 in the first embodiment is replaced with a fuel assembly 1U shown in FIG. 42, and other configurations are the same as those in the fourteenth embodiment. The configuration of the present embodiment will be described with respect to the differences from the first embodiment. The core used in this embodiment is also a parfait-type core.

  The fuel assembly 1U is formed by forming five regions of the fuel assembly 1 as shown in FIG. The other structure of the fuel assembly 1U is the same as that of the fuel assembly 1. In the fuel assembly 1U, the height of the upper blanket region 5 is 120 mm, the height of the upper fuel region 6 is 226 mm, the height of the inner blanket region 7 is 450 mm, the height of the lower fuel region 8 is 224 mm, and the lower blanket region The height of 9 is 180 mm. When the fuel assembly 1S is a new fuel assembly with zero burnup, all the fuel rods of the fuel assembly 1U are filled with depleted uranium in three blanket regions, and the upper fuel region 6 and the lower fuel region 8 When the weight of TRU is 100, a mixed oxide fuel in which deteriorated uranium is mixed at a weight ratio of 172 is filled. The enrichment of the fissile Pu of this mixed oxide fuel is 18 wt%. Each blanket area does not contain a TRU. The fuel assembly 1US also includes fuel rods 3A to 3E, and each mixed oxide fuel existing in the upper fuel region 6 and the lower fuel region 8 includes TRU having the composition shown in Table 1. The fuel assembly 1U is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 44 wt%.

  The core flow rate control device 33 sets a set core flow rate determined based on the ratio information (44 wt%) and the characteristics shown in FIG. 2, and controls the internal pump 26 to increase the core flow rate until the set core flow rate is reached. The core flow rate is maintained at the set core flow rate until the operation of the BWR 19 in the operation cycle is stopped.

  Since the present embodiment is also adjusted to have a set core flow rate determined by 44 wt%, which is the ratio of Pu-239 in the TRU in the new fuel assembly 1U loaded in the core, as in the first embodiment, the TRU The ratio of isotopes can be maintained. Therefore, in this embodiment, the burnup can be further increased and multiple recycling of TRU is possible. In the present embodiment, the amount of TRU in the fuel assembly 1U taken out from the core can be reduced as compared with the new fuel assembly 1US.

According to the present embodiment, in the BWR 19 that generates the same electrical output 1350 MW as the ABWR in a reactor pressure vessel of the same size as the current ABWR, the upper and lower fuel areas and the inner blanket area excluding the upper and lower blanket areas Including the core burnout region of the burnout 45 Gwd / t, the burnout burnout 52 Gwd / t of the core region higher than the light water reactor breeding reactor described in Patent Document 1, including the upper and lower blanket regions A core take-off burnup of 45 Gwd / t can be realized. In this example, the MCPR is 1.3 and the void coefficient is −2 × 10 ⁻⁵ Δk / k /% void so that the negative absolute value is larger than that in Example 1 and the ratio of the TRU isotopes is maintained. And a growth ratio of 1.01 can be realized.

  1, 1A to 1E, 1H, 1J to 1N, 1P to 1U, 1a to 1e ... Fuel assembly, 2, 2A ... Control rod, 3, 3A, 3J to 3N, 3P to 3T ... Fuel rod, 4, 4A ... Channel box, 5 ... Upper blanket region, 6 ... Upper fuel region, 7 ... Internal blanket region, 8 ... Lower fuel region, 9 ... Lower blanket region, 19 ... BWR, 20, 20A-20G ... Core, 26 ... Internal pump 30 ... Control rod drive control device, 33 ... Core flow rate control device.

Claims (17)

  In the core of a light water reactor in which a plurality of fuel assemblies containing a plurality of isotopes of super uranium nuclides are loaded, the loaded fuel assemblies having zero burnup are all the above-mentioned ultra uraniums included in the fuel assemblies. A core of a light water reactor, wherein a ratio of Pu-239 in a nuclide is in a range of 40% to 45%.   In the core of a light water reactor in which a plurality of fuel assemblies containing a plurality of isotopes of super uranium nuclides are loaded, the loaded fuel assemblies having zero burnup are included in all Pu contained in the fuel assemblies. The core of a light water reactor characterized in that the proportion of Pu-239 is in the range of 40% to 45% and the proportion of Pu-240 in all Pu is in the range of 35% to 45% .   The ratio of the plurality of isotopes of the super uranium nuclide present in the fuel assembly taken out from the core is the ratio of the plurality of isotopes present in the fuel assembly of zero burnup loaded in the core. The core of a light water reactor according to claim 1 or claim 2, wherein the core is substantially the same.   The core of a light water reactor according to claim 1, wherein a ratio of Pu-239 in the super uranium nuclide present in the fuel assembly taken out from the core is 40% or more and 45% or less.   The fuel assembly having a channel box and a plurality of fuel rods arranged in the channel box is a ratio of a cross-sectional area of fuel pellets in the fuel rods to a cross-sectional area of a unit fuel rod lattice in the channel box. The core of a light water reactor according to any one of claims 1 to 4, wherein the core is 30% or more and 55% or less.   The light water reactor according to any one of claims 1 to 5, wherein the upper blanket region, the upper fuel region, the inner blanket region, the lower fuel region, and the lower blanket region are arranged in this order from above in the axial direction. Core.   The core of the light water reactor according to any one of claims 1 to 5, wherein the upper blanket region, the fuel region, and the lower blanket region are arranged in this order from above in the axial direction.   When having a nuclear fuel material containing a plurality of isotopes of transuranium nuclides and having a burnup of zero, the proportion of Pu-239 in all the transuranium nuclides contained in the nuclear fuel material is 40% or more and 45% or less A fuel assembly characterized by being in range.   The plurality of isotopes contained in the nuclear fuel material when the ratio of the isotopes of the transuranium nuclides contained in the nuclear fuel material when it is removed from the core is zero. The fuel assembly according to claim 8, which is substantially the same as the ratio.   The fuel assembly according to claim 8, wherein a ratio of Pu-239 in the transuranium nuclide contained in the nuclear fuel material when extracted from the core is 40% or more and 45% or less.   When having a nuclear fuel material containing a plurality of isotopes of transuranium nuclides and having a burnup of zero, the proportion of Pu-239 in all the transuranium nuclides contained in the nuclear fuel material is 40% or more and 45% or less A fuel assembly characterized in that the ratio of Pu-240 in all Pu is in the range of 35% to 45%.   The ratio of the cross-sectional area of the fuel pellets in the fuel rods to the cross-sectional area of the unit fuel rod lattice in the channel box is 30% or more, which has a channel box and a plurality of fuel rods arranged in the channel box The fuel assembly according to any one of claims 8 to 11, which is 55% or less.   The upper blanket region, the upper fuel region, the inner blanket region, the lower fuel region, and the lower blanket region are arranged in the fuel effective length region from the upper side in this order in the axial direction. The fuel assembly according to item.   The fuel assembly according to any one of claims 8 to 12, wherein the upper blanket region, the fuel region, and the lower blanket region are arranged in an effective fuel length region from above in this order in the axial direction.   15. In the state of zero burnup, each of the blanket regions includes depleted uranium and does not include the transuranium nuclide, and each of the fuel regions includes the nuclear fuel material including the isotope. The fuel assembly according to 1.   In the core of a light water reactor in which a plurality of fuel assemblies containing transuranium nuclides are loaded, the ratio of Pu-239 in all Pu contained in the fuel assemblies having zero burnup loaded in the core is 40. % To 50%, and the proportion of Pu-240 in all of the Pu is in the range of 35% to 45%. A channel box and a plurality of fuel rods arranged in the channel box The fuel assembly has a ratio of the cross-sectional area of the fuel pellets in the fuel rods in the cross-sectional area of the unit fuel rod lattice in the channel box in the range of 30% to 55%, and is taken out from the core. The core of the light water reactor is characterized in that the proportion of Pu-239 in the transuranium nuclide present in the fuel assembly is 40% or more and 50% or less   The fuel assembly having a channel box and a plurality of fuel rods arranged in the channel box is a ratio of a cross-sectional area of fuel pellets in the fuel rods to a cross-sectional area of a unit fuel rod lattice in the channel box. The core of a light water reactor according to claim 16, wherein the core is in the range of 30% to 55%.

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