JP5762611B2 - Light water reactor core and fuel assembly - Google Patents

Description

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

  Actinide nuclides with many isotopes are used in various nuclear reactions such as neutron capture and nuclear decay when burned in the core in the state of being enriched with nuclear fuel material in the fuel assemblies loaded in the light water reactor core. To sequentially move between isotopes. In 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 by fast neutrons. The abundance of each isotope varies 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 limited, the fuel assemblies used in light water reactors can be used as depleted uranium, natural uranium, thorium, depleted uranium, etc., which are residues of uranium enrichment, from spent fuel assemblies of light water reactors. It is necessary to sequentially replace the extracted super uranium nuclide (hereinafter referred to as TRU) with a recycle type fuel assembly using a nuclear fuel material enriched. Depleted uranium, natural uranium, thorium, depleted uranium, TRU, and the like are referred to as nuclear fuel materials. A fuel assembly having nuclear fuel material is loaded into the core of a light water reactor. Is the newly generated U-233 regenerated as neutrons absorbed by TRU and thorium over a fairly long period of time expected to require a commercial reactor, while the amount of TRU and U-233 always increases? It is desirable to keep it almost constant.

  A technology for realizing a breeding reactor in which the amount of fissile Pu is increased or almost constant through the burning of nuclear fuel in a light water reactor that currently occupies most of commercial reactors is disclosed in Japanese Patent No. 3428150 and R. TAKEDA et al., Proc. of International Conference on Evaluation of Emerging Nuclear uel Cycle Systems. GLOBAL '95 Versailles, France, September, 1995, P.938. The breeder reactor described in Japanese Patent No. 3428150 and R.TAKEDA et al., Proc. Of Internatioal Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems.GLOBAL '95 Versailles, France, September, 1995, P.938 was realized. In a light water reactor, 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 are arranged in the core. The core of this light water reactor reduces the amount of water around the fuel rods due to the dense arrangement of the fuel rods, increases the proportion of resonance and fast energy neutrons, and lowers the height of the mixed oxide fuel region of the TRU. The blanket areas of depleted uranium are placed above and below to maintain the negative void coefficient, which is a safety standard. The core is made up of two core layers by applying the concept of a puffer core described in GA Ducat et al., “EVALUATION OF THE PARFAIT BLANKET CONCEPT FOR FAST BREEDER REACTORS”, MITNE-157, January, 1974. The growth ratio is set to 1 or more while ensuring economy and safety.

  In order to recycle TRU, reprocessing of spent fuel is essential. However, due to the fear that civilian TRUs can be diverted to weapons of mass destruction, the demand for nuclear non-proliferation has become increasingly stringent, and restrictions on TRU recycling have become stricter.

  Also, sometime in the future, there will always be a better power generation system to replace the fission reactor. At that time, TRUs from troublesome, long-lived waste from highly useful nuclear fuel equivalent to enriched uranium. It will fall to. Therefore, in order to widely disseminate light water reactors that use uranium as nuclear fuel to the world, a method for disposing of TRU remaining in spent nuclear fuel, that is, a so-called TRU extinguishing reactor that fissions TRU into fission products is prepared. This is the most important issue in nuclear development.

  JP 2008-215818 and R.TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems.GLOBAL '07 Boise, USA, September, 2007, P.1725 reprocesses spent nuclear fuel. In order to realize a multiple cycle in which recycle of the TRU obtained in this way is reused as a new nuclear fuel, a light water breeding reactor that recycles the TRU while maintaining a substantially constant ratio of each TRU isotope, A TRU extinguishing furnace that fissions TRU is proposed.

  This light water breeder reactor has a reactor core loaded with a fuel assembly in which nuclear fuel is recycled while maintaining or increasing the amount of TRU, increasing the burnup and increasing the proliferation resistance. TRU extinguishing reactors gradually aggregate TRUs while reducing the TRUs recovered by nuclear fuel reprocessing by nuclear fission in order to avoid the TRUs becoming ultra-life radioactive waste when the time for the light water reactors to finish their mission is approaching In addition, it is a nuclear reactor that fissions all TRUs except for the last core.

  R.TAKEDA et al., Proc. Of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems.GLOBAL '95 Versailles, France, September, 1995, P.938, and R.TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems. GLOBAL '07 Boise, USA, September, 2007, P.1725, in light water reactors that recycle these TRUs recovered from spent nuclear fuel, abnormal transients and accidents, etc. In order to meet the design standards of the above, long-term stable supply of energy has been realized by maintaining the TRU amount constant with a sufficient safety margin, effectively using TRU as a seed fuel, and burning all depleted uranium . In addition, when the fission reactor has finished its mission and the TRU is no longer needed, it has become possible to realize a recycle furnace that can fission all the TRU and prevent it from becoming a long-lived waste.

Japanese Patent No. 3428150 JP 2008-215818 A 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 R.TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems.GLOBAL '07 Boise, USA, September, 2007, P.1725 W.S.Yang et al., A Metal Fuel Core Concept for 1000MWt Advanced Burner Reactor GLOBAL '07 Boise, USA, September, 2007, P.52

  In addition, with the spread of light water reactors, the fear that TRUs newly produced in light water reactors will become long-lived radioactive waste and the fear that TRUs will be diverted to weapons of mass destruction are major obstacles. Before being able to serve as a seed fuel for fissioning depleted uranium for long-term stable supply of energy, the TRU from a light water reactor will be fissioned anyway and recycled with a high non-proliferation resistant isotope ratio. By repeating this, there is a strong movement to establish a technology that ultimately reduces the number of spent fuel assemblies as much as possible and to remove obstacles for the spread of light water reactors. Then, there is a new need that would be more desirable if these technologies could be implemented simply by replacing the fuel assembly in the currently operating light water reactor.

  In recent years, there has been a movement to tighten the concept of reactor safety. For example, if the core flow rate suddenly drops for some reason, it is possible to cope with accidents outside of the design criteria such as a combined event where all control rods cannot be inserted (Anticipated Transient Without Scram (ATWS)). There is a need for a core with a safety margin and a higher safety potential.

  Therefore, the inventors assumed the most severe event, a state in which the entire core is filled with steam (a state where the entire core is 100% voided), and further allowance for safety inherent in light water reactors. The improvement was examined. If the entire core is 100% voided, positive reactivity is introduced into the core. This positive reactivity input must be avoided and the margin for the safety inherent in light water reactors must be further improved.

  The objective of this invention is providing the core and fuel assembly of a light water reactor which can further increase a safety margin, without impairing the economical efficiency of a light water reactor.

  A feature of the present invention that achieves the above-described object is that a nuclear fuel material region having a nuclear fuel material containing a transuranium nuclide present in the core and having a height in the range of 20 cm to 250 cm is above the nuclear fuel material region. The neutron absorbing member is arranged.

  In a core having a nuclear fuel material region whose height is in the range of 20 cm to 250 cm, the amount of neutrons leaking from the nuclear fuel material region is large. Therefore, for some reason during the operation of the light water reactor, the entire core is assumed to be 100%. When it becomes a void state, the neutron absorbing member disposed above the nuclear fuel material region absorbs neutrons leaked from the nuclear fuel material region. For this reason, since the amount of leaked neutrons reflected back to the structural member existing outside the nuclear fuel material region and returned to the nuclear fuel material region is extremely small, if the entire core is in a 100% void state, No positive reactivity is injected into the nuclear fuel material area. This can improve the margin for the inherent safety of the light water reactor, and can thereby further increase the safety margin without damaging the economics of the light water reactor.

  The region where the nuclear fuel material is disposed in the core is the nuclear fuel material region. The height of the nuclear fuel material region is the same as the effective fuel length of the fuel assembly.

  In each fuel rod included in the fuel assembly loaded in the reactor core, the outer diameter of the plenum formed above the nuclear fuel material region containing the super-uranium nuclide of the fuel rod is 3 mm or more and the fuel rod in the nuclear fuel material region The above object can also be achieved by making it smaller than the outer diameter. The length of the plenum is in the range of 400 mm to 2500 mm.

  The outer diameter of the fuel rod plenum, which is in the range of 400 mm to 2500 mm in length, is 3 mm or more and smaller than the outer diameter of the fuel rod in the nuclear fuel material region. Structural member that delimits the plenum even when a composite event exceeding the design standard (according to the first non-design standard accident described later) occurs in which the core flow rate suddenly decreases and all control rods are not inserted into the core. The amount of leaking neutrons reflected back to the nuclear fuel material filling region is reduced. For this reason, even if an accident outside the first design standard occurs, the void coefficient becomes negative due to the inherent safety of the BWR. Therefore, the operation of the high-pressure core water injection system automatically causes the fuel rod to cool down. The safety margin of the core increases. In addition, the fuel rod health increases as the plenum volume increases. Therefore, the safety margin can be further increased without impairing the economics of the light water reactor.

  In the nuclear fuel material region in the core, in the axial direction of the core, an upper blanket region, an upper fuel region in which nuclear fuel material including super uranium nuclides exists, an inner blanket region, and a lower fuel region in which nuclear fuel materials including super uranium nuclides exist. The lower blanket region is formed in this order from above, and the ratio of fissile plutonium in the total superuranium nuclides in the lower fuel region is made larger than the ratio of fissile plutonium in the total superuranium nuclides in the upper fuel region. The above object can also be achieved.

  Since the ratio of fissile plutonium in the total fuel material in the lower fuel region is larger than the ratio of fissile plutonium in the total fuel material in the upper fuel region, the thermal margin in the lower fuel region is reduced and the upper fuel region is reduced. The thermal margin at is increased. Since the void rate in the upper fuel region is larger than the void rate in the lower fuel region, the degree of increase in the thermal margin in the upper fuel region is larger than the degree of decrease in the thermal margin in the lower fuel region. Overall, the thermal margin increases. Thus, since the thermal margin increases and the safety margin increases, the proportion of fissile plutonium in the total nuclear fuel material in the lower fuel region is greater than the proportion of fissile plutonium in the total nuclear fuel material in the upper fuel region. By making it large, the safety margin can be further increased without impairing the economics of the light water reactor.

  Loaded with multiple fuel assemblies that contain super uranium nuclides and differ in the number of recycles of super uranium nuclides, and among these fuel assemblies, multiple fuel assemblies that contain trans uranium nuclides with the least number of recycles Is located in the center of the core, and between the center and the outermost layer of the core, fuel assemblies containing super-uranium nuclides with a higher number of recycles are placed closer to the outermost layer of the core. is there.

  By configuring the core in which the fuel assemblies are arranged as described above based on the number of recycling of the ultra-uranium nuclides contained in the fuel assemblies, the number of spent fuel assemblies can be reduced. That is, among the plurality of fuel assemblies loaded in the core, the fuel assemblies containing super-uranium nuclides with a high number of recycles are arranged on the outermost layer side of the core, so that the entire core is assumed to be 100% voided. In this case, the shift of the power distribution in the radial direction toward the core center can be mitigated, and the number of spent fuel assemblies can be reduced.

  Nuclear fuel reprocessing is performed on the spent nuclear fuel contained in the spent fuel assembly removed from the nuclear reactor. By the nuclear fuel reprocessing of the spent nuclear fuel, the transuranium nuclides contained in the spent nuclear fuel are recovered, and a new fuel assembly is manufactured using the recovered transuranium nuclides. This new fuel assembly is loaded into the reactor and used in the reactor for a predetermined number of operating cycles, and then removed from the reactor as a spent fuel assembly. Nuclear fuel reprocessing is performed on the spent nuclear fuel contained in the taken-out spent fuel assembly, and the transuranium nuclide is recovered. Thus, transuranium nuclides are recycled and used. The number of recycles of the super uranium nuclide is the number of times that the super uranium nuclide is recovered from the spent nuclear fuel by nuclear fuel reprocessing and included in a new fuel assembly and used in the nuclear reactor.

  (A1) A plurality of fuel assemblies having a nuclear fuel material containing a plurality of isotopes of transuranium nuclides are loaded, have a nuclear fuel material region where the nuclear fuel material exists, and the height of the nuclear fuel material region is 20 cm to 250 cm. In the core of the light water reactor in which the neutron absorbing member is disposed above the nuclear fuel material region, a preferable configuration will be described below.

  Preferably, (A2) in A1 described above, the fuel assembly includes a lower fuel support member that supports lower end portions of the plurality of fuel rods forming the nuclear fuel material region therein, and upper end portions of the plurality of fuel rods. It is desirable to have an upper fuel support member to support, a plenum is formed above the nuclear fuel material region inside the fuel rod, and the neutron absorbing member is preferably disposed below the upper fuel support member.

  Preferably, (A3) In the above A2, it is desirable that the neutron absorbing member is disposed between the plenums of adjacent fuel rods.

  Preferably, (A4) in any one of A1 to A3, the length of the neutron absorbing member in the axial direction of the core is in the range of 20 mm to 700 mm, and the upper end of the nuclear fuel material region and the neutron absorbing member It is desirable that the distance between the lower ends is in the range of 230 mm to 500 mm.

  Preferably, (A5) in any one of the above A1 to A4, the total cross-sectional area of all the neutron absorbing members is preferably in the range of 10 to 50% of the cross-sectional area of the fuel assembly lattice. .

  Preferably, (A6) In the above A1, it is desirable that another neutron absorbing member is disposed below the nuclear fuel material region.

  Preferably, (A7) In the above A2 or A3, it is desirable that the neutron absorber filling region is formed below the nuclear fuel material region in the fuel rod.

  Preferably, (A8) In A7 described above, the length of the neutron absorber filling region in the axial direction of the core (or fuel assembly) is in the range of 10 mm to 150 mm.

  Preferably, (A9) in A7 or A8 above, the outer diameter of the fuel rod in the neutron absorber filling region is equal to or greater than the outer diameter in the nuclear fuel material region of the fuel rod, It is desirable that the distance between the outer surfaces in the neutron absorber filling region is within a range in which 1.3 mm or more is formed.

  Preferably, (A10) In the above A2 or A3, the outer diameter of the plenum portion of the fuel rod is within a range of 3 mm or more smaller than the outer diameter of the nuclear fuel material region portion of the fuel rod, The length of the core (or fuel assembly) in the axial direction is preferably in the range of 400 mm to 2500 mm.

  Preferably, (A11) in the above A2 or A3, the plenum includes a first region and a second region disposed above the first region, and an outer diameter of a portion of the first region of the fuel rod is a fuel rod The outer diameter of the portion of the nuclear fuel material region of the fuel rod is smaller than the outer diameter of the portion of the fuel rod in the second region of the fuel rod and less than the outer diameter of the portion of the nuclear fuel material region of the fuel rod. It is desirable that the neutron absorbing member is disposed between the lower end of the second region and the upper end of the nuclear fuel material region.

  Preferably, (A12) in the above A2 or A3, the plenum includes the first region and the second region disposed above the first region, and the outer diameter of the portion of the first region of the fuel rod is the fuel It is larger than the outer diameter of the portion of the second region of the rod and smaller than the outer diameter of the portion of the nuclear fuel material region of the fuel rod, and the neutron absorbing member is disposed above the upper end of the first region. desirable.

  Preferably, (A13) In any one of the above A1 to A12, it is desirable that the neutron absorbing member contains either boron or hafnium.

  Preferably, (A14) In any one of the above A7 to A9, it is desirable that the neutron absorbing material filling region contains either boron or hafnium.

Preferably, (A15) in the above A1, the nuclear fuel material region includes an upper blanket region, an upper fuel region, an inner blanket region and a lower fuel region, and an upper blanket region, an upper fuel region, an inner blanket region and a lower fuel region. The regions are arranged in this order in the axial direction of the core, and the upper fuel region and the lower fuel region contain a plurality of isotopes,
The ratio of fissile plutonium in all the upper uranium nuclides in the lower fuel region is larger than the ratio of fissile plutonium in all the upper uranium nuclides in the upper fuel region, including the fuel assemblies with zero burnup. It is desirable.

  Preferably, (A16) in the above A15, in which the fuel assemblies having zero burnup are included, the total height of the lower fuel region and the upper fuel region is in the range of 350 mm to 600 mm, and the upper fuel It is desirable that the height of the region is in the range of 1.1 to 2.1 times the height of the lower fuel region.

  Preferably, (A17) in the above A15 or A16, in a state where a fuel assembly having a burnup of zero is included, the enrichment of fissile plutonium of all the ultra-uranium nuclides in the lower fuel region and the entire super-fuel region of the upper fuel region The average enrichment of fissile plutonium in uranium nuclides is in the range of 16% to 20%, and the enrichment of fissile plutonium in all super uranium nuclides in the lower fuel region is the same as that of all super uranium nuclides in the upper fuel region. It is desirable to be in the range of 1.05 to 1.6 times the enrichment of fissile plutonium.

  Preferably, (A18) In any one of the above A15 to A17, it is desirable that the lower blanket region is disposed below the lower fuel region in the nuclear fuel material region.

  Preferably, (A19) In any one of the above A1 to A18, the ratio of plutonium-239 in all the transuranium nuclides contained in the nuclear fuel material region is in the range of 40% to 60% and 5% or more It is desirable to be within any range of less than 40%.

  Preferably, (A20) In any one of the above A1 to A19, the ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is preferably in the range of 30% to 55%.

  Preferably, a plurality of fuel rods, a lower fuel support member that supports lower end portions of the plurality of fuel rods, an upper fuel support member that supports upper end portions of the plurality of fuel rods, and a neutron absorbing member, The rod forms a nuclear fuel material region having a height in the range of 20 cm to 250 cm, in which a nuclear fuel material containing a plurality of isotopes of transuranium nuclides is present, and a plenum formed above the nuclear fuel material region. It is desirable to add the constituent elements (A2) to (A20) for the core of the light water reactor (A1) described above to the fuel assembly in which the neutron absorbing member is disposed above the nuclear fuel material region. . In the fuel assembly, the “state including the fuel assembly with zero burnup” in (A15) to (A17) is changed to “the state with zero burnup”.

  (B1) A plurality of fuel assemblies having nuclear fuel material are loaded, and an upper blanket region, an upper fuel region, an inner blanket region, and a lower fuel region are arranged in this order in the axial direction in the nuclear fuel material region where the nuclear fuel material exists. However, in the state where multiple isotopes of super uranium nuclides are contained in the upper fuel region and lower fuel region and the fuel assemblies with zero burnup are included, the fissionability of all the ultra uranium nuclides in the lower fuel region In the core of the light water reactor in which the plutonium ratio is larger than the fissile plutonium ratio in all the super uranium nuclides in the upper fuel region, a preferable configuration will be described below.

  Preferably, (B2) in B1 described above, the total of the height of the lower fuel region and the height of the upper fuel region is in the range of 350 mm to 600 mm in a state including a fuel assembly having zero burnup, and the upper fuel It is desirable that the height of the region is in the range of 1.1 to 2.1 times the height of the lower fuel region.

  Preferably, (B3) in B1 or B2 above, in a state where a fuel assembly having a burnup of zero is included, the enrichment of fissile plutonium of all the ultra-uranium nuclides in the lower fuel region and the entire super-fuel region of the upper fuel region The average enrichment of fissile plutonium in uranium nuclides is in the range of 16% to 20%, and the enrichment of fissile plutonium in all super uranium nuclides in the lower fuel region is the same as that of all super uranium nuclides in the upper fuel region. It is desirable to be in the range of 1.05 to 1.6 times the enrichment of fissile plutonium.

  Preferably, (B4) in any one of the above B1 to B3, the lower blanket region is disposed below the lower fuel region in the nuclear fuel material region.

  Preferably, (B5) In any one of the above B1 to B4, the ratio of plutonium-239 in all the transuranium nuclides included in the nuclear fuel material region is in the range of 40% to 60% and 5% or more It is desirable to be within any range of less than 40%.

  Preferably, (B6) In any one of the above B1 to B5, the ratio of the cross-sectional area of the fuel pellets to the cross-sectional area of the unit fuel rod lattice is in the range of 30% to 55%.

  Preferably, a plurality of fuel rods, a lower fuel support member that supports lower end portions of the plurality of fuel rods, an upper fuel support member that supports upper end portions of the plurality of fuel rods, and a neutron absorbing member, The rod forms a nuclear fuel material region in which a nuclear fuel material containing a plurality of isotopes of transuranium nuclides exists, and the nuclear fuel material region includes an upper blanket region, an upper fuel region, an inner blanket region, and a lower fuel region. The upper blanket region, the upper fuel region, the inner blanket region, and the lower fuel region are arranged in this order in the axial direction, and the upper fuel region and the lower fuel region contain a plurality of isotopes, and the burnup is zero. In this state, the ratio of fissile plutonium to all the super uranium nuclides in the lower fuel region is equal to the fissile plutonium in the all super uranium nuclides in the upper fuel region. The fuel assembly which is larger than the ratio, it is desirable to add the respective configuration requirements for the core of a light water reactor described above (B1) (B2) ~ (B6). In the fuel assembly, the “state including the fuel assembly having zero burnup” in (B2) and (B3) is changed to “the state having zero burnup”.

  According to the present invention, the safety margin can be further increased without impairing the economics of the light water reactor.

It is a characteristic view which shows each relationship between the distance between the upper end of a nuclear fuel material area | region and the lower end of a neutron absorption member, a void coefficient, and injection | pouring reactivity when the whole core will be in a 100% void state. Each of the length of the neutron absorbing member, the input reactivity when the entire core is in a 100% void state, and the pressure loss in the region existing between the upper end of the nuclear fuel material region and the upper end of the neutron absorbing member It is a characteristic view which shows a relationship. It is a characteristic view which shows each relationship between the height from the lower end of the nuclear fuel material area | region in a core, and the power distribution and void ratio distribution in a core axis direction. It is explanatory drawing which shows the thermal neutron flux distribution in the axial direction of a lower reflector area | region and a nuclear fuel material area | region. Characteristic diagram showing the relationship between the ratio of the cross-sectional area of the neutron absorber filling region below the nuclear fuel material region to the cross-sectional area of the fuel assembly lattice and the input reactivity when the entire core is 100% voided It is. It is a characteristic view which shows the relationship between the plenum length and the input reactivity when the entire core is in a state of 100% void. FIG. 6 is a characteristic diagram showing the relationship between the ratio of the enrichment of fissile Pu in the upper fuel region to the enrichment of fissionable Pu in the lower fuel region and the height of each region. The ratio of the enrichment of the fissile Pu in the upper fuel region to the enrichment of the fissile Pu in the lower fuel region, the void coefficient, and the input reactivity when the entire core becomes 100% void, respectively. It is a characteristic view which shows the relationship. It is explanatory drawing which shows the power distribution in the radial direction of a core. It is explanatory drawing which shows arrangement | positioning of each fuel assembly containing TRU from which the frequency | count of recycle differs in the balanced core of a light water reactor. BRIEF DESCRIPTION OF THE DRAWINGS It is a longitudinal cross-sectional view of the boiling water reactor to which the core of the light water reactor of Example 1 which is one suitable Example of this invention is applied. It is a cross-sectional view of the core shown in FIG. FIG. 13 is a cross-sectional view of the fuel assembly lattice shown in FIG. 12. FIG. 13 is an explanatory diagram showing the arrangement of fuel assemblies in the equilibrium core state of the core shown in FIG. 12. It is explanatory drawing which shows the opening distribution of the orifice in the equilibrium core shown in FIG. It is explanatory drawing which shows the enrichment distribution of the height of each area | region of the axial direction of the new fuel assembly loaded to the equilibrium core shown in FIG. 14, and the fissionable Pu in a fuel area | region. It is II sectional drawing of FIG. 13, Comprising: It is a longitudinal cross-sectional view of a fuel assembly. It is II-II sectional drawing of FIG. It is explanatory drawing which showed the enrichment distribution of the fissile Pu in the cross section of the upper fuel area | region of the fuel assembly shown in FIG. It is explanatory drawing which showed distribution of the enrichment degree of fissile Pu in the cross section of the lower fuel area | region of the fuel assembly shown in FIG. It is a cross-sectional view in the arrangement position of the neutron absorption member in other examples of a fuel assembly. It is a cross-sectional view in the arrangement position of the neutron absorption member in other examples of a fuel assembly. It is a cross-sectional view of the fuel assembly lattice in the core of the light water reactor according to embodiment 2, which is another embodiment of the present invention. FIG. 6 is an explanatory diagram showing the arrangement of fuel assemblies in an equilibrium core state of a light water reactor core according to a second embodiment. It is III-III sectional drawing of FIG. 23, Comprising: It is a longitudinal cross-sectional view of a fuel assembly. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the fuel area | region and the height of each area | region of the axial direction of the new fuel assembly loaded to the equilibrium core shown in FIG. It is IV-IV sectional drawing of FIG. It is a longitudinal cross-sectional view of the fuel assembly loaded in the core of the light water reactor in Example 3 of this invention. It is a longitudinal cross-sectional view of the fuel assembly loaded in the core of the light water reactor in Example 4 of this invention. It is VV sectional drawing of FIG. It is a longitudinal cross-sectional view of the fuel assembly loaded in the core of the light water reactor in Example 5 of this invention. It is a cross-sectional view of the fuel assembly lattice in the core of the light water reactor according to embodiment 6, which is another embodiment of the present invention. FIG. 10 is an explanatory diagram showing the arrangement of fuel assemblies in an equilibrium core state of a light water reactor core according to a sixth embodiment. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the fuel area | region and the height of each area | region of the axial direction of the new fuel assembly loaded to the equilibrium core shown in FIG. It is a transverse cross section of the core of the light water reactor of Example 7 which is other examples of the present invention. FIG. 36 is a cross-sectional view of the fuel assembly shown in FIG. 35. Explanatory drawing which shows the enrichment degree distribution of the height of each area | region of the axial direction of the new fuel assembly loaded into the equilibrium core in the core of the light water reactor of Example 8 which is another Example of this invention, and the fuel area | region. It is. FIG. 38 is an explanatory diagram showing an enrichment distribution of fissile Pu in a cross section of an upper fuel region and a lower fuel region of the fuel assembly shown in FIG. 37. It is a longitudinal cross-sectional view of the fuel assembly loaded in the core of the light water reactor in Example 9 which is another Example of this invention. Explanatory drawing which shows the enrichment degree distribution of the height of each area | region of the axial direction of the new fuel assembly loaded to the equilibrium core in the core of the light water reactor of Example 10 which is another Example of this invention, and the fuel area | region. It is. Explanatory drawing which shows the enrichment degree distribution of the height of each area | region of the axial direction of the new fuel assembly loaded in the equilibrium core in the core of the light water reactor of Example 11 which is another Example of this invention, and the fuel area | region. It is. FIG. 15 is a characteristic diagram showing the relationship between the core of each TRU recycle generation in Example 11, the weight of the current reactor discharge TRU added to the fuel assembly with zero burnup, and the TRU weight in the spent fuel assembly. The core of each recycle generation in Example 11, the weight ratio of Pu-239 contained in the TRU in the fuel assembly having zero burnup, and the weight ratio of Pu-239 contained in the TRU in the spent fuel assembly It is a characteristic view which shows each relationship of these. It is a characteristic view which shows each relationship between the core of each recycling generation in Example 11, a void coefficient, and the input reactivity when the whole core will be in a state of 100% void. The core of each recycle generation in the core of the light water reactor of Example 12, which is another embodiment of the present invention, the weight of the current reactor exhaust TRU added to the fuel assembly with zero burnup, and included in the spent fuel assembly It is a characteristic view which shows each relationship with the weight of TRU. The core of each recycle generation in Example 12, the weight ratio of Pu-239 contained in the TRU in the fuel assembly having zero burnup and the weight ratio of Pu-239 contained in the TRU in the spent fuel assembly It is a characteristic view which shows each relationship. It is a characteristic view which shows each relationship between the core of each recycling generation in Example 12, a void coefficient, and the input reactivity when the whole core will be in a state of 100% void. The core of each recycling generation in the core of the light water reactor according to the thirteenth embodiment which is another embodiment of the present invention, the weight of the current reactor discharge TRU added to the fuel assembly having zero burnup, and the TRU included in the spent fuel assembly It is a characteristic view which shows each relationship with the weight of. The core of each recycle generation in Example 13, the weight ratio of Pu-239 contained in the TRU in the fuel assembly having zero burnup and the weight ratio of Pu-239 contained in the TRU in the spent fuel assembly It is a characteristic view which shows each relationship. FIG. 11 is a characteristic diagram showing the relationship between the core of each recycling generation in Example 13, which is another example of the present invention, the void coefficient, and the input reactivity when the entire core is in a 100% void state. It is.

  The inventors have made various studies in order to realize a light water reactor that can further increase the safety margin without impairing the economics of the light water reactor. As a result, (1) the neutron absorbing member is disposed above the nuclear fuel material region that is present in the core and has the nuclear fuel material containing the super uranium nuclide and whose height is in the range of 20 cm to 250 cm. (2) The outer diameter of the plenum formed above the nuclear fuel material region and having a length in the range of 400 mm to 2500 mm is 3 mm or more and smaller than the outer diameter of the fuel rod in the nuclear fuel material region, and (3) Either of the following: the ratio of fissile plutonium (hereinafter referred to as fissile Pu) in the total nuclear fuel material in the lower fuel region is made larger than the ratio of fissile Pu in the total nuclear fuel material in the upper fuel region It has been found that the safety margin can be further increased in the core having the configuration without impairing the economics of the light water reactor.

  In addition, the inventors also examined multiple recycling of nuclear fuel materials containing transuranium nuclides. As a result, (4) among the plurality of fuel assemblies having different numbers of recycling of the ultra-uranium nuclides, the plurality of fuel assemblies including the ultra-uranium nuclides having the smallest number of recycling are arranged in the center of the core, A fuel assembly containing super-uranium nuclides with a high number of recycles between the central part and the core outermost layer region can be reduced in number by reducing the number of spent fuel assemblies by placing them on the outermost core region side. I found it. The nuclear fuel material includes a fissile material (such as U-235 and Pu-239) and a parent material (such as Th-232 and U-238).

  The safety margin is handled in the following three stages. Level 1 is a design basis accident, level 2 is a first design basis accident, and level 3 is a second design basis accident.

  Design basis accidents are events (abnormal transient changes and accidents) that are subject to safety review. In response to this design basis accident, the reactor's inherent safety and normal safety systems are activated, so that "abnormal transient changes" are designed to keep the MCPR drop within the range where the fuel rod does not burn out. It is required to do. The fuel rod is reusable. In the “accident”, it is required to design the fuel rod so that the maximum cladding tube temperature is 1200 ° C. or less, the fuel rod maintains its shape, and the fuel rod can be continuously cooled.

  The accident outside the first design standard is currently not subject to safety review, but is an event that is considered in the design of light water reactors. The accident that is considered to be the most severe of accidents outside this design standard is that all coolant supply pumps (recirculation pumps or internal pumps) that supply coolant to the core stop, and all control rods are It is a complex event where accidents that do not operate occur simultaneously. For this combined event, the emergency core cooling high-pressure core water injection pump (capacity is about 5% of the total capacity of the coolant supply pump) is activated, and the negative reactivity coefficient due to the inherent safety of the BWR Therefore, it is required to design the fuel rod so that it automatically drops to a coolable output at the flow rate of the high-pressure water injection pump.

  The accident outside the second design standard is an event that assumes that the entire core is in a 100% void state regardless of the accident scenario. For accidents outside the design standard, it is required to design so that positive reactivity is not input.

  The above (1) is a core of a light water reactor that realizes a safety margin of level 3 (accident outside the second design standard). The above (2) is the core of a light water reactor that realizes a safety margin of level 2 (accident outside the first design standard).

  Each structure of said (1)-(4) is demonstrated in detail below. Each of the above-described configurations (1) to (4) is applied to a light water reactor that recycles the super uranium nuclide recovered from spent nuclear fuel by reprocessing nuclear fuel into each fuel rod of a new fuel assembly. The

  Of these light water reactors, a core of a light water reactor having improved performance as a breeding reactor will be described. For example, a boiling water breeder reactor in which the remaining ratio of fissile Pu is 1 or more was first realized in Japanese Patent No. 3428150. In order to realize a breeder reactor in a light water reactor, it is necessary to keep the energy of neutrons in the core high. However, in a fast breeder reactor, the mass of hydrogen atoms forming water used as a coolant in a light water reactor is usually smaller than the mass of Na used as a coolant. The energy lost in a single collision is large, and the proportion of coolant per unit volume of nuclear fuel material must be reduced. When recycling is performed with nuclear fuel material in which the proportion of Pu-239 in all TRUs is greater than 60%, (a) the cooling capacity for the nuclear fuel material in the core is insufficient, (b) fuel There are inconveniences such as the burnup of the assembly is lowered and the fuel economy is impaired, and (c) the fuel rod gap constituting the fuel assembly becomes too narrow, making it difficult to manufacture the fuel assembly. When recycled with nuclear fuel material in which the proportion of Pu-239 in all TRUs is less than 40%, (d) the proportion of odd nuclides with a large fission cross-section is the proportion of even nuclides with a small fission cross-section. Compared with that, it becomes difficult to realize a residual ratio of 1 or more of fissile Pu, and (e) the core becomes large in order to maintain the critical state, and the void coefficient which is a safety index is deteriorated. , Etc. occur. Therefore, in the light water breeder reactor, the ratio of Pu-239 contained in all TRUs needs to be in the range of 40% to 60%.

  In addition, when recycling is performed with nuclear fuel material in which the proportion of Pu-240 in all TRUs is less than 35%, the above disadvantages (a), (b), and (c) occur. . When recycled with nuclear fuel material in which the proportion of Pu-240 in all TRUs is greater than 45%, inconveniences such as (d) and (e) occur. Therefore, in the light water breeder reactor, the ratio of Pu-240 contained in all TRUs needs to be in the range of 35% or more and 45% or less.

  Next, the TRU that is being considered to be disposed of as long-lived radioactive waste when it is no longer needed is used as nuclear fuel material, and finally, all TRUs other than one core TRU are fissioned. The core of the light water reactor (TRU extinguishing reactor) will be described. When the TRU is no longer necessary, the inventors reduce the number of TRUs by fissioning them, aggregate the TRUs distributed in many cores according to the amount of reduction, and finally the TRUs only in one core. I thought about making it remain. At this time, in order to prevent the TRU from becoming a long-lived radioactive waste, when the nuclear fuel material is recycled with a Pu-239 ratio of 40% or more in all TRUs, the rate at which the TRU decreases is reduced. It takes too much time to combine TRUs into one core. In addition, when recycling is performed using nuclear fuel materials in which the proportion of Pu-239 in all TRUs is less than 5%, the core becomes large and the void coefficient deteriorates. Therefore, in the TRU extinguishing furnace, the ratio of Pu-239 contained in all TRUs needs to be in the range of 5% or more and less than 40%.

  Also, in order to prevent TRU from becoming a long-lived waste, when the nuclear fuel material is recycled with the ratio of Pu-240 contained in all TRUs being 35% or less, the rate of decrease in TRU is slow. It takes too much time to consolidate them into one core. In addition, when recycling is performed using nuclear fuel material in which the proportion of Pu-240 in the total TRU is 45% or more, the core becomes large and the void coefficient deteriorates. Therefore, in the TRU extinguishing furnace, the ratio of Pu-240 contained in all TRUs needs to be in the range of 35% to 45%.

  Here, an outline of a parfait core, which is a kind of core of a light water reactor, will be described. The puffer core is a new fuel assembly to be loaded (with a burnup of zero). The lower blanket region, the lower fuel region, the inner blanket region, the upper fuel region, and the upper blanket region are arranged in this order from the lower end to the upper end. The arranged fuel assembly is used. Therefore, 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 of the nuclear fuel material area toward the upper end of the nuclear fuel material area. The lower fuel region and the upper fuel region have TRU oxide fuel (or a mixed oxide fuel of TRU oxide and uranium oxide) containing fissile material. The lower blanket region, the inner blanket region, and the upper blanket region have a uranium oxide fuel with a low content of fissile material and a high content of a parent material such as U-238.

  Each fuel rod included in the fuel assembly loaded in the core of the light water reactor forms a plenum inside. The plenum accumulates volatile fission products (FP gas) generated by fission of fissionable material contained in the nuclear fuel material filled in the fuel rod, and suppresses the increase in internal pressure of the fuel rod.

  The configuration (1) described above will be described.

  Cooling water (coolant) that cools the fuel assembly loaded in the BWR core flows into the core from below as subcooled water of about 5 ° C. to 10 ° C., saturated water, and steam while cooling the fuel assembly And a two-phase flow containing water. This cooling water is a two-phase flow having a void volume ratio of about 60% to 90% at the core outlet. Therefore, the distribution of hydrogen atoms in the core axis direction, which greatly contributes to neutron deceleration, decreases from the lower end of the core toward the upper end of the core. Such a BWR increases the void ratio at the core outlet even when the core output increases and the core flow rate decreases for some reason, and the fuel rod temperature rises and the fuel integrity may be impaired. As a result, the amount of neutron leakage to the upper part of the core increases and negative reactivity is introduced into the core, and the reactor power is automatically reduced to maintain the fuel rod soundness.

  Inventors examined the improvement plan of the further safety margin in BWR which has the above-mentioned characteristic. In this examination, the above-mentioned accident outside the second design standard was taken into consideration. An outline of this study will be described.

  The inventors have a nuclear fuel material region including a TRU obtained by nuclear fuel reprocessing at the core of a light water reactor to be studied when examining a measure for improving safety margin. The core of the BWR. In a BWR that includes a TRU and forms a nuclear fuel material region having a height in the range of 20 cm to 250 cm, the amount of neutrons leaking above and below the nuclear fuel material region increases even during operation of the BWR.

  When the height of the nuclear fuel material region is less than 20 cm, even in a core in which fuel rods are densely arranged, the amount of nuclear fuel material loaded decreases, and the fuel assembly is frequently replaced to continue the rated power operation. Need arises. For this reason, the operation rate of a nuclear power plant falls and economical efficiency is impaired. When the height of the nuclear fuel material region is greater than 250 cm, less neutrons leak from the nuclear fuel material region, and even if a neutron absorbing member is disposed above the nuclear fuel material region, the entire core is 100%. When this happens, a positive reactivity is introduced into the nuclear fuel material region. Therefore, the height of the nuclear fuel material region is set to a range of 20 cm to 250 cm.

  If the entire core is in a 100% void state for some reason, the self-control function, which is a BWR-specific safety function, is inhibited. This self-control function allows the void fraction of the two-phase flow in the reflector region that is formed above the nuclear fuel material region when the core flow rate suddenly decreases for some reason and the void rate in the nuclear fuel material region suddenly increases. This increases the neutron leakage rate from the nuclear fuel material region and decreases the neutron effective multiplication factor in the nuclear fuel material region, thereby automatically reducing the reactor power.

  A part of the neutron leaked upward from the nuclear fuel material region is reflected by a structural member that defines the fuel rod plenum (in the current fuel rod, a part of a cladding tube made of a zirconium alloy) and is reflected in the nuclear fuel material region. Returned. If the entire core becomes 100% void, the ratio of the two-phase flow existing between each plenum of adjacent fuel rods to the structural member defining the plenum decreases, and neutrons leak upward from the nuclear fuel material region. The amount of increases. This increases the amount of neutrons reflected back to the structural member defining the plenum and returned to the nuclear fuel material region. However, the increase in the amount of neutron leakage from the nuclear fuel material region is larger than the increase in the infinite neutron multiplication factor when the void rate in the nuclear fuel material region becomes the above-mentioned 100% void state from the value at the rated power operation. Since it is small, positive reactivity is injected into the core, specifically the nuclear fuel material region.

As a result of various studies, the inventors have arranged a neutron absorber (for example, B ₄ C and Hf) above the nuclear fuel material region, so that neutrons leaked upward from the nuclear fuel material region are caused by the neutron absorbing member. Since it was absorbed, it was newly confirmed that no positive reactivity entered the core even when the entire core was 100% voided. Since the arrangement of the neutron absorber as described above can prevent the positive reactivity from being input, the margin for the inherent safety of the BWR is improved, and as a result, the safety margin of the BWR is improved. Accordingly, the inventors have found a new finding that by applying the configuration of (1) to the core of a light water reactor, the safety margin can be improved even in multiple accidents while maintaining the TRU breeding ratio.

  The examination results described above will be described in detail.

  The results of studies conducted by the inventors on the core of a light water breeder reactor will be described below. The light water breeding reactor to be studied is, for example, a BWR core having a breeding ratio of 1.01 having an electric output of 1350 MW, 720 fuel assemblies loaded in the core, and 271 fuel rods per fuel assembly. is there. Each fuel assembly has nuclear fuel material including TRU obtained by nuclear fuel reprocessing, and the proportion of Pu-239 in all TRUs included at the time of zero burnup is 40% to 60%. The value is within the range of%.

Based on the knowledge about the configuration of (1) above, the inventors set the neutron absorbing member (including B ₄ C, Hf, etc.) in the fuel assembly to the position of the plenum in the axial direction of the fuel assembly. (See, for example, FIGS. 18, 21, and 22). In each fuel rod, the plenum is located above the upper end of the effective fuel length (nuclear fuel material region). For this reason, the neutron absorber is located above the nuclear fuel material region and below the lower end of the upper fuel support member (for example, the upper tie plate) of the fuel assembly that holds the upper end of each fuel rod. Placed in. In the case of using B ₄ C, for example, a neutron absorbing member configured by filling B ₄ C in a sealed container is disposed between the plenums. When Hf is used, the metal Hf is formed into a plate shape or a rod shape, for example, and disposed between the plenums as a neutron absorbing member.

  During the operation of the BWR, a gas-liquid two-phase flow flows between the fuel rods at the plenum position in the fuel assembly. Even when the BWR is stopped, cooling water exists in the core. A two-phase flow or cooling water existing between the fuel rods above the nuclear fuel material region functions as a neutron reflector. For this reason, it can be said that the neutron absorber is disposed in the reflector above the nuclear fuel material region. A region where the two-phase flow or cooling water exists above the nuclear fuel material region is referred to as a reflector region.

  The inventors examined the arrangement of the neutron absorbing member between the fuel rods at the plenum position. FIG. 1 shows that the entire core is 100% void by the distance between the upper end of the nuclear fuel material region and the lower end of the neutron absorbing member disposed between the plenums (the distance between the nuclear fuel material region and the neutron absorbing member). The changes in the reactivity of the reactor core and the void coefficient are shown. Characteristic A shows the relationship between the distance between the nuclear fuel material region and the neutron absorbing member and the void coefficient, and property B shows the relationship between the distance and the input reactivity. The distance between the nuclear fuel material region and the neutron absorbing member is a distance in the core axis direction. Characteristics A and B are characteristics obtained for the core of a light water breeding reactor loaded with a fuel assembly in which a neutron absorbing member having a length of 500 mm is arranged adjacent to each fuel rod as shown in FIG.

  If the lower end of the neutron absorbing member is too close to the upper end of the nuclear fuel material region, the effect of reflecting neutrons to the nuclear fuel material region will be reduced during operation of the BWR due to the influence of the neutron absorbing member. As a result, the effective neutron multiplication factor of the nuclear fuel material region is reduced, and it is necessary to increase the height of the nuclear fuel material region in order to compensate for this reduction. If the entire core becomes 100% voided, the core Increases the reactivity to the input. When the distance between the nuclear fuel material region and the neutron absorbing member is shorter than 230 mm, if the entire core is in a 100% void state, positive reactivity is introduced into the core. For this reason, if the entire core is in a 100% void state, the distance between the nuclear fuel material region and the neutron absorbing member must be 230 mm or more in order to avoid the introduction of positive reactivity in the core.

Further, according to the characteristic B, when the distance between the nuclear fuel material region and the neutron absorbing member is shortened, the volume of the two-phase flow region (reflector region) between the nuclear fuel material region and the neutron absorbing member is reduced, The change in effective neutron multiplication factor due to the change in core void fraction becomes smaller and the void coefficient worsens. If the distance between the nuclear fuel material region and the neutron absorbing member becomes too long, the influence of the reflector region on the nuclear fuel material region becomes small. For this reason, the probability that the neutron leaked to the reflector region will return to the nuclear fuel material region again increases, and the void coefficient deteriorates. When the distance between the nuclear fuel material region and the neutron absorbing member exceeds 500 mm, the void coefficient becomes −1 × 10 ⁻⁴ % Δk / k /% void or less, and there is a problem in the core transient characteristics (for example, satisfying the constraints of MCPR) Event that cannot be performed). From the above, it is preferable that the distance between the nuclear fuel material region and the neutron absorbing member in the core axis direction is in the range of 230 mm to 500 mm.

  Even during the rated operation of the BWR, the neutron absorbing member disposed in the reflector region absorbs neutrons leaking upward from the nuclear fuel material region. If the neutron absorbing member is too close to the upper end of the nuclear fuel material region, the amount of neutrons returned from the reflector region to the nuclear fuel material region is reduced during the rated operation of the reactor due to the neutron absorbing function of the neutron absorbing member. This reduces the reactor power at the upper end of the nuclear fuel material region. Such a problem does not occur when the distance between the nuclear fuel material region and the neutron absorbing member is set to 230 mm or more.

  Next, the inventors examined the length of the neutron absorbing member in the axial direction of the fuel assembly. FIG. 2 shows the relationship between the length of the neutron absorbing member, the input reactivity, and the pressure loss between the upper end of the nuclear fuel material region and the upper end of the neutron absorbing member. Characteristic C shows the change in the input reactivity depending on the length of the neutron absorbing member. In this input reactivity, the characteristic D indicates the change in pressure loss between the upper end of the nuclear fuel material region and the upper end of the neutron absorbing member due to the length of the neutron absorbing member. Characteristics C and D are characteristics when the distance between the upper end of the nuclear fuel material region and the upper end of the neutron absorbing member is 300 mm.

  When the length of the neutron absorbing member is less than 20 mm, if the entire core is in a 100% void state, positive reactivity is introduced into the core (see characteristic C in FIG. 2). For this reason, the length of a neutron absorption member shall be 20 mm or more. When the length of the neutron absorbing member exceeds 700 mm, the increase in pressure loss between the nuclear fuel material region and the neutron absorbing member generated by the flow of the two-phase flow is 20% or more of the pressure loss of the entire core. become. The influence of the increase in pressure loss between the nuclear fuel material region and the neutron absorbing member on the core characteristics cannot be ignored. For this reason, the length of the neutron absorbing member is set to 700 mm or less.

  Therefore, Preferably, the length of a neutron absorption member is set to the range of 20 mm-700 mm.

  The neutron absorbing member may be disposed below the nuclear fuel material region. When the neutron absorbing member is disposed below the nuclear fuel material region, if the entire core is in a 100% void state, neutrons leaking downward from the nuclear fuel material region can be absorbed by the neutron absorbing member. . For this reason, by disposing the neutron absorbing member below the nuclear fuel material region, if the entire core is in a 100% void state, the amount of neutrons leaking below the nuclear fuel material region and returning to the nuclear fuel material region Is very low. This also prevents positive reactivity from being introduced into the core when the entire core is in a 100% void state.

In order to dispose the neutron absorbing member below the nuclear fuel material region, the neutron absorbing member may be disposed at the lower end of each fuel rod included in the fuel assembly. Specifically, a neutron absorbing material (for example, B ₄ C and Hf) is filled in the lower end portion of the cladding tube of the fuel rod (see, for example, FIGS. 17 and 27). A plurality of fuel pellets containing TRU are filled into the cladding tube above the neutron absorber filling region. The outer diameter of the lower end portion of the fuel rod where the neutron absorber filling region exists is the same as the outer diameter of the fuel rod above the neutron absorber filling region. The length of the neutron absorber filling region is in the range of 10 mm to 150 mm. In order to minimize the amount of neutrons leaking below the nuclear fuel material region and returning to the nuclear fuel material region, it is preferable that the upper end of the neutron absorber filling region is in contact with the lower end of the nuclear fuel material region. However, even if the upper end of the neutron absorber filling region and the lower end of the nuclear fuel material region are separated by a maximum of 5 mm, even if the entire core is in a 100% void state, it is avoided that positive reactivity is introduced into the core. it can. The upper end of the neutron absorber filling region arranged below the nuclear fuel material region is separated downward from the lower end of the nuclear fuel material region, so that the neutron is absorbed in the neutron absorber filling region to the output at the lower end of the nuclear fuel material region. The adverse effect of can be reduced. The neutron absorber filling region is provided separately from the control rod inserted into the core from below.

  In BWR, control rods are inserted between the fuel assemblies loaded in the core from below the core. For this reason, in addition to the control rod, a neutron absorber is disposed below the nuclear fuel material region. Arranging the neutron absorbing member below the nuclear fuel material region can remarkably suppress neutrons leaking below the nuclear fuel material region from returning to the nuclear fuel material region.

  Next, the core of the TRU extinguishing reactor described in R.TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems.GLOBAL '07 Boise, USA, September, 2007, P.1725 The examination results of the inventors will be described. The core of the TRU extinguishing reactor to be studied is, for example, a BWR core in which 720 fuel assemblies having an electrical output of 1350 MW and 397 fuel rods per fuel assembly are loaded.

  When TRU recycling is repeated for the purpose of reducing TRU, that is, Pu-239 of nuclear fuel material containing TRU obtained by nuclear fuel reprocessing and occupied in all TRUs included at the time of zero burnup When loading fuel assemblies with a ratio in the range of 5% or more and less than 40% to the core in every operation cycle, the reaction rate ratio due to fast neutron flux is larger than that of the core with a growth ratio of 1. In some cases, the neutron absorbing member disposed above the nuclear fuel material region must be thicker than the core near the growth ratio of 1. In this case, as a matter of course, the outer diameter of the plenum portion of the fuel rod is smaller than the outer diameter of the fuel rod at the portion below the plenum. However, in the core of the TRU extinguishing reactor, preferably, the distance between the nuclear fuel material region and the neutron absorbing member is set in the range of 230 mm to 500 mm, and the length of the neutron absorbing member is set in the range of 20 mm to 700 mm.

Also in the core of the TRU extinguishing reactor, a neutron absorbing member (B ₄ C or Hf) may be disposed below the nuclear fuel material region as in the light water breeder reactor. In the TRU extinguishing furnace, the control rod is inserted into the core from below. As in the light water breeder reactor, when the neutron absorbing member is disposed at the lower end of the fuel rod below the nuclear fuel material region, in order to obtain the necessary and sufficient effect by the neutron absorbing member, the neutron absorber filling region is provided inside. It may be desirable to make the outer diameter of the lower end of the formed fuel rod larger than the outer diameter of the fuel rod above the neutron absorber filling region. However, considering the pressure loss of the fuel assembly, the width of the gap formed between the fuel rods should not be less than 1.3 mm at the lower end of the fuel rods where the neutron absorber filling region is formed.

  Similar to the light water breeder reactor, the length of the neutron absorbing material filling region is set in the range of 10 mm to 150 mm. The upper end of the neutron absorber filling region and the lower end of the nuclear fuel material region can be separated by a maximum of 5 mm.

  The inventors examined various characteristics of the core of a light water reactor when the entire core is in a 100% void state for some reason. First, the axial power distribution and the axial void ratio distribution in the axial direction of the nuclear fuel material region will be described with reference to FIG. In FIG. 3, characteristic E shows the average power distribution in the core axis direction during the rated power operation of the reactor. Characteristic F shows the average power distribution in the core axis direction when the entire core is in a state of 100% void for some reason during rated power operation, which is one of the most severe complex events as an accident outside the design criteria. Yes. A characteristic G represents an average void ratio distribution in the core axis direction corresponding to the power distribution of the characteristic E. Each characteristic shown in FIG. 3 was obtained for the TRU extinguishing furnace, but each characteristic showing the same tendency as each characteristic shown in FIG. 3 could be obtained also in the light water breeding reactor.

  Based on FIG. 3, if the entire core is in a 100% void state, the power distribution in the axial direction of the core shifts to the lower end side of the nuclear fuel material region, and the lower reflector region exists below the nuclear fuel material region. However, it turns out that it plays the role of the dump tank of surplus neutrons generated at the time of the accident. The lower reflector region is located below the nuclear fuel material region, below the lower end of the nuclear fuel material region, and between the fuel rods in the lower tie plate and below the lower tie plate fuel holder. This is the area where the cooling water below exists.

  A state in which all the control rods are completely extracted from the nuclear fuel material region, that is, the upper end of the neutron absorber filling region of the control rod is arranged near the lower end of this region below the nuclear fuel material region, and there is no lower blanket region In the nuclear fuel material region, the thermal neutron flux distribution in the core axis direction when the entire core is in a 100% void state for some reason is represented by the characteristic H in FIG. The characteristic J in FIG. 4 is that the neutron absorber is disposed in a portion below the fission material region in the fuel rod, and the outer diameter of the fuel rod in the portion filled with the neutron absorber is larger than the portion filled with the neutron absorber. The thermal neutron flux distribution in the core axis direction in the case where the outer diameter of the upper fuel rod is the same is shown. Characteristic K in FIG. 4 is that the neutron absorber is disposed in the fuel rod below the fission material region, and the outer diameter of the fuel rod in the portion filled with the neutron absorber is larger than the portion filled with the neutron absorber. The distribution of thermal neutron flux in the core axis direction when the outer diameter of the upper fuel rod is larger is shown. The characteristics J and K are the control rod drawing state and the nuclear fuel material region in which the characteristic H is obtained. The thermal neutron flux in the axial direction of the core when the entire core becomes 100% void for some reason. Represents the distribution. The thermal neutron flux distribution of characteristics J and K is significantly lower than that of characteristic H. By placing a neutron absorber below the nuclear fuel material region, the entire core is 100% voided for some reason. Even in this case, the lower reflector region located below the nuclear fuel material region serves as a neutron dump tank. For this reason, generation | occurrence | production of a surplus reactivity can be prevented beforehand.

  The ratio of the cross-sectional area of the neutron absorber filling region below the nuclear fuel material region to the cross-sectional area of the fuel assembly lattice, and the reactivity that is input to the core when the entire core becomes 100% void This relationship is shown in FIG. When the ratio is 35% or more, even if the entire core is in a 100% void state, no positive reactivity is introduced into the core. For this reason, the ratio of Pu-239 in all TRUs that have nuclear fuel materials including TRUs obtained by nuclear fuel reprocessing and are included at the time of zero burnup is within the range of 5% or more and less than 40%. The safety margin can be improved in the core of a light water reactor loaded with a certain fuel assembly.

  By arranging a neutron absorbing member above a nuclear fuel material region that has a nuclear fuel material containing a super-uranium nuclide in the core and has a height in the range of 20 cm to 250 cm, the economy of a light water reactor The safety margin can be further increased without impairing the performance. Multiple recycling of TRUs can also be continued. Preferably, the distance between the nuclear fuel material region and the neutron absorbing member in the core axis direction is in the range of 230 mm to 500 mm, and the length of the neutron absorbing member is in the range of 20 mm to 700 mm.

  A neutron absorbing member is disposed above the nuclear fuel material region, and further, a neutron absorber filling region is disposed below the nuclear fuel material region, so that if the entire core becomes 100% voided, it is introduced into the core. The reactivity can be further negative.

  In the light water breeding reactor and TRU extinguishing reactor, when the ratio of the cross-sectional area of the fuel pellets (filled in the fuel rods) to the cross-sectional area of the unit fuel rod lattice in the channel box exceeds 55%, the sensitivity between the fuel rods is 1 mm. As a result, the assembly of the fuel assembly becomes extremely difficult. 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.

  Furthermore, the inventors examined how many neutron absorbing members should be arranged per fuel assembly lattice above the nuclear fuel material region. The neutron absorbing member is disposed adjacent to the plenum formed in the fuel rod between the nuclear fuel material region and the upper fuel support member (eg, upper tie plate). In order to prevent positive reactivity from being introduced into the core when the entire core becomes 100% void for some reason, the cross-sectional area of all neutron absorbing members arranged above the nuclear fuel material region Must be 10% or more of the cross sectional area of the fuel assembly lattice. The fuel assembly lattice includes an annular region that surrounds one fuel assembly disposed in the core with a width that is half of a gap (water gap in BWR) formed between adjacent fuel assemblies, and This is a region including the cross section of the one fuel assembly. The cross sectional area of the fuel assembly lattice is the sum of the cross sectional area of the annular region and the cross sectional area of the fuel assembly.

  During the operation of the light water reactor, it is necessary to flow a two-phase flow with a predetermined flow rate between the neutron absorbing members disposed between the plenum portions of the adjacent fuel rods. The sum should be no more than 50% of the cross-sectional area of the fuel assembly lattice.

  Based on the above examination results, it is preferable to set the total cross-sectional area of all neutron absorbing members within a range of 10 to 50% of the cross-sectional area of the fuel assembly lattice.

  Even below the nuclear fuel material region, excluding the control rod cross-sectional area, the total cross-sectional area of all neutron absorber filling regions formed in the fuel assembly is in the range of 10 to 50% of the cross-sectional area of the fuel assembly lattice. It is preferable to set within.

  It is also possible to add the configuration (2) to the configuration (1). That is, a neutron absorbing member is disposed above a nuclear fuel material region that is present in the core and has a nuclear fuel material containing a super-uranium nuclide and has a height in the range of 20 cm to 250 cm. The outer diameter of the plenum existing in the range of 400 mm to 2500 mm formed above is made 3 mm or more smaller than the outer diameter of the fuel rod in the nuclear fuel material region. As a result, even if the core is in a 100% void state, it is possible to prevent positive reactivity from being introduced into the nuclear fuel material region, and the soundness of the fuel rod is increased. Since the outer diameter of the plenum is 3 mm or more and smaller than the outer diameter of the fuel rod in the nuclear fuel material region, the amount of the neutron absorbing member disposed between the plenums of adjacent fuel rods (for example, the thickness of the neutron absorbing member) ) Can be increased. As a result, if the core is in a 100% void state, the reactivity introduced into the nuclear fuel material region can be made more negative.

  A pressurized water reactor (PWR) in which cluster control rods are inserted from above the core into a plurality of guide tubes provided in a fuel assembly loaded in the core, and a high speed at which control rods are inserted into the core from above Also in the breeder reactor (FBR), a nuclear fuel material region including TRU may be formed in the core, and neutron absorbing members may be disposed above and below the nuclear fuel material region.

  The configuration (2) described above will be described.

  A fuel rod included in a fuel assembly loaded in a core of a light water reactor, for example, a BWR core, contains therein a plurality of fuel pellets including TRU. Even when the release rate of volatile fission products from this fuel pellet is higher than that of uranium oxide pellets, TRU recycling continues while maintaining the safety potential of the BWR while ensuring the integrity of the fuel rods. For this purpose, it is necessary to increase the volume of the plenum formed in the fuel rod and to keep the void coefficient within a predetermined range. Also, commercial reactors that are put into practical use require higher burnup of fuel assemblies that increase the amount of volatile fission products generated from the viewpoint of fuel economy, so the volume of the plenum in the fuel rod is increased. There must be.

  When the volume of the plenum provided at the upper part of the fuel rod is increased, the self-control function which is a safety function unique to the BWR is hindered. When the volume of the plenum is increased, the self-control function is hindered when an accident outside the first design standard occurs for some reason.

  For some reason, accidents outside the first design standard, that is, all coolant supply pumps (recirculation system pumps or internal pumps) that supply coolant to the core stop, and all control rods do not operate at the same time. Even when a composite event occurs, it is necessary to improve the safety margin of the BWR core.

  The inventors examined a safety margin improvement measure that can improve the safety margin of the core of a light water reactor when an accident outside the first design standard occurs even without using the configuration of (1). The core of the light water reactor used for this examination is a BWR core having a nuclear fuel material region including TRUs having a height in the range of 20 cm to 250 cm and obtained by nuclear fuel reprocessing. When an accident outside the first design standard occurs, the high pressure core water injection system of the emergency core cooling system can be operated.

  As a result of the study by the inventors, even when the volume of the plenum in the fuel rod is increased, the outer diameter of the plenum portion of the fuel rod is made larger than the outer diameter of the fuel rod in the nuclear fuel material filling region below the plenum portion. As a result of the reduction, the inventors discovered a new finding that when an accident outside the first design standard occurs, the reactivity introduced into the core is reduced. Based on this new knowledge, the inventors set the outer diameter of the plenum formed above the nuclear fuel material region in the range of 400 mm to 2500 mm to be 3 mm or more and the fuel rod in the nuclear fuel material region. The conclusion was reached that the outer diameter should be smaller than the outer diameter.

  By applying the configuration of (2) in this way, the amount of leaked neutrons that are reflected back to the structural member defining the plenum and returned to the nuclear fuel material filling region even when a composite event outside the first design standard occurs. And increasing the plenum volume increases fuel rod health. Therefore, the safety margin can be further increased without impairing the economics of the light water reactor.

  The examination results described above will be described in detail. The inventors examined the core of a light water reactor having a nuclear fuel material region including TRU obtained by nuclear fuel reprocessing. The multiplication ratio of the nuclear fissionable Pu is 1.01.

  FIG. 6 shows the result obtained by the examination, and shows the change in the input reactivity with respect to the plenum length. This input reactivity is the input reactivity when the entire core is in a void state.

  The hydrogen scattering cross-section is relatively large in the energy region of 500 keV or less, but rapidly decreases as it approaches 1 Mev. Therefore, fast neutrons of 1 MeV or more penetrate from the nuclear fuel material region to the reflector region by deep penetration. On the other hand, since the mass of hydrogen atoms constituting the two-phase flow passing through the fuel assembly is almost the same as that of neutrons, the neutrons lose a lot of energy by one collision with hydrogen atoms. For this reason, when the upper reflector existing above the nuclear fuel material region is composed only of a two-phase flow containing water and this vapor, the fast neutrons once leaked from the nuclear fuel material region to the upper reflector again become nuclear fuel material. The probability of returning to the region is small. However, if the component that defines the plenum in the fuel rod (eg, Zircaloy tube) is present in the upper reflector region, the zirconium atom mass in the Zircaloy tube is greater than the neutron mass. The energy that neutrons lose due to a single collision with a zirconium atom is extremely small. For this reason, while the neutrons collide with the zirconium atom several times, some neutrons return to the nuclear fuel material region.

  In FIG. 6, characteristic L is relative to the plenum length when the outer diameter of the portion of the plenum formed in the fuel rod is the same as the outer diameter of the portion of the nuclear fuel material filling region below the plenum portion of the fuel rod. The change in input reactivity is shown. Characteristic M indicates the input reactivity with respect to the plenum length when the outer diameter of the portion of the plenum formed in the fuel rod is smaller than the outer diameter of the portion of the nuclear fuel material filling region below the plenum portion of the fuel rod. Shows changes. Specifically, the cross-sectional area of the plenum portion is half the cross-sectional area of the nuclear fuel material filling region portion of the fuel rod.

  For some reason, the total number of coolant supply pumps that supply coolant to the reactor core stops, and a combined event (all accidents outside the first design standard) that prevents all control rods from operating causes an increase in output and fuel rods. As the temperature of the fuel pellets increases, the rate of release of volatile fission products from the fuel pellets increases. Furthermore, the internal pressure of the fuel rod cladding tube increases, the gap between the cladding tube and fuel pellets widens, the heat transfer rate from the fuel pellets to the cladding tube decreases, and the temperature of the fuel pellets further increases. The occurrence of the above composite event causes such a positive feedback condition. However, by increasing the plenum length by increasing the plenum length, it is possible to prevent the occurrence of such a positive feedback state and improve the fuel rod soundness.

  As shown by the characteristic M in FIG. 6, when the outer diameter of the portion of the plenum formed in the fuel rod is smaller than the outer diameter of the portion of the nuclear fuel material filling region below the plenum portion of the fuel rod, When the entire core is in a 100% void state, the reactivity introduced into the nuclear fuel material region is less than $ 1. For this reason, if the outer diameter of the plenum portion in the fuel rod is smaller than the outer diameter of the fuel rod in the nuclear fuel material filling region, even if a composite event of an accident outside the first design standard occurs By the operation of the high-pressure core water injection system, the output is automatically reduced to the output at which the fuel rod can be cooled with the flow rate of the cooling water injected into the core, and the safety of the BWR is maintained. Therefore, the safety margin of the BWR core can be improved by making the outer diameter of the plenum portion smaller than the outer diameter of the nuclear fuel material filling region.

  In the fuel rod, when the outer diameter of the plenum portion is the same as the outer diameter of the nuclear fuel material filling region portion, the length of the plenum is about 200 mm, from the outer diameter of the nuclear fuel material filling region portion below the plenum portion. However, if the length of the plenum is about 200 to 300 mm, it is possible to avoid the introduction of positive reactivity even when the entire core is in a state of 100% void.

  The configuration (3) described above will be described.

  It has been proposed to multiplex-recycle TRUs obtained by nuclear fuel reprocessing (Japanese Patent Laid-Open No. 2008-215818 and R. TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems. Boise, USA, September, 2007, P.1725). In order to realize multiple recycling of TRU, it is necessary to use nuclear fuel material recovered from spent nuclear fuel generated from various light water reactors (BWR and PWR).

  Even in light water reactors, BWR and PWR have different neutron energy spectra when the fissile material contained in the nuclear fuel material existing in the core burns. In addition, there are various types of spent fuel assemblies that have just been removed from the reactor core and spent fuel assemblies that have been stored for a long time in the fuel storage pool. is there. The spent nuclear fuel contained in the spent fuel assembly stored in the fuel storage pool is also different in the nuclear decay of the isotope due to the difference in the storage period of the spent fuel assembly, and the composition of the contained TRU is different. Is different.

  From such various spent nuclear fuels, a plurality of new fuel assemblies manufactured using nuclear fuel materials including TRU recovered by nuclear fuel reprocessing must be loaded into the core of one light water reactor. Due to the difference in the composition of the TRUs in the recovered nuclear fuel material containing TRU, there is a risk that the variation in output of each fuel assembly manufactured and loaded in the core will increase, and the thermal margin of the core will be reduced There is. For this reason, it is desired to increase the thermal margin of the core of the light water reactor.

  The inventors have made various studies in order to realize a core of a light water reactor with an increased thermal margin. As a result of this examination, the inventors determined that the ratio of fissile Pu in the total nuclear fuel material in the lower fuel region formed in the nuclear fuel material region is the fission in the total fuel material in the upper fuel region formed in the nuclear fuel material region. It has been found that the thermal margin of the core of the light water reactor can be increased without impairing the economics of the light water reactor by making it larger than the ratio of the characteristic Pu. With the configuration (3), it is possible to increase the thermal margin of the fuel rod power density, the fuel rod center temperature, the MCPR, and the like. Furthermore, multiple recycling of TRUs can be realized.

  The configuration of (3) can be realized by a parfait core in which a lower blanket region, a lower fuel region, an inner blanket region, an upper fuel region, and an upper blanket region are formed in order from the lower end of the nuclear fuel material region toward the upper end of the nuclear fuel material region. .

  In order to increase the thermal margin, it is desirable to increase the height of the nuclear fuel material region in the core, that is, to increase the total length of the fuel rod in the axial direction. In the BWR, the density of cooling water, which is a neutron moderator, decreases in the reactor core from the lower end of the nuclear fuel material region toward the upper end of the nuclear fuel material region. Therefore, the nuclear fuel material in the lower fuel region in the lower fuel region where the enrichment of the fissile Pu in the whole nuclear fuel material in the upper fuel region is lowered to increase the height of the upper fuel region and the cooling water density is higher than that in the upper fuel region. Increase the enrichment of fissionable Pu in the neutron and improve the utilization rate of neutrons. The core growth ratio and void coefficient do not deteriorate.

  Increasing the enrichment of fissile Pu in the whole fuel material in the lower fuel region will reduce the height of the lower fuel region. However, the increase in the height of the upper fuel region is larger than the decrease in the height of the lower fuel region, and as a result, the height of the nuclear fuel material region increases.

  The ratio of fissile Pu in the total nuclear fuel material in the lower fuel region formed in the nuclear fuel material region is larger than the ratio of the fissile Pu in the total nuclear fuel material in the upper fuel region formed in the nuclear fuel material region. The core of a light water reactor with an increased margin is the fuel in the core depending on the capacity of coolant that can be supplied from the high-pressure core injection system of the emergency core cooling system, even if an accident outside the first design occurs for some reason. The output can be automatically reduced to an output that can cool the assembly. Further, the core of such a light water reactor is not charged with a positive reactivity even if the entire core is in a 100% void state for some reason.

  In the core of the light water reactor having the configuration of (3), the thermal margin is increased, so that the safety margin can be further increased without impairing the economics of the light water reactor.

  The inventors have devised the following configurations (I) and (II) as specific configurations for realizing the configuration (3). In the configuration (I), the sum of the height of the lower fuel region and the height of the upper fuel region in the nuclear fuel material region is within a range of 350 mm to 600 mm, and the height of the upper fuel region is set to the height of the lower fuel region. Within the range of 1.1 to 2.1 times. The configuration of (II) is such that the average of the enrichment degree of fissile Pu in the whole nuclear fuel material in the lower fuel region and the enrichment degree of fissile Pu in the whole nuclear fuel material in the upper fuel region is 14 to 22%. And the enrichment of the fissile Pu in the whole nuclear fuel material in the lower fuel region is within a range of 1.05 to 1.6 times the enrichment of the fissile Pu in the whole nuclear fuel material in the upper fuel region. To do. With either of the configurations (I) and (II), the thermal margin can be further increased without impairing the economics of the light water reactor.

  The examination results described above will be described in detail.

  The above discussion is based on a light water breeder reactor core, for example, a BWR with a growth ratio of 1.01 having an electrical power of 1350 MW, 720 fuel assemblies loaded into the core and 271 fuel rods per fuel assembly. This was done for the core.

  In light water breeder reactors, under the effective neutron multiplication factor of 1, which is a critical constraint, it is possible to satisfy a well-balanced negative void coefficient, which is one of important indicators for breeding ratio, thermal margin and safety. is important.

  As a result of the investigations by the inventors conducted on the core of the BWR to be examined above, the enrichment of the fissile Pu in the whole nuclear fuel material in the upper fuel region is lowered, and the fissile Pu in the whole nuclear fuel material in the lower fuel region is reduced. It has been found that by increasing the enrichment, as described above, the thermal margin of the core can be increased without deteriorating the growth ratio and void coefficient.

  In general, when the fissile Pu enrichment is increased, the neutron spectrum in the fuel region where the fissile material is present is shifted to a higher energy side, and the number of neutrons generated when the TRU undergoes fission increases. At the same time, fast fission of the parent substance, such as U-238, increases. For this reason, the number of neutrons leaking from the fuel region to the blanket region increases, contributing to an increase in the growth ratio. However, since the height of the fuel region necessary to keep the core critical is reduced, the overall length of the fuel rod is shortened, and the thermal margin is reduced. On the other hand, the absolute value of the negative void coefficient of the core increases, and the safety margin increases.

  However, if the enrichment of fissile Pu in the whole nuclear fuel material in the upper fuel region is lowered and the enrichment of the fissile Pu in the whole nuclear fuel material in the lower fuel region is increased, as described above, the nuclear fuel material region The height of can be increased. For this reason, the thermal margin of the core increases.

  The inventors examined a reactor core in which the ratio of fissile Pu in the total nuclear fuel material in the lower fuel region is larger than the ratio of fissile Pu in the total nuclear fuel material in the upper fuel region formed in the nuclear fuel material region. . FIG. 7 shows one of the examination results.

  In the new fuel assembly loaded in the equilibrium core of the light water breeder reactor, the ratio of the enrichment of the fissile Pu in the upper fuel region to the enrichment of the fissionable Pu in the lower fuel region (hereinafter simply referred to as the fissionable Pu enrichment). The change in the height of the upper fuel region, the lower fuel region, and the nuclear fuel material region when the ratio was changed was investigated. FIG. 7 shows the relationship between the ratio of enrichment of fissile Pu and the height of each region. The characteristic P shows the change in the height of the upper fuel region according to the ratio of the enrichment of fissile Pu. The characteristic Q shows the change in the height of the lower fuel region depending on the ratio of the enrichment of fissile Pu. The characteristic R shows the change in the height of the nuclear fuel material region depending on the ratio of the enrichment of fissile Pu.

  Considering the criticality of the core and the flattening of the power distribution in the core axis direction, the enrichment of the fissile Pu in all TRUs in the upper fuel region is 17%, and the fissile Pu in all TRUs in the lower fuel region is When the enrichment is 19%, the height of the upper fuel region is about 1.1 times the height of the lower fuel region. When the enrichment of the fissile Pu in the total nuclear fuel material in the upper fuel region is 14% and the enrichment of the fissile Pu in the total nuclear fuel material in the lower fuel region is 22%, the height of the upper fuel region is the lower fuel region. About 2.1 times the height of Each characteristic shown in FIG. 11 is evaluated when the average of the enrichment degree of fissile Pu in the whole nuclear fuel material in the upper fuel region and the enrichment degree of fissile Pu in the whole nuclear fuel material in the lower fuel region is 18%. It is a result. Even if the average of these enrichments is varied between 16% and 20%, the height of the upper fuel region and the height of the lower fuel region relative to the ratio of fissile Pu enrichment is determined by the average of the enrichment. The change was the same as in the case of 18%.

  In the new fuel assembly loaded in the equilibrium core of the light water breeder reactor, the void coefficient when the ratio of enrichment of fissile Pu is changed, and when the entire core is in the state of 100% void, it is thrown into the core Changes in reactivity were examined. FIG. 8 obtained as a result of this examination shows the relationship between the ratio of enrichment of fissile Pu, the void coefficient, and the input reactivity when the entire core is in a 100% void state. Characteristic S shows a change in the void coefficient depending on the ratio of enrichment of fissile Pu. Characteristic T shows a change in the input reactivity depending on the ratio of enrichment of fissile Pu.

  The BWR has a density distribution of hydrogen atoms responsible for the neutron moderation function in the core axis direction. Therefore, the enrichment of the fissionable Pu in the lower fuel region where the hydrogen atom density is high is that in the upper fuel region where the hydrogen atom density is low. It is desirable to make it higher than the degree of enrichment. If the enrichment of fissionable Pu is set higher than 22%, the effect of increasing the enrichment of fissionable Pu is diminished by the self-shielding effect of various TRUs in the resonance energy region. For this reason, the amount of fissile Pu necessary to keep the core critical is unnecessarily increased, and the economics of the BWR are impaired. Moreover, if the fissile Pu enrichment is made lower than 14%, the neutron energy spectrum shifts to the low energy side, the growth ratio decreases, the void coefficient worsens by more than 10%, and the economics of BWR and There is a risk of sacrificing safety.

  However, as represented by characteristic T in FIG. 8, the enrichment of fissile Pu in the total nuclear fuel material in the upper fuel region is 1.05 (the enrichment of fissile Pu in the total nuclear fuel material in the lower fuel region. If it is smaller than 19/16), the input reactivity when it is assumed that the entire core is in a state of 100% void exceeds 1 dollar (about 0.34% ΔK) and enters the prompt critical region. Entering the prompt critical region must be avoided. For this reason, in the unlikely event that an accident outside the first design standard occurs and the high-pressure core water injection system is activated, the enrichment of the fissionable Pu in the whole nuclear fuel material in the upper fuel region is reduced to the entire lower fuel region. It must be less than 1.05 times the enrichment of fissile Pu in nuclear fuel material. If the enrichment of fissile Pu in the total fuel material of the upper fuel region exceeds 1.6 (22/14) times the enrichment of fissile Pu in the total fuel material of the lower fuel region, The absolute value of the void coefficient may be small, and it may be difficult to meet safety standards depending on abnormal transient changes and accident types. For this reason, the enrichment of the fissile Pu in the whole nuclear fuel material in the upper fuel region is made 1.6 times or less than the enrichment of the fissile Pu in the whole nuclear fuel material in the lower fuel region.

  Based on the characteristic T shown in FIG. 8, the enrichment of fissile Pu in the total nuclear fuel material in the upper fuel region is 1.25 (20/16) of the enrichment of fissile Pu in the total nuclear fuel material in the lower fuel region. ) By multiplying it by a factor of 2 or more, even if the entire core becomes 100% voided for some reason, it is possible to avoid a positive reactivity from entering the core.

  For this reason, by making the enrichment of fissile Pu in the whole nuclear fuel material in the upper fuel region within the range of 1.05 to 1.6 of the enrichment of the fissile Pu in the whole nuclear fuel material in the lower fuel region. The thermal margin of the core can be increased. As a result, the safety margin of the core can be improved without impairing the economics of the core. Preferably, the enrichment of fissile Pu in the total fuel material of the upper fuel region is within the range of 1.25 to 1.6 of the enrichment of fissile Pu in the total fuel material of the lower fuel region. desirable. The above shows the result of the study on the equilibrium core, but the same can be said for the initial loading core and the transition core that reaches the equilibrium core.

  The configuration (4) described above will be described.

  Spent nuclear fuel contained in spent fuel assemblies generated in large quantities from light water reactors shall be handled either by reprocessing nuclear fuel and recycling TRUs, or by direct geological disposal of spent fuel assemblies Is considered. However, since it is difficult to determine the geological disposal site of the spent fuel assembly, an intermediate storage path for the spent fuel assembly is also considered. On the other hand, there is concern that the fear that TRU newly generated by the operation of the light water reactor will become a long-lived radioactive waste hinders the addition of the light water reactor. Therefore, as an immediate measure for the popularization of light water reactors, the inventors studied to fission the TRU using the BWR currently in operation to greatly reduce the number of spent fuel assemblies.

  As an example of recycling TRUs in currently operating light water reactors, only Pu among TRUs that are so-called Pu thermal use is recycled in Europe and the like once. However, if you continue to recycle TRUs, you will not be able to comply with safety constraints, so you will need to repeat TRU multiple recycling while keeping the safety constraints, greatly increasing the number of spent fuel assemblies. It is necessary to reduce it.

  The inventors examined measures that can reduce the number of spent fuel assemblies. As a result, the inventors have loaded a plurality of fuel assemblies that contain super uranium nuclides and have different numbers of recycles of the super uranium nuclides, and among these fuel assemblies, the ultra uranium nuclides with the least number of recycles. A plurality of fuel assemblies containing nuclides are arranged in the center of the core, and the fuel assemblies containing super-uranium nuclides that are recycled many times between the center and the outermost layer of the core are arranged on the outermost layer side of the core. It was found that the number of spent fuel assemblies can be reduced by doing (configuration (4)).

  The core configuration (4) can be realized, for example, as follows. That is, TRUs with different numbers of recycles are included in separate fuel assemblies, and these fuel assemblies are loaded into the core of one light water reactor. When the burnup is zero, the fuel assemblies that are enriched with the same number of recycles of TRUs and have different core stay periods are loaded adjacent to each other in the core. Multiple fuel assemblies containing ultra-uranium nuclides with the least number of recycles are arranged in the central part of the core, and between these central parts and the outermost layer of the core, fuel assemblies containing ultra-uranium nuclides with the highest number of recycles , Arranged on the outermost layer region side of the core.

  TRUs with different numbers of recycles need to be loaded separately into different fuel assemblies without mixing. The same for each fuel assembly containing the same recycle number of TRUs when TRUs obtained by reprocessing spent nuclear fuel generated in light water reactors using low-enriched uranium and TRUs as nuclear fuel are enriched in uranium and subjected to multiple recycling It is preferable to determine the TRU loading amount of the new fuel assembly so that the average value of the infinite effective multiplication factor of all the fuel assemblies including the recycle number of TRUs and having different core residence years is substantially the same value.

  The higher the number of recycled TRUs, the smaller the percentage of Pu-239 in the TRUs. Therefore, in the configuration of (4), when TRU is subjected to multiple recycling, a plurality of fuel assemblies containing nuclear fuel material having the highest ratio of Pu-239 in the TRU are arranged in the central portion of the core. A fuel assembly containing nuclear fuel material having a smaller ratio of Pu-239 in the TRU between the core outermost layer regions is the same as that disposed on the core outermost layer region side.

  By adopting the configuration of (4), it is possible to improve the safety of recycled light water reactors that use nuclear fuel materials including TRU obtained by nuclear fuel reprocessing, increase the thermal margin, and reduce the number of spent fuel assemblies. You can aim.

  The examination described above will be described in detail. This study was conducted on the core of ABWR currently in operation. The target ABWR core has a low enrichment with an average enrichment of 4.8%, for example, with an electric power of 1350 MW, 872 fuel assemblies loaded in the core and 74 fuel rods per fuel assembly. It is a core using uranium.

  A fuel assembly using low-enriched uranium as a nuclear fuel material is loaded, for example, in the core of ABWR. Newly manufactured using nuclear fuel material obtained by enriching only the Pu recovered by reprocessing the spent nuclear fuel in the spent fuel assembly generated in this ABWR with depleted uranium, natural uranium or depleted uranium, etc. For example, a BWR core loaded with a fuel assembly is generally called a Pu thermal core. Of the Pu thermal furnace, a core that is a fuel assembly having no nuclear fuel material containing low enriched uranium and having a nuclear fuel material containing Pu recovered by nuclear fuel reprocessing. Is referred to as a Full MOX core.

  In a core of a light water reactor loaded with not only Pu but also a fuel assembly having nuclear fuel material including all TRUs recovered by nuclear fuel reprocessing, the core loaded with only the fuel assembly including TRU that has been recycled once is TRU. Called the first generation recycling core. A core loaded with a fuel assembly including a TRU having a recycle count of 2 times obtained by reprocessing spent nuclear fuel contained in the spent fuel assembly taken out of the TRU first generation recycling core Called the second generation recycling core. A reactor core loaded with a fuel assembly including a TRU of 3 recycles obtained by reprocessing the spent nuclear fuel contained in the spent fuel assembly taken out of the TRU second generation recycling core This is called the third generation recycling core. In this way, the number of generations of the core increases as the fuel assembly including the TRU having increased in the number of recycles is loaded.

  As described above, when the number of TRUs recycled is increased, the proportion of even-numbered nuclides in the TRU increases, and the absolute value of the negative void coefficient decreases. For this reason, there is no safety margin in the core, and TRU multiple recycling cannot be continued. According to WSYang et al., A Metal Fuel Core Concept for 1000MWt Advanced Burner Reactor GLOBAL '07 Boise, USA, September, 2007, P.52 Is being considered. The Advanced Burner Reactor (ABR) uses depleted uranium containing TRU obtained by reprocessing spent fuel in a light water reactor as nuclear fuel material in TRU first generation recycling. In TRU second generation recycling, all the TRUs obtained by reprocessing spent nuclear fuel contained in the spent fuel assemblies taken out from the TRU first generation recycling core are filled into the fuel assemblies and recycled. At the same time, attempts have been made to make up for the shortage of TRU, which has decreased due to combustion in the TRU first-generation recycling core, with TRU obtained by reprocessing spent nuclear fuel in light water reactors.

  Thus, as long as the light water reactor and the ABR are operated in parallel, TRUs from spent nuclear fuel generated in the light water reactor will continue to be accommodated in the core of the light water reactor and the fuel cycle facility. For this reason, storage outside the reactor of the TRU can be avoided for the time being. When this concept of ABR was tried at ABWR, it was possible to continue the multiple recycling of TRUs longer than the multiple MOX multiple recycling core. However, although multiple TRU recycling is limited to the TRU 4th generation recycling core, the number of spent fuel assemblies generated is reduced by a factor of 10 compared to when TRU recycling is not performed. The When the configuration of (4) is applied and multiple TRU recycling is continued up to the TRU 8th generation recycling core, it occurs when the number of spent fuel assemblies generated does not recycle TRU. It can be reduced to less than 1% of its body number.

  As a result of investigations by the inventors for the TRU multiple recycle core of a normal Full MOX, the cause of the void coefficient shifting to the positive side when the TRU is continuously recycled is as follows: (I) The ratio of even-numbered nuclei in the TRU As the void ratio in the core increases and the void ratio in the core increases, it has been found that the radial power distribution is higher in the core and lower in the periphery. .

  In FIG. 9, the characteristic V indicates the power distribution in the core radial direction during the rated power operation of the BWR, and the characteristic X indicates the power distribution in the core radial direction when the entire core is 100% voided. Show. As the recycling generation of the TRU recycling furnace progresses, the rate of increase of the infinite neutron multiplication factor of the fuel assembly when the void ratio of the coolant increases increases. Utilizing this phenomenon, a plurality of fuel assemblies containing super uranium nuclides and different numbers of recycles of super uranium nuclides are loaded, and among these fuel assemblies, the ultra uranium nuclides with the least number of recycles Are arranged in the center of the core, and between the center and the outermost layer region of the core, the fuel assemblies containing the most uranium nuclides are arranged closer to the outermost layer region of the core. This makes it possible to alleviate the shift of the power distribution in the radial direction toward the core center. As a result, multiple recycling of TRUs is possible while meeting safety standards, and the number of spent fuel assemblies generated can be reduced.

  The mitigation of the power distribution shift to the core center will be specifically described. If the entire core becomes 100% void, one of the main reasons for the large positive reactivity being introduced into the nuclear fuel material region is that the entire core is 100 from the state of void distribution at the rated power of the BWR. The inventors have found that the power distribution in the radial direction may shift to the central part of the core where the neutron importance is high when shifting to the state of the% void fraction. When multiple TRU recycling is performed, the proportion of Pu-239 in all TRUs decreases sequentially as the number of TRU recycling increases, and the infinite neutron multiplication factor of the fuel assembly containing TRU increases when the void ratio increases The amount increases. Therefore, by loading a fuel assembly containing TRU with a small number of recycles in the center of the core and loading a fuel assembly containing TRU with a high number of recycles in the periphery of the core, the void distribution during rated power operation is reduced. It is possible to mitigate the shift of the radial power distribution to the center of the core that occurs when the entire core shifts from the state to the state of 100% void ratio. For this reason, even when the entire core is in a 100% void state, a core in which no positive reactivity is introduced can be realized. Note that the power distribution in the radial direction of the core is flattened by adjusting the ratio of the number of loaded bodies among fuel assemblies including TRUs with different numbers of recycles.

  FIG. 10 shows an example of the core of a light water reactor to which the configuration (4) is applied, and the state at the start of the operation cycle. In the core of this light water reactor, the number of recycles from one to eight times from a plurality of fuel assemblies A including a TRU that is recycled once to a plurality of fuel assemblies that include a TRU that is recycled eight times. Fuel assemblies A to H, each containing a number of TRUs, are loaded. In FIG. 10, alphabets A, B, C, D, E, F, G, and H indicate the number of times TRU is recycled. In FIG. 10, numerals 1, 2, 3, 4 and 5 attached after the alphabet indicate the period (number of operating cycles) of the corresponding fuel assembly staying in the core. For example, the fuel assembly B3 is a fuel assembly that includes a TRU that is recycled twice and is experiencing operation in the third operation cycle after being loaded in the core. The number “5” is the fuel assembly that is experiencing the fifth operating cycle.

  The fuel assemblies A to C and a part of the fuel assembly D are taken out from the reactor as spent fuel assemblies after the operation of the fourth operation cycle is completed after being loaded on the core. . The rest of the fuel assembly D and the fuel assemblies F to H stay in the core until the operation in the fifth operation cycle is completed after being loaded in the core.

  In the balanced core, the total amount of TRU recovered by reprocessing spent nuclear fuel contained in the fuel assembly A4 taken out from the core as a spent fuel assembly is a plurality of newly produced fuel assemblies B1. Are dispersed and filled. The total amount of TRU recovered by reprocessing spent nuclear fuel contained in the fuel assembly B4 taken out from the core as a spent fuel assembly is dispersed and filled in a plurality of newly manufactured fuel assemblies C1. Is done. Similarly, the total amount of TRU recovered from spent nuclear fuel contained in the fuel assembly C4 taken out from the core is dispersed and filled in a plurality of new fuel assemblies D1, and the fuel assembly E5 taken out from the core is loaded. The total amount of TRU recovered from the contained spent nuclear fuel is distributed and filled in a plurality of new fuel assemblies F1. The total amount of TRU recovered from the spent nuclear fuel contained in the fuel assembly G5 is distributed and filled in a plurality of new fuel assemblies H1, and finally only the fuel assembly H5 remains as a spent fuel assembly.

  The number of fuel assemblies of the fuel assemblies A to H is determined so that the infinite effective multiplication factor of each fuel assembly is substantially the same so that the power distribution in the radial direction of the core can be kept flat. In the core shown in FIG. 10, the number of fuel assemblies A to H at the time of zero burnup is 100, 40, 24, 16, 12, 12, 8, 4, respectively. It is. A plurality of fuel assemblies including the same number of recycles of TRUs are arranged adjacent to fuel assemblies having different core stay periods.

By placing fuel assemblies containing TRUs with a higher number of recycles on the outermost layer region side of the core, the void ratio of the core is lower than in a core loaded with only fuel assemblies containing TRUs with the same number of recycles. When it rises, the increase of the infinite neutron effective multiplication factor at the core center becomes relatively smaller than the increase of the infinite neutron effective multiplication factor at the core periphery. For this reason, the shift of the power distribution in the radial direction to the center of the core is reduced. As a result, despite the core loaded with a nuclear fuel assembly that contains each TRU that has been recycled up to eight times, the light water reactor is maintained with a void coefficient of -4 × 10 ⁻⁴ % Δk /% void. It becomes possible to drive. It can be seen that this core can reduce the number of spent fuel assemblies to 0.5% or less compared to the case where TRU is not recycled.

  The case where each fuel assembly from a fuel assembly including a TRU with one recycling to a fuel assembly including a TRU with eight recyclings coexists in one core has been described as an example. A suitable core configuration may be used. For example, a core loaded with only a fuel assembly including a TRU with one recycle as a fuel assembly, a fuel assembly including a TRU with one recycle, and a fuel assembly including a TRU with two recycles A loaded core, a fuel assembly including a TRU with one recycling, a fuel assembly including a TRU with two recyclings, a core loaded with a fuel assembly including a TRU with three recyclings, etc. Conceivable.

  Also, the case of recycling all TRUs recovered from each spent fuel assembly taken out of the light water reactor has been discussed. However, when only Pu among the recovered TRUs is recycled, some nuclides in the TRU In the case of specifying and recycling together with Pu, the concept of recycling all TRUs can be applied as it is.

  The safety margin can be further improved by combining some of the configurations (1), (2), and (3). For example, when (2) is combined with (1), the safety margin is higher than that of (1) alone, and when (3) is further combined with (1) and (2) (1) ) And (2), the safety margin is further improved. These are also true for the combination of the other two configurations including (2) and the combination of the other two configurations including (3).

  An embodiment of the present invention to which the concept described above is applied will be described in detail below with reference to the drawings.

  The core of the 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 core 20 of the light water reactor of the present embodiment has the above-described configurations (1), (2), and (3).

  The core 20 of the light water reactor is a core for an electric output of 1350 MW, but the output scale is not limited to this. By changing the number of fuel assemblies loaded in the core 20, it is possible to realize a core of another output scale to which the present embodiment can be applied.

  An outline of a boiling water reactor (BWR), which is a light water reactor for electric power 1350 MW, to which the core 20 of the present embodiment is applied will be described with reference to FIG. The BWR 1 has a reactor core 20, a steam / water separator 21, and a steam dryer 22 disposed in a reactor pressure vessel 27. The core 20 is surrounded by a core shroud 25 in the reactor pressure vessel 27. A core support plate 17 disposed at the lower end of the core 20 is disposed in the core shroud 25 and installed in the core shroud 25. The upper lattice plate 18 disposed at the upper end of the core 20 is disposed in the core shroud 25 and installed in the core shroud 25. A plurality of control rods 42 are arranged at positions where they can be inserted into the core 20. These control rods 42 are inserted into the core from below. 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 29 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. The BWR 1 is provided with a low pressure core water injection system 31 and a high pressure core water injection system 32 as an emergency core cooling system when the coolant supplied to the core is lost for some reason.

  As shown in FIG. 12, 720 fuel assemblies 41 are loaded in the core 20. Three fuel assemblies 41 are provided with Y-shaped control rods 42 at a ratio of one, and 223 control rods 42 are arranged. Each control rod 42 is connected to a separate control rod drive provided at the bottom of the reactor pressure vessel 27. The control rod drive device is motor-driven and can finely adjust the movement of the control rod 42 in the axial direction. The control rod drive unit performs the operations of pulling out the control rod 42 from the core and inserting the control rod 42 into the core. About 1/5 of the 223 control rods 42 are control rods that adjust the output of the reactor by being inserted into or pulled out from the core of the operating BWR 1, and the remaining about 4/5 are operating rods. The control rod 42 is inserted into the core when the reactor is stopped in a state of being pulled out from the core of the BWR 1.

  The fuel assembly 41 has a nuclear fuel material region 16 loaded with nuclear fuel material. Within this nuclear fuel material region 16, an upper blanket region 5, an upper fuel region 6, an inner blanket region 7, a lower fuel region 8, and Five regions of the lower blanket region 9 are sequentially formed. Further, the fuel assembly 41 has a region that forms the upper reflector region 10 in a state of being loaded in the core 20 above the upper blanket region 5, and further, in the core 20 below the lower blanket region 9. And has another region that forms the lower reflector region 11 (see FIG. 16).

  The core 20 has a nuclear fuel material region 12 containing a nuclear fuel material, an upper reflector region 10A, and a lower reflector region 11A. The upper reflector region 10 </ b> A is formed above the nuclear fuel material region 12 and is formed by the upper reflector region 10 of the entire fuel assembly 41 loaded in the core 20. The lower reflector region 11 </ b> A is formed below the nuclear fuel material region 12 and is formed by the lower reflector region 11 of the entire fuel assembly 41 loaded in the core 20.

  The nuclear fuel material region 12 of the core 20 is constituted by the nuclear fuel material region 16 of the entire fuel assembly 41. The nuclear fuel material region 12 includes an upper blanket region 5A formed by the upper blanket region 5, an upper fuel region 6A formed by the upper fuel region 6, an inner blanket region 7A formed by the inner blanket region 7, and a lower fuel region 8 The lower fuel area 8A and the lower blanket area 9A formed by the lower blanket area 9 are provided. The upper blanket region 5A, the upper fuel region 6A, the inner blanket region 7A, the lower fuel region 8A, and the lower blanket region 9A are arranged in this order from the upper end of the nuclear fuel material region 12 toward the lower end of the nuclear fuel material region 12. The core 20 is a parfait core. Regions 10A, 5A, 6A, 7A, 8A, 9A, and 11A are at the same position as the respective regions 10, 5, 6, 7, 8, 9, and 11 of the fuel assembly 41 in the height direction of the core 20. .

As shown in FIG. 13, the cross section of the region of the fuel assembly 41 where the nuclear fuel material is loaded has 271 fuel rods 44 having an outer diameter of 10.1 mm in the channel box 13 which is a hexagonal cylinder. They are arranged in an equilateral triangular lattice. The cross section of the fuel assembly 41 has a hexagonal shape, and the gap between the plurality of fuel rods 44 included in the fuel assembly 41 is 1.3 mm. Nine fuel rods 44 are arranged on one side of the outermost fuel rod row. The control rod 42 having a Y-shaped cross section has three wings extending outward from a tie rod located at the center. Each wing includes a plurality of neutron absorbing members 3 filled with B ₄ C, which is a neutron absorbing material, and is disposed around the tie rod with a 120 degree interval. The control rod 42 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.

The configuration of each fuel assembly 41 will be described with reference to FIG. The fuel assembly 41 includes an upper tie plate (upper fuel support member) 14, a lower tie plate (lower fuel support member) 15, a plurality of neutron absorbing members (for example, neutron absorbing rods) 3, a plurality of fuel rods 44, and a channel. It has a box 13. The lower end portion of each fuel rod 44 is supported by the lower tie plate 15, and the upper end portion of each fuel rod 44 is supported by the upper tie plate 14. Each fuel rod 44 has a sealed cladding tube made of zirconium alloy, and the plenum 2, the nuclear fuel material region 16, and the neutron material filling region in the cladding tube of the fuel rod 44 in the axial direction from the upper end downward. 6 are arranged in this order. A nuclear fuel material region (active fuel length) 16 located above the neutron material filling region 6 filled with B ₄ C, which is a neutron absorbing material, is filled with a plurality of fuel pellets containing nuclear fuel material. A hafnium rod may be disposed in the neutron material filling region 6. 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 13 is 53%.

  The outer diameter of the fuel rod 44 (the outer diameter of the cladding tube) at the respective positions of the neutron material filling region 6 and the nuclear fuel material region 16 is 10.1 mm, which is the same. The outer diameter of the fuel rod 44 at the position of the plenum 2 (outer diameter of the cladding tube) is 5.8 mm, which is smaller than the outer diameter of the fuel rod 44 at the position of the nuclear fuel material region 16. The length of the plenum 2 is 1100 mm. The plenum 2 communicates with the neutron material filling region 6 and the nuclear fuel material region 16 in the fuel rod 44.

  Each fuel rod 44 is held by a fuel spacer (not shown) at several locations in the axial direction in the nuclear fuel material region 16. These fuel spacers keep the spacing between the fuel rods 44 at a predetermined width. The portion of the plenum 2 of each fuel rod 44 is also supported by the fuel spacer 33 at about three locations.

Each neutron absorbing member 3 is held on the upper tie plate 14 by a support rod (support member) 45 made of zirconium alloy. The neutron absorbing member 3 is filled with B ₄ C pellets in a sealed tube having an outer diameter of 6 mm. This tube is attached to the support bar 45. The neutron absorbing member 3 may be configured by filling a hafnium rod in a tube. Each neutron absorbing member 3 is disposed between the plenums 2 of adjacent fuel rods 44, and one neutron absorbing member 3 is provided in one fuel rod 44 (see FIG. 18). Each neutron absorbing member 3 is disposed between the upper end of the nuclear fuel material region 12, that is, between the upper end of the nuclear fuel material region 16 and the lower end of the upper tie plate 14. The length of the neutron absorbing member 3 is 500 mm, and the distance between the upper end of the nuclear fuel material region 16 and the lower end of the neutron absorbing member 3 is 300 mm. In the present embodiment, the ratio of the total cross sectional area of all the neutron absorbing members 3 to the cross sectional area of the fuel assembly lattice is 16.8%. The ratio of the total cross-sectional area of all neutron absorber filling regions 3 to the cross-sectional area of the fuel assembly lattice is 49.3%. This 49.3% does not include the cross-sectional area of the control rod 42.

  When the BWR 1 is in operation, the cooling water in the downcomer 29 is pressurized by the rotation of the internal pump 26 and supplied to the core 20. The cooling water supplied into the reactor core 20 is guided into each fuel assembly 41 and heated by heat generated by fission of the fissile material, and a part thereof becomes steam. For this reason, the gas-liquid two-phase flow including cooling water and steam rises in the upper reflector region 10 in the fuel assembly 41. This gas-liquid two-phase flow is guided from the core 20 to the steam separator 21 to separate the steam. The separated steam is further dehumidified 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 to rotate the turbine. 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. This condensed water is introduced into the reactor pressure vessel 27 through the water supply pipe 24 as water supply. The cooling water separated by the steam separator 22 is mixed with the above-mentioned water supply in the downcomer 29 and is pressurized by the internal pump 26 again.

  The arrangement of the core fuel assemblies 41 in an equilibrium state will be described with reference to FIGS. 14 and 15. The fuel assembly 41E having the longest residence time in the reactor and the fifth operation cycle is arranged in the core 20 having a low neutron importance. The first cycle fuel assembly 41A having the highest infinite neutron multiplication factor and the shortest residence time in the reactor is loaded in the outer core region 48 located inside the outermost core region 46 of the core 20, and the core. The output distribution in the radial direction is flattened. The fuel assemblies 41B, 41C, and 41D in the second, third, and fourth cycles in the reactor stay period are arranged in the core inner region 50 that exists inside the core outer region 48, respectively. With such an arrangement, the power distribution in the core inner region 50 is flattened.

  The fuel assemblies 41A, 41B, 41C, 41D, and 41E are the fuel assemblies 41 shown in FIG. 13 and FIGS. 19 and 20 described later, respectively. The lower tie plates 15 of these fuel assemblies are supported by a plurality of fuel support fittings (not shown) provided on the core support plate 17. A coolant passage for guiding cooling water to the fuel assembly 41 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. In the core 20, three regions of a core outermost layer region 46, a core outer region 48 and a core inner region 50 are formed in the radial direction (see FIG. 15). The orifice of the orifice located in the outermost core region 46 of the core where the output of the fuel assembly 41 is the smallest is set to be the smallest, and the orifice is set to increase in the order of the orifice located in the core outer region 48 and the orifice located in the core inner region 50. To do. The diameter of the orifice located in the core inner region 50 is the largest.

  As shown in FIG. 16, the height of each region in the nuclear fuel material region 16 of the fuel assembly 41 is as follows. The height of the upper blanket region 5 (upper blanket region 5A) is 70 mm, the height of the upper fuel region 6 (upper fuel region 6A) is 283 mm, and the height of the inner blanket region 7 (inner blanket region 7A) is The height of the lower fuel region 8 (lower fuel region 8A) is 194 mm, and the height of the lower blanket region 9 (lower blanket region 9A) is 280 mm. Further, an upper reflector region 10 (upper reflector region 10A) having a length of 1100 mm upward from the upper end of the nuclear fuel material region 16 is formed. The upper reflector region 10 includes cooling water (gas-liquid two-phase flow during operation of the BWR 1) that exists between the plenums 2 of the fuel rods 41. A lower reflector region 11 (lower reflector region 11A) having a length of 70 mm downward from the lower end of the nuclear fuel material region 16 is formed. The lower reflector region 11 includes cooling water that exists between the neutron material filling regions 6 of the fuel rods 41. The numerical values of the length of the upper reflector region 10 and the length of the lower reflector region 11 indicate the length in the axial length of the fuel rods arranged in the fuel assembly. In each example described later. The same applies to the length of the upper reflector region 10 and the length of the lower reflector region 11.

  The neutron absorbing member 3 and the support bar 45 are disposed in the upper reflector region 10 (upper reflector region 10A).

  When the fuel assembly 41 has a burnup of zero, all fuel rods 44 of the fuel assembly 41 (the fuel rods 44A to 44E shown in FIG. 19) have three blanket regions, the upper blanket region 5 and the internal blanket. Region 7 and lower blanket region 9 are filled with depleted uranium, and upper fuel region 6 is mixed with depleted uranium at a rate of 213 when the weight of TRU is 100. Filled with mixed oxide fuel, and the lower fuel region 8 is filled with mixed oxide fuel with a fissionable Pu enrichment of 20.2 wt% mixed with deteriorated uranium at a weight ratio of 143 when the weight of TRU is 100 doing. The upper blanket area 5, the inner blanket area 7 and the lower blanket area 9 do not contain TRUs. The average fissile Pu enrichment in the upper fuel region 6 and the lower fuel region 8 is 17.5 wt%. TRU is a material recovered by nuclear fuel reprocessing from nuclear fuel material (spent nuclear fuel) contained in the fuel assembly 41 taken out from the reactor pressure vessel 27 as a spent fuel assembly. 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 41 has a plurality of fuel rods 44A to 44E as the fuel rods 44, and these fuel rods are arranged as shown in FIGS. 19 shows a cross section of the fuel assembly 41 in the upper fuel region 6, and FIG. 20 shows a cross section of the fuel assembly 41 in the lower fuel region 8. The mixed oxide fuel filled in the upper fuel region 6 of each of the fuel rods 44A to 44E has the following fissionable Pu enrichment in the state of zero burnup (see FIG. 19). Fuel rod 44A has a fissile Pu enrichment of 8.4 wt%, fuel rod 44B has a fissile Pu enrichment of 11.2 wt%, fuel rod 44C has a fissile Pu enrichment of 14.5 wt%, fuel rod 44D has a fissile Pu enrichment of 15.9 wt% and fuel rod 44E has a fissile Pu enrichment of 17.2 wt%.

  The mixed oxide fuel filled in the lower fuel region 8 of each of the fuel rods 44A to 44E has the fissile Pu enrichment shown below in a state where the burnup is zero (see FIG. 20). Fuel rod 44A has a fissile Pu enrichment of 13.1 wt%, fuel rod 44B has a fissile Pu enrichment of 15.9 wt%, fuel rod 44C has a fissile Pu enrichment of 19.2 wt%, fuel rod In 44D, the fissile Pu enrichment is 20.7 wt%, and in the fuel rod 44E, the fissile Pu enrichment is 21.4 wt%.

  Although no TRU exists in each blanket region of the fuel rods 44A to 44E, each mixed oxide fuel in the upper fuel region 6 and the lower fuel region 8 of each fuel rod 44A to 44E contains TRUs having the composition shown in Table 1. It is out. The fuel assembly 41 has a burnup of zero, and the ratio of Pu-239 in all TRUs is 44 wt%. Table 1 shows that after the fuel assembly 41 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. The composition of TRUs present in the nuclear fuel material contained in the fuel assembly loaded into the core obtained by reprocessing the nuclear fuel material in the spent fuel assembly is shown.

  During operation of the BWR 1, volatile fission products generated by fission of fissile material in each fuel rod 44 are stored in the plenum 2. Since the plenum 2 has a length of 1100 mm, it can store a sufficient amount of volatile fission products generated by fission of the fissile material. For this reason, the soundness of the fuel rod 44 can be ensured.

  According to this embodiment, a plurality of neutron absorbing members 3 having a length of 500 mm are disposed at a position 300 mm above the upper end of the nuclear fuel material region 12, and a plurality of neutron absorptions are provided below the lower end of the nuclear fuel material region 12. Since the material filling region 4 is arranged, even if it is assumed that the entire core is in a state of 100% void, which is an event that cannot occur as an originating event in ABWR, it leaks upward and downward from the nuclear fuel material region 12 Neutrons to be absorbed can be absorbed by the neutron absorbing member 3 and the neutron absorbing material filling region 4. For this reason, even if the entire core is in a state of 100% void, it is possible to avoid the positive reactivity being introduced into the nuclear fuel material region 12. When this happens, a negative reactivity is introduced into the nuclear fuel material region 12.

  Further, the core 20 has an upper fuel region 6 having a fissile Pu enrichment of 15.7 wt% and a height of 283 mm, and a lower fuel region 8 having a fissile Pu enrichment of 20.2 wt% and a height of 194 mm. The average fissile Pu enrichment in the upper fuel region 6 and the lower fuel region 8 is 17.5 wt%. The sum of the height of the lower fuel region 8 and the height of the upper fuel region 6 is 477 mm, the height of the upper fuel region 6 is 1.46 times the height of the lower fuel region 8, and the fissile Pu enrichment of the lower fuel region 8 The degree is 1.29 times the fissile Pu enrichment of the upper fuel region 6. For this reason, the core 20 has a growth ratio of 1 or more, and can further increase the thermal margin. As a result, the core 20 of the present embodiment can reduce the maximum linear power density by 2% compared to the case where the upper and lower fuel regions have the same fissile Pu enrichment, and the void coefficient is negative. The BWR 1 having such a core 20 can continue TRU multiple recycling.

  Since the present embodiment has the configurations of (1), (2) and (3), even if the entire core is in a 100% void state, the nuclear fuel material region 12 has a positive reactivity. When not inserted, the fuel rods are more sound and the thermal margin is increased. Therefore, this embodiment can further increase the safety margin without impairing the economics of the light water reactor.

  In the core 20 of the present embodiment, when an electric power 1350 MW that is the same as that of ABWR is generated using a reactor pressure vessel 27 that is approximately the same size as the current ABWR, the upper fuel excluding the upper blanket region 5A and the lower blanket region 9A. The extraction burn-up of the nuclear fuel material region 12 including the region 6A, the lower fuel region 8A, and the inner blanket region 7A is 53 GWd / t, and the extraction burn-up of the nuclear fuel material region 12 including the upper blanket region 5A and the lower blanket region 9A is 45 GWd / t. In the core 20, the void coefficient is −3 × 10 −4 Δk / k /% void and the MCPR is 1.3. Therefore, in the core 20, a growth ratio of 1.01 can be realized with the ratio of each isotope of TRU kept substantially constant as described above.

  In this embodiment, since the neutron absorbing member 3 is arranged in the upper reflector region 10, the flow area of the gas-liquid two-phase flow formed between the plenums 2 of the fuel rod 44 in the II-II cross section of FIG. It becomes narrower and the pressure loss in the upper reflector region 10 increases. Since the pressure loss in the upper reflector region 10 is smaller than the pressure loss in the core 20, there is no particular problem. In the upper blanket region 5 and the upper reflector region 10, the pressure loss of the upper reflector region 10 can be reduced by forming a plurality of openings penetrating the side wall of the channel box 13.

Instead of the neutron absorbing member 3 shown in FIGS. 17 and 18, a neutron absorbing member 3A shown in FIG. 21 and a neutron absorbing member 3B shown in FIG. 22 may be used. A fuel assembly 41F shown in FIG. 21 has a configuration in which the neutron absorbing member 3 is replaced with a neutron absorbing member 3A in the fuel assembly 41 shown in FIG. The other structure of the fuel assembly 41F is the same as that of the fuel assembly 41. The neutron absorbing member 3 </ b> A is an annular body and is disposed so as to surround the plenum 2 of each fuel rod 44. The neutron absorbing member 3A is formed between an outer surface of the fuel rod 44 and the sealed container by attaching an annular sealed container 38 having an outer diameter of 8.8 mm disposed around the outer surface of the fuel rod 44 to the outer surface of the fuel rod 44. The annular region is filled with B ₄ C. The upper and lower ends of the sealed container are sealed. When the neutron absorbing member 3A is applied, the indicator bar 45 is not necessary.

A fuel assembly 41G shown in FIG. 22 has a configuration in which the neutron absorbing member 3 is replaced with a neutron absorbing member 3B in the fuel assembly 41 shown in FIG. The other structure of the fuel assembly 41G is the same as that of the fuel assembly 41. The neutron absorbing member 3B is an Hf plate and is arranged between the fuel rods 44 array. Both ends of each Hf plate having a thickness of 1.5 mm are attached to a frame member 5 having a hexagonal cross section. The frame member 5 is disposed along the inner surface of the channel box 13 and is attached to the upper tie plate 14 by a plurality of support bars 45. The plenum 2 is arranged so as to surround the plenum 2. The neutron absorbing member 3A is filled with B ₄ C in an annular sealed container having an outer diameter of 8.8 mm. Since this sealed container is attached to the support rod 45, the neutron absorbing member 3 </ b> A is held by the upper tie plate 14 as with the fuel assembly 41.

  Even if the neutron absorbing members 3A and 3B are used, the same effect as that of the first embodiment can be obtained.

  The core of the 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. The core 20A of the light water reactor according to the present embodiment has the above-described configurations (1) and (2).

  The core 20A of the light water reactor has a configuration in which the fuel assembly 41 is replaced with the fuel assembly 41H in the core 20 of the first embodiment. The other configuration of the core 20A is the same as that of the core 20. The difference between the core 20A and the core 20 will be described. Similarly to the core 20, the core 20 </ b> A is a parfait core. A BWR, which is a light water reactor to which the core 20A is applied, has a configuration in which the core 20 is replaced with the core 20A in the BWR1. The BWR is a TRU extinguishing reactor having the same configuration as the BWR 1 except for the core 20 and having a core 20A.

  In the fuel assembly 41H (see FIGS. 23 and 25) loaded in the core 20A, 397 fuel rods 44F having an outer diameter of 7.2 mm are arranged in an equilateral triangular lattice in the channel box 13. The gap between the fuel rods 44F is 2.2 mm, and 11 fuel rods 44F are arranged on one side of 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 13 is 36%. As shown in FIG. 24, the reactor core 20A is provided with fuel assemblies 41A to 41D having different numbers of operating cycles as shown in FIG. The fuel assembly 41D having the fourth operation cycle number is disposed in the core outermost layer region 46 (see FIG. 15). The fuel assembly 41A having the first operation cycle number is arranged in the core outer region 48, and the fuel assemblies 41B, 41C, 41D having the operation cycle number second to fourth cycles are disposed in the core inner region 50. Are distributed. Between the core inner region 50 and the core outer region 48, there is an intermediate region in which a plurality of fuel assemblies 41B are annularly arranged. Such power distribution in the radial direction of the core is further flattened. Each of the fuel assemblies 41A to 41E shown in FIG. 24 is a fuel assembly 41H.

  The nuclear fuel material region 16A (see FIG. 26) where the nuclear fuel material of the fuel assembly 41H exists has a configuration in which the lower blanket 9 is removed from the fuel assembly 41. In the nuclear fuel material region 16A, as shown in FIG. 26, the height of the upper blanket region 5 is 20 mm, the height of the upper fuel region 6 is 218 mm, the height of the inner blanket region 7 is 560 mm, and the height of the lower fuel region 8 Is 224 mm. The height of the upper reflector region 10 formed above the upper blanket region 5 is 1100 mm, and the height of the lower reflector region 11 formed below the lower fuel region 8 is 70 mm.

  The nuclear fuel material region 12 of the core 20A does not have the lower blanket region 9A. The nuclear fuel material region 12 includes an upper blanket region 5A, an upper fuel region 6A, and an inner blanket region 7A having the same height as each of the upper blanket region 5, the upper fuel region 6, the inner blanket region 7, and the lower fuel region 8. And a lower fuel region 8A.

  The structure of each fuel assembly 41H will be described with reference to FIG. The fuel assembly 41H has the same configuration as the fuel assembly 44 except that the fuel rods 44 of the fuel assembly 41 are replaced with fuel rods 44F. The fuel rod 44F has the nuclear fuel material region 16A described above, the plenum 2 above the nuclear fuel material region 16A, and the neutron absorber filling region 4A below the nuclear fuel material region 16A. The outer diameter of the plenum 2 is 3.7 mm, and the length of the plenum 2 is 1100 mm. An upper blanket region 5, an upper fuel region 6, an inner blanket region 7, and a lower fuel region 8 exist in the nuclear fuel material region 16A of the fuel rod 44F. The outer diameter of the fuel rod 44F in the neutron absorber filling region 4A is 8.1 mm, which is larger than the outer diameter of the fuel rod 44F in the nuclear fuel material region 16A (see FIG. 25 and FIG. 25). (See FIG. 27). A neutron absorbing member 3 having an outer diameter of 6.2 mm is disposed between the plenums 2. The ratio of the total cross-sectional area of all the neutron absorber filling regions 3 to the cross-sectional area of the fuel assembly lattice is 44.0%. This 44.0% does not include the cross-sectional area of the control rod 42.

  When the fuel assembly 41H has a burnup of zero, all the fuel rods 44F of the fuel assembly 41H (fuel rods 44A to 44E shown in FIG. 19) have deteriorated uranium in the upper blanket region 5 and the inner blanket region 7. The upper fuel region 6 and the lower fuel region 8 are filled with TRU oxide fuel containing TRU having the composition shown in Table 2 in a state of zero burnup. The fissionable Pu enrichment of this TRU oxide fuel is 13.0 wt%, and the ratio of Pu-239 in TRU is 8.5 wt%. The TRU used for the fuel assembly 41H is obtained by reprocessing spent nuclear fuel contained in the spent fuel assembly. Each blanket region is not filled with its mixed oxide fuel and does not contain TRUs. In each blanket region, natural uranium or depleted uranium recovered from the spent fuel assembly may be used instead of depleted uranium.

  During operation of the BWR, a sufficient amount of volatile fission products generated by fission of fissionable material within each fuel rod 44F can be stored in the plenum 2 having a length of 1100 mm. For this reason, the soundness of the fuel rod 44 is increased.

  According to the present embodiment, a plurality of neutron absorbing members 3 having a length of 500 mm are disposed at a position 300 mm above the upper end of the nuclear fuel material region 12, and a plurality of neutron absorbing members 3 are disposed below the lower end of the nuclear fuel material region 12. Since the neutron absorber filling region 4A is arranged, even if it is assumed that the entire core is in a state of 100% void, which is an event that cannot occur as an originating event in ABWR, a positive reaction is caused in the nuclear fuel material region 12 It is possible to avoid the degree being input. When this happens, a negative reactivity is introduced into the nuclear fuel material region 12.

  Furthermore, the present embodiment can obtain the effects produced in the first embodiment. Since the present embodiment has the configurations of (1) and (2), even if the entire core is in a 100% void state, no positive reactivity is introduced into the nuclear fuel material region 12, Fuel rod health increases. Therefore, this embodiment can further increase the safety margin without impairing the economics of the light water reactor.

  In the core 20A of the present embodiment, when a reactor pressure vessel 27 having the same size as that of the current ABWR is used and the same electric output 1350 MW as that of the ABWR is generated, the take-off burnup is 65 GWd / t and the void coefficient is −3. × 10-4Δk / k /% void, MCPR is 1.3. In the core 20A, reprocessing of spent nuclear fuel in the spent fuel assembly three years after the fuel assembly 41H is taken out from the core 20A as a spent fuel assembly while maintaining the ratio of TRU isotopes. The TRU weight obtained in (1) can be reduced by 8.3% from the TRU weight of the fuel assembly 41H newly loaded in the core. Further, in the period from when the fuel assembly 41H is loaded to the core 20A until it is taken out, the TRU fission efficiency, which is the ratio of the TRU fission weight to the total fission weight of the nuclear fuel material in the fuel assembly 41H, is 55%. is there. The retention of the TRU isotope ratio means that the TRU isotope ratio is the same in the nTRU recycling generation and the (n + 1) nTRU recycling generation.

  Also in this embodiment, any one of the neutron absorbing members 3 </ b> A and 3 </ b> B may be used instead of the neutron absorbing member 3.

  A core of a light water reactor according to embodiment 3, which is another embodiment of the present invention, will be described with reference to FIG. Similar to the core 20 of the first embodiment, the core of the light water reactor of the present embodiment has the above-described configurations (1), (2), and (3).

  The core of the present embodiment has a configuration in which the fuel assembly 41 is replaced with the fuel assembly 41I in the core 20 of the first embodiment. Other configurations of the core of the present embodiment are the same as those of the core 20. The fuel assembly 41I has a configuration in which the fuel rod 44 in the fuel assembly 41 is replaced with a fuel rod 44G. The other configuration of the fuel assembly 41I is the same as that of the fuel assembly 41.

  The fuel rod 44G included in the fuel assembly 41I has a configuration in which the plenum 2 of the fuel rod 44 is replaced with a plenum 2A. The fuel rod 44G includes the plenum 2A, the nuclear fuel material region 16 and the neutron absorber filling region 4 similar to the fuel rod 44. The plenum 2 </ b> A is disposed above the nuclear fuel material region 16, and the neutron absorber filling region 4 is disposed below the nuclear fuel material region 16. The outer diameter of each part of the nuclear fuel material region 16 and the neutron absorber filling region 4 of the fuel rod 44G is 10.1 mm. 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 13 is 53%.

  The plenum 2A has a first region 35A and a second region 35B. The outer diameter of the plenum 2A in the portion of the first region 35A is 4.8 mm, and the outer diameter of the plenum 2A in the portion of the second region 35B is 4.4 mm. The outer diameter at the first region 35A is larger than the outer diameter at the second region 35B. The first region 35A is a large diameter portion and the second region 35B is a small diameter portion. The second area 35B is located above the first area 35A. The length of the first region 35A is 300 mm, and the upper end of the first region 35A is located at a position 300 mm upward from the upper end of the nuclear fuel material region 16 (lower end of the neutron absorbing member 3). The lower end of the second area 35B is at the same position as the upper end of the first area 35A.

  The neutron absorbing member 3 is disposed between the second regions 35B which are the narrow diameter portions of the adjacent fuel rods 44G. The outer diameter of the neutron absorbing member 3 is 7.4 mm, which is larger than the outer diameter (6 mm) of the neutron absorbing member 3 used in Example 1. The ratio of the total cross sectional area of all the neutron absorbing members 3 to the cross sectional area of the fuel assembly lattice is 26.7%.

  The core of the light water reactor according to the present embodiment can satisfy all the constraints and maintain the breeding ratio of 1.01. Furthermore, in the present embodiment, even when it is assumed that the entire core is in a 100% void state, which cannot occur as an originating event in ABWR, no positive reactivity is input to the core. In particular, in this embodiment, since the outer diameter of the neutron absorbing member 3 is larger than the outer diameter of the neutron absorbing member 3 in the first embodiment, if the entire core is in a 100% void state, The reactivity introduced into is less negative than the reactivity introduced at that time in Example 1.

  Furthermore, the present embodiment can obtain the effects produced in the first embodiment. Since the present embodiment has the configurations (2) and (3), the soundness of the fuel rods is increased and the thermal margin is increased.

  Therefore, this embodiment can further increase the safety margin without impairing the economics of the light water reactor.

  A core of a light water reactor according to embodiment 4 which is another embodiment of the present invention will be described with reference to FIGS. 29 and 30. FIG. Similar to the core 20 of the first embodiment, the core of the light water reactor of the present embodiment has the above-described configurations (1) and (3).

  The core of the present embodiment has a configuration in which the fuel assembly 41 is replaced with the fuel assembly 41J in the core 20 of the first embodiment. Other configurations of the core of the present embodiment are the same as those of the core 20. The fuel assembly 41J has a configuration in which the fuel rod 44 in the fuel assembly 41 is replaced with a fuel rod 44H. The other structure of the fuel assembly 41J is the same as that of the fuel assembly 41.

  The fuel rod 44H included in the fuel assembly 41J has a configuration in which the plenum 2 of the fuel rod 44 is replaced with a plenum 2B. The fuel rod 44H includes the plenum 2B, the nuclear fuel material region 16 and the neutron absorber filling region 4 similar to the fuel rod 44. The plenum 2 </ b> B is disposed above the nuclear fuel material region 16, and the neutron absorber filling region 4 is disposed below the nuclear fuel material region 16. The outer diameter of each part of the nuclear fuel material region 16 and the neutron absorber filling region 4 of the fuel rod 44H is 10.1 mm. 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 13 is 53%. The plenum 2B has a first region 35C and a second region 35D. The outer diameter of the plenum 2B in the first region 35C is 5.8 mm, and the outer diameter of the plenum 2B in the second region 35D is 7.4 mm. The outer diameter at the portion of the second region 35D is larger than the outer diameter at the portion of the first region 35C. The first region 35C is a small diameter portion and the second region 35D is a large diameter portion. The second area 35D is located above the first area 35C. The length of the first region 35C is 800 mm, and the length of the second region 35D is 300 mm. The lower end of the second region 35D is at a position 800 mm upward from the upper end of the nuclear fuel material region 16 (the upper end of the neutron absorbing member 3), and is at the same position as the upper end of the first region 35C.

  The neutron absorbing member 3 having an outer diameter of 6 mm is disposed between the first regions 35C, which are the narrow diameter portions, of the adjacent fuel rods 44H.

  The core of the light water reactor according to the present embodiment can satisfy all the constraints and maintain the breeding ratio of 1.01. In the present embodiment, since the second region 35D of the plenum 2B existing above the upper end of the neutron absorbing member 3 has a large diameter portion, the volume of the plenum 2B is larger than that of the plenum 2 in the first embodiment. It is getting bigger. For this reason, the pressure in the fuel rod 44H is further reduced, and the soundness of the fuel rod 44H used in the present embodiment is increased as compared with the fuel rod 44 used in the first embodiment.

  Furthermore, the present embodiment can obtain the effects produced in the first embodiment. Since the present embodiment has the configurations of (1) and (3), even if the entire core is in a 100% void state, it is possible to avoid introducing positive reactivity into the core, and Thermal margin increases. Therefore, this embodiment can further increase the safety margin without impairing the economics of the light water reactor.

  The core of a light water reactor according to embodiment 5 which is another embodiment of the present invention will be described with reference to FIG. The core of the light water reactor according to the present embodiment has the configurations (1) and (3) described above. The light water reactor to which the core of the light water reactor of this embodiment is applied is a TRU extinguishing reactor.

  The core of the light water reactor according to the present embodiment has a configuration in which the fuel assembly 41H is replaced with the fuel assembly 41K in the core 20A of the second embodiment. Other configurations of the core of the present embodiment are the same as those of the core 20A. The fuel assembly 41K has a configuration in which the fuel rod 44F is replaced with the fuel rod 44I in the fuel assembly 41H. The other configuration of the fuel assembly 41K is the same as that of the fuel assembly 41H. 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 13 is 36%.

  The fuel rod 44I included in the fuel assembly 41K has a configuration in which the plenum 2 of the fuel rod 44F is replaced with a plenum 2B. The fuel rod 44J includes the plenum 2B, the nuclear fuel material region 16 and the neutron absorber filling region 4A similar to the fuel rod 44. The plenum 2B having the first region 35C and the second region 35D in the fourth embodiment is disposed above the nuclear fuel material region 16, and the neutron absorber filling region 4A is disposed below the nuclear fuel material region 16. The outer diameter of the plenum 2B in the portion of the first region 35C is 3.7 mm, and the outer diameter of the plenum 2B in the portion of the second region 35D is 5.6 mm. The outer diameter at the portion of the second region 35D is larger than the outer diameter at the portion of the first region 35C. The second region 35D is disposed above the first region 35C. The lower end of the second region 35D is at a position 800 mm upward from the upper end of the nuclear fuel material region 16 (the upper end of the neutron absorbing member 3), and is at the same position as the upper end of the first region 35C.

  The neutron absorbing member 3 having an outer diameter of 8.1 mm is disposed between the first regions 35C, which are narrow diameter portions, of the adjacent fuel rods 44I.

  The core of the light water reactor according to the present embodiment can satisfy all the constraints and maintain the breeding ratio of 1.01. In the present embodiment, since the plenum 2B existing above the upper end of the neutron absorbing member 3 has the second region 35D, the volume of the plenum 2B is the same as that of the plenum 2 of the fuel rod 44F used in the second embodiment. It becomes larger than the volume. For this reason, the pressure in the fuel rod 44I can be reduced, and the soundness of the fuel rod 44I used in the present embodiment is increased as compared with the fuel rod 44F used in the second embodiment.

  Further, the present embodiment can obtain the effects produced in the second embodiment. Since the present embodiment has the configurations of (1) and (3), even if the entire core is in a 100% void state, it is possible to avoid introducing positive reactivity into the core, and Thermal margin increases. Therefore, this embodiment can further increase the safety margin without impairing the economics of the light water reactor.

  A core of a light water reactor according to embodiment 6 which is another embodiment of the present invention will be described with reference to FIGS. 32 to 34 and Table 2. FIG. The core 20B of the light water reactor according to the present embodiment has the above-described configurations (1) and (2). The light water reactor to which the core of the light water reactor of this embodiment is applied is a TRU extinguishing reactor.

  The core 20B of the present embodiment has a configuration in which the fuel assembly 41H is replaced with the fuel assembly 41L in the core 20A of the second embodiment. The other structure of the core 20B of the present embodiment is the same as that of the core 20A. The fuel assembly 41L has a configuration in which the fuel rod 44F is replaced with the fuel rod 44J in the fuel assembly 41H. The other structure of the fuel assembly 41L is the same as that of the fuel assembly 41H. The shape of the vertical cross section of the fuel assembly 41L is the same as the shape of the vertical cross section of the fuel assembly 41H shown in FIG.

  The core 20B of the present embodiment is a one-region core having an electric output of 450 MW, and is a core applied to a TRU extinguishing reactor. In the fuel assembly 41L loaded in the core 20B, 331 fuel rods 44J having an outer diameter of 8.7 mm are arranged in a regular triangular lattice in the channel box 13. The gap between the fuel rods 44J is 1.6 mm, and ten fuel rods 44J are arranged on one side of 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 13 is 46%. As shown in FIG. 33 in the state of the equilibrium core, the reactor core 20B is provided with fuel assemblies 41A to 41D having different operating cycle numbers. The fuel assembly 41D whose operation cycle number is the fourth cycle is arranged in the core outermost layer region 46. The fuel assembly 41A having the first operation cycle number is arranged in the core outer region 48, and the fuel assemblies 41B, 41C, 41D having the operation cycle number second to fourth cycles are disposed in the core inner region 50. Are arranged in a distributed manner. Between the core inner region 50 and the core outer region 48, there is an intermediate region in which a plurality of fuel assemblies 41B are annularly arranged. The power distribution in the radial direction of the core 20B is further flattened. Each of the fuel assemblies 41A to 41D shown in FIG. 33 is a fuel assembly 41L.

  The nuclear fuel material region 16B (see FIG. 34) where the nuclear fuel material of the fuel assembly 41L exists has an upper blanket region 5, a fuel region 34, and a lower blanket region 9. An upper reflector region 10 exists above the upper end of the upper blanket region 5, and a lower reflector region 11 exists below the lower end of the lower blanket region 9. The height of the upper blanket region 5 is 20 mm, the height of the fuel region 34 is 201 mm, and the height of the lower blanket region 9 is 20 mm. The height of the upper reflector region 10 is 1100 mm, and the height of the lower reflector region 11 is 70 mm.

  Although the nuclear fuel material region 12 of the core 20B is not shown, the upper blanket region 5A, the fuel region 34A, and the lower blanket region have the same height as the upper blanket region 5, the fuel region 34, and the lower blanket region 9, respectively. 9A is included. The upper blanket region 5A, the fuel region 34A, and the lower blanket region 9A are arranged in this order in the axial direction of the core 20B.

  The outer diameter of the fuel rod 44J at the plenum 2 portion is 4.2 mm, and the outer diameter of the neutron absorbing member 3 is 6.7 mm. The outer diameter of the fuel rod 44J in the portion of the neutron absorber filling region 4A is 9.0 mm. The ratio of the total cross sectional area of all the neutron absorbing members 3 to the cross sectional area of the fuel assembly lattice is 26.0%. The ratio of the total cross-sectional area of all the neutron absorber filling regions 4A to the cross-sectional area of the fuel assembly lattice is 46.7%.

  When the fuel assembly 41L has a burnup of zero, all the fuel rods 44J of the fuel assembly 41L (the fuel rods 44A to 44E shown in FIG. 19) have deteriorated uranium in the upper blanket region 5 and the lower blanket region 9. The fuel region 34 is filled with TRU oxide fuel containing TRU having the composition shown in Table 3 in a state of zero burnup. The fissionable Pu enrichment of this TRU oxide fuel is 7.4 wt%, and the ratio of Pu-239 in TRU is 4.0 wt%. The TRU included in the fuel assembly 41L is obtained by reprocessing the spent nuclear fuel included in the spent fuel assembly. Each blanket region is not filled with its mixed oxide fuel and does not contain TRUs. In each blanket region, natural uranium or depleted uranium recovered from the spent fuel assembly may be used instead of depleted uranium.

  During operation of a light water reactor to which the core 20B of the present embodiment is applied, a sufficient amount of volatile fission products generated by fission of fissionable material in each fuel rod 44J is stored in the plenum 2 having a length of 1100 mm. be able to. For this reason, the soundness of the fuel rod 44 can be ensured.

  According to the present embodiment, a plurality of neutron absorbing members 3 having a length of 500 mm are disposed at a position 300 mm above the upper end of the nuclear fuel material region 12, and a plurality of neutron absorbing members 3 are disposed below the lower end of the nuclear fuel material region 12. Since the neutron absorber filling region 4A is arranged, even if it is assumed that the entire core is in a state of 100% void, which is an event that cannot occur as an originating event in ABWR, a positive reaction is caused in the nuclear fuel material region 12 It is possible to avoid the degree being input. When this happens, a negative reactivity is introduced into the nuclear fuel material region 12.

  Further, the present embodiment can obtain the effects produced in the second embodiment. Since the present embodiment has the configurations (1) and (2), the safety margin can be further increased without impairing the economics of the light water reactor.

  In the core 20B of the present embodiment, when an electric output of 450 MW is generated using a reactor pressure vessel having the same size as that of the current ABWR, the take-off burnup is 75 GWd / t and the void coefficient is −3 × 10−. 5Δk / k /% void, MCPR becomes 1.3. Reactor of spent nuclear fuel in the spent fuel assembly 3 years after the fuel assembly 41L is taken out from the core 20B as the spent fuel assembly, while maintaining the ratio of TRU isotopes in the core 20B The TRU weight obtained in (1) can be reduced by 7.4% from the TRU weight of the fuel assembly 41L newly loaded in the core. Further, in the period from when the fuel assembly 41L is loaded to the core 20B until it is taken out, the TRU fission efficiency, which is the ratio of TRU fission weight to the total fission weight of the nuclear fuel material in the fuel assembly 41L, is 80%. is there.

  A core of a light water reactor according to embodiment 7, which is another embodiment of the present invention, will be described with reference to FIGS. 10, 35, 36 and Table 4. FIG. The core 20C of the light water reactor of the present embodiment has the configuration (4). The core 20C of the light water reactor according to this embodiment is an ABWR core having an electric output of 1350 MW and loaded with 872 fuel assemblies 41M. A plurality of control rods 47 having a cross-shaped cross section are taken in and out of the core 20C in order to control the reactor power.

  One fuel assembly 41M has 74 fuel rods, and the cross section of the fuel assembly 41M has a square shape. In the fuel assembly 41M, as shown in FIG. 36, 74 fuel rods 44M having an outer diameter of 11.2 mm are arranged in a channel box 13A having a square cross section. These fuel rods 44K are arranged in a square lattice pattern. Two water rods 39 are arranged at the center of the cross section of the fuel assembly 41M.

  The height of the nuclear fuel material region in which the nuclear fuel material of the core 20C is loaded is 3.71 m. The plurality of fuel assemblies 41J loaded in the core 20C include fuel assemblies A, B, C, D, E, F, G, and H, as shown in FIG. In these fuel assemblies, the fuel rods 44K are filled with TRUs having different numbers of recycles as nuclear fuel materials.

  Specifically, each fuel rod 44K of the fuel assembly A with zero burnup is a TRU (recycled once) obtained by reprocessing the spent nuclear fuel in the spent fuel assembly taken out from the equilibrium core. TRU). Each fuel rod 44K of the fuel assembly B with a burnup of zero is a TRU (recycle number of 2 is 2) obtained by reprocessing the spent nuclear fuel of the fuel assembly A, which is a spent fuel assembly taken out from the equilibrium core. Times TRU). Each fuel rod 44K of the fuel assembly C with zero burnup has a TRU (recycle number of 3) obtained by reprocessing the spent nuclear fuel of the fuel assembly B, which is a spent fuel assembly taken out from the equilibrium core. Times TRU). Similarly, each fuel rod 44K of the fuel assembly D with zero burnup is filled with TRU obtained from the spent nuclear fuel of the fuel assembly C (4 times TRU), and the fuel assembly with zero burnup Each fuel rod 44K of E is filled with TRU obtained from spent nuclear fuel of the fuel assembly D (TRU with 5 recycles). Each fuel rod 44K of the fuel assembly F with zero burnup is filled with TRU (TRU with 6 recycles) obtained from the spent nuclear fuel of the fuel assembly E, and each of the fuel assemblies G with zero burnup The fuel rods 44K are filled with TRU obtained from the spent nuclear fuel of the fuel assembly F (the TRU is recycled seven times), and each fuel rod 44K of the fuel assembly H with zero burnup is the use of the fuel assembly G. TRUs obtained from spent nuclear fuel (TRUs with 7 recycles) are filled. Use of fuel assemblies in which fuel rods 44K of fuel assembly A are filled with low-enriched uranium that does not include TRU when the burnup is zero (TRU with one recycle) Recovered from spent nuclear fuel. TRUs with different numbers of recycles are filled separately into fuel rods of different fuel assemblies (for example, fuel assemblies A, B, C, etc.) without being mixed.

  In the core 20C, among the fuel assemblies A to H having different TRU recycling numbers, a plurality of fuel assemblies A including the TRU having the lowest number of recyclings are arranged in the central portion. Between the outer layer regions, the fuel assemblies including TRUs that are recycled more frequently are arranged on the outermost layer region side of the core. Specifically, the fuel assemblies B, C, D, E, F, G, and H are arranged in alphabetical order toward the central portion where the fuel assemblies A are arranged and the core outermost layer region. Is done.

  In the core 20C, there are 100 fuel assemblies A1, 40 fuel assemblies B1, 24 fuel assemblies C1, 16 fuel assemblies D1, 12 fuel assemblies E1, and 8 fuel assemblies F1. Body, four fuel assemblies G1, and four fuel assemblies H1. These numbers are the numbers when the fuel assemblies A1, B1, C1, D1, E1, F1, G1, and H1 have a burnup of zero.

  Each of the fuel assemblies A to H loaded in the core 20C includes fuel assemblies having different in-reactor stay periods (number of operation cycles). The numbers 1, 2, 3, 4, and 5 attached after the alphabet (for example, A to H described next to the fuel assembly) for identifying the fuel assembly containing TRUs having different recycle numbers are applicable. The period of stay in the furnace (the number of operation cycles) of the fuel assembly (for example, fuel assembly A, fuel assembly B, etc.) is shown. The larger the number, the longer the period of stay in the furnace. The fuel assembly marked with “1” is the fuel assembly in the first cycle during the stay in the furnace, and the fuel assembly marked with “5” is the fuel assembly in the fifth cycle during the stay in the furnace. Is the body.

  For example, the fuel assembly A1 is a fuel assembly that includes a TRU that is recycled one time and is experiencing operation in the first operation cycle after being loaded in the core 20C. The fuel assembly E5 includes a TRU that is recycled five times, and is a fuel assembly that is experiencing operation in the fifth operation cycle after being loaded in the core 20C. The fuel assemblies A to C and a part of the fuel assembly D are taken out from the reactor as spent fuel assemblies after the operation of the fourth operation cycle is completed after being loaded on the core 20C. It is. The rest of the fuel assembly D and the fuel assemblies F to H are taken out from the reactor as spent fuel assemblies after the operation in the fifth operation cycle is completed after being loaded on the core 20C. It is.

  In a plurality of fuel assemblies including TRUs having the same number of recycles, fuel assemblies having different residence times in the furnace are arranged adjacent to each other. For example, in a certain fuel assembly A1, the fuel assemblies A4 are disposed adjacent to the left and right of FIG. 10, and the fuel assemblies A3 are disposed adjacent to the top and bottom of FIG.

  Table 4 shows the TRU weight and TRU composition of each of the fuel assemblies A to H and the fuel assemblies taken out as spent fuel assemblies. A to H shown in Table 4 correspond to the fuel assemblies A to H. The fuel assembly taken out as a spent fuel assembly is a fuel assembly H5.

  According to the core 20C of the present embodiment, when the void ratio of the core 20C increases, the increase of the infinite neutron effective multiplication factor at the center of the core 20C is relative to the increase of the infinite neutron effective multiplication factor in the core outermost layer region. Become smaller. For this reason, the shift of the power distribution toward the center of the core is reduced (see FIG. 9). Therefore, the fuel assembly H5 taken out from the core 20C as a spent fuel assembly is taken out even though the fuel assemblies H1 to H5 including the TRU that is recycled eight times are loaded on the core 20C. The burnup is 45 GWd / t, and the void coefficient of the core 20C is −4 × 10−4% Δk /% void. The core 20C can reduce the number of spent fuel assemblies generated to 0.5% or less compared to the case where the TRU is not recycled.

  In the fuel assembly loaded in the core 20C of the present embodiment, the mixed oxide fuel of TRU and deteriorated uranium was used as the nuclear fuel material. However, instead of the deteriorated uranium, recovered from the natural uranium and the spent fuel assembly. Depleted uranium may be used. Instead of TRU, Pu extracted from TRU, or some minor actinide nuclides and Pu in TRU may be used.

  A core of a light water reactor according to embodiment 8, which is another embodiment of the present invention, will be described with reference to FIGS. Similar to the core 20 of the first embodiment, the core of the light water reactor of the present embodiment has the above-described configurations (1), (2), and (3).

  The core of the present embodiment has a configuration in which the fuel assembly 41 is replaced with the fuel assembly 41N in the core 20 of the first embodiment. Other configurations of the core of the present embodiment are the same as those of the core 20. The fuel assembly 41N has a configuration in which the fuel rod 44 in the fuel assembly 41 is replaced with a fuel rod 44L. The other structure of the fuel assembly 41N is the same as that of the fuel assembly 41.

  The fuel rod 44L included in the fuel assembly 41N includes the plenum 2, the nuclear fuel material region 16, and the neutron absorber filling region 4 in the same manner as the fuel rod 44 included in the fuel assembly 41. The outer diameter of each part of the plenum 2, the nuclear fuel material region 16 and the neutron absorber filling region 4 of the fuel rod 44L is the same as the outer diameter of those portions of the fuel rod 44.

  The nuclear fuel material region 16 where the nuclear fuel material of the fuel assembly 41N exists has an upper blanket region 5, an upper fuel region 6, an inner blanket region 7, a lower fuel region 8, and a lower blanket region 9, as shown in FIG. An upper reflector region 10 exists above the upper end of the upper blanket region 5, and a lower reflector region 11 exists below the lower end of the lower blanket region 9. The height of the upper blanket region 5 is 70 mm, the height of the upper fuel region 6 is 242 mm, the height of the inner blanket region 7 is 520 mm, the height of the lower fuel region 8 is 220 mm, and the height of the lower blanket region 9 is 280 mm. is there. The sum of the height of the lower fuel region 8 and the height of the upper fuel region 6 is 462 mm, and the height of the upper fuel region 6 is 1.10 times the height of the lower fuel region 8. The height of the upper reflector region 10 is 1100 mm, and the height of the lower reflector region 11 is 70 mm. In this embodiment, the fuel pellets filled in the upper fuel region 6 of the fuel rod 44L are all hollow pellets, unlike the other embodiments. In the fuel rod 44L, all fuel pellets filled in the upper blanket region 5, the inner blanket region 7, the lower fuel region 8 and the lower blanket region 9 other than the upper fuel region 6 are used in other embodiments. Solid pellet.

  When the fuel assembly 41N has a burnup of zero, all the fuel rods 44L of the fuel assembly 41N fill three blanket regions with depleted uranium, and mixed oxide fuel in the upper fuel region 6 and the lower fuel region 8 Filled. In the upper fuel region 6 and the lower fuel region 8 of the fuel assembly 41N, the enrichment degree of fissile Pu in which deteriorated uranium is mixed at a ratio of 173 weight when the weight of TRU is 100 is 18.0 wt%, respectively. is there. The TRU is recovered from the spent nuclear fuel contained in the fuel assembly 41N, which is a spent fuel assembly, by reprocessing. Each blanket region is not filled with its mixed oxide fuel and does not contain TRUs. In each blanket region, natural uranium and depleted uranium recovered from the spent fuel assembly may be used instead of depleted uranium.

  Fuel rods 44N to 44R are used as the fuel rods 44L disposed in the fuel assembly 41N. The fuel rods 44N to 44R are arranged in the channel box 13 as shown in FIG. In the fuel assembly 41N with zero burnup, the fissionable Pu enrichment in each of the upper fuel region 6 and the lower fuel region 8 is 10.7 wt% for the fuel rod 44N, 13.5 wt% for the fuel rod 44O, and the fuel rod 44P is 16.8 wt%, fuel rod 44Q is 18.2 wt%, and fuel rod 44R is 19.5 wt%.

  In the present embodiment, each effect produced in the first embodiment can be obtained. Since the present embodiment has the configurations of (1), (2) and (3), even if the entire core is in a 100% void state, the nuclear fuel material region 12 has a positive reactivity. When not inserted, the fuel rods are more sound and the thermal margin is increased. Therefore, this embodiment can further increase the safety margin without impairing the economics of the light water reactor.

  The core of the present embodiment satisfies all the constraints and can maintain a growth ratio of 1.01. Further, in this embodiment, the fissionable Pu enrichment is made equal in each of the upper fuel region 6 and the lower fuel region 8 of the fuel rods 44N to 44R in the fuel assembly 41N. For this reason, in this embodiment, the fissile Pu enrichment degree is higher than that of the fuel assembly 41 used in the first embodiment in which the upper fuel area 6 and the lower fuel area 8 have different fissionable Pu enrichment degrees. The number of types can be reduced from nine to five, and the types of fuel pellets to be manufactured can be reduced accordingly.

  A core of a light water reactor according to embodiment 9, which is another embodiment of the present invention, will be described with reference to FIG. The core of the light water reactor of the present embodiment has the configuration (2) described above.

  The core of the present embodiment has a configuration in which the fuel assembly 41 is replaced with the fuel assembly 41Q in the core 20 of the first embodiment. Other configurations of the core of the present embodiment are the same as those of the core 20. The fuel assembly 41Q has a configuration in which the neutron absorbing member 3 and the neutron absorbing material region 4 are removed from the fuel assembly 41. The other structure of the fuel assembly 41Q is the same as that of the fuel assembly 41. The fuel assembly 41Q has a plurality of fuel rods 44L. The fuel rod 44S has a configuration in which the neutron absorber region 4 is removed from the fuel rod 44 used in the first embodiment.

  In this embodiment, the outer diameter of the fuel rod 44S at the plenum 2 portion is 5.8 mm, and the length of the plenum 2 is 1100 mm. The outer diameter of the fuel rod 44S in the nuclear fuel material region 16 is 10.1 mm. 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 13 is 53%.

  For this reason, since the volume of the plenum 2 increases, the soundness of the fuel rod can be increased. Furthermore, the outer diameter of the portion of the plenum 2 formed in the fuel rod 44S is smaller than the outer diameter of the portion of the nuclear fuel material filling region 16 below the plenum portion of the fuel rod 44S. When the state becomes 100% void, the reactivity introduced into the nuclear fuel material region 12 becomes 1 dollar or less. For this reason, even if a composite event of an accident outside the first design standard occurs, the output automatically reaches the output at which the fuel rod can be cooled with the flow rate of the coolant injected into the core by the operation of the high pressure core injection system. Decreases and BWR safety is maintained. Therefore, this embodiment can further increase the safety margin without impairing the economics of the light water reactor.

  The core of the present embodiment satisfies all the constraints and can maintain a growth ratio of 1.01.

  A core of a light water reactor according to embodiment 11 which is another embodiment of the present invention will be described with reference to FIG. Similar to the core 20A of the second embodiment, the core of the light water reactor of the present embodiment has the above-described configurations (1), (2), and (3).

  The core of the present embodiment has a configuration in which the fuel assembly 41H in the second embodiment is replaced with a fuel assembly 41R (see FIG. 41). Other configurations of the core of the present embodiment are the same as those of the core 20A.

  The nuclear fuel material region 16A in the fuel assembly 41R loaded in the core of the present embodiment has an upper blanket region 5, an upper fuel region 6, an inner blanket region 7, and a lower fuel region 8 like the fuel assembly 41H. The height of the upper blanket region 5 is 50 mm, the height of the upper fuel region 6 is 183 mm, the height of the inner blanket region 7 is 560 mm, and the height of the lower fuel region 8 is 173 mm. An upper reflector region 10 having a height of 1100 mm exists above the upper blanket region 5, and a lower reflector region 11 having a height of 70 mm exists below the lower fuel region 8. The height of the upper fuel region 6 is 1.06 times the height of the lower fuel region 8.

  In the fuel assembly 41R, 397 fuel rods having an outer diameter of 7.6 mm are arranged in an equilateral triangular lattice in the channel box 13. The gap between the fuel rods is 1.8 mm, and 11 fuel rods are arranged on one side of the outermost fuel rod row. When the fuel assembly 41R has a burnup of zero, all the fuel rods of the fuel assembly 41R fill the upper blanket region 5 and the inner blanket region 7 with depleted uranium, and the upper fuel region 6 and the lower fuel region 8 TRU oxide fuel containing TRU having the composition shown in Table 5 in a state of zero burnup is filled. The fissionable Pu enrichment of this TRU oxide fuel is 18.0 wt%, and the ratio of Pu-239 in TRU is 13.4 wt%.

  In the present embodiment, each effect produced in the second embodiment can be obtained. The core of the present embodiment can efficiently extinguish TRU even with a TRU composition different from the TRU composition of the fuel assembly 41H loaded in the core 20A of the second embodiment.

  A core of a light water reactor according to embodiment 11, which is another embodiment of the present invention, will be described with reference to FIG. Similar to the core 20A of the second embodiment, the core of the light water reactor of the present embodiment has the above-described configurations (1) and (2).

The core of the present embodiment has a configuration in which the fuel assembly 41H in the second embodiment is replaced with a fuel assembly 41S (see FIG. 41). Other configurations of the core of the present embodiment are the same as those of the core 20A.

  The nuclear fuel material region 16A in the fuel assembly 41S loaded in the core of the present embodiment has an upper blanket region 5, an upper fuel region 6, an inner blanket region 7, and a lower fuel region 8 in the same manner as the fuel assembly 41H. The height of the upper blanket region 5 is 20 mm, the height of the upper fuel region 6 is 217 mm, the height of the inner blanket region 7 is 560 mm, and the height of the lower fuel region 8 is 224 mm. The height of the upper reflector region 10 is 1100 mm, and the height of the lower reflector region 11 is 70 mm.

  The cross section of the fuel assembly 41S is the same as FIG. When the fuel assembly 41S has a burnup of zero, all the fuel rods of the fuel assembly 41S fill the upper blanket region 5 and the inner blanket region 7 with thorium oxide. The upper fuel region 6 and the lower fuel region 8 of the fuel assembly 41S loaded on the core 20A of the present embodiment are taken out from the current ABWR core when the fuel assembly 41S has a burnup of zero. The TRU (hereinafter referred to as the current reactor discharge TRU) having the composition shown in Table 6 obtained by reprocessing the spent fuel assemblies (including low enriched uranium) and the mixed oxide fuel of thorium are included. The core loaded with the fuel assemblies 41S having zero burnup is the TRU first generation recycle core (hereinafter referred to as RG1 core). The current ABWR core is loaded with a fuel assembly containing low enriched uranium. To the TRU obtained by reprocessing the fuel assembly 41S taken out from the RG1 core as a spent fuel assembly, the current reactor exhaust TRU having the composition shown in Table 6 is added in an amount that makes the core critical. A core loaded with a fuel assembly 41S of zero burnup containing the mixed oxide fuel of TRU and thorium obtained in this addition in the upper fuel region 6 and the lower fuel region 8 is a TRU second generation recycle core (hereinafter referred to as a TRU second generation recycle core). , RG2 core). Thereafter, each time TRU recycling is repeated, the TRU obtained by reprocessing the fuel assembly 41S, which is the spent fuel assembly generated from the TRU recycling core of each generation, is replaced with the current reactor discharge TRU, and the core becomes critical. A fuel assembly 41S of zero burnup containing the mixed oxide fuel of TRU and thorium obtained in this addition in the upper fuel region 6 and the lower fuel region 8 is loaded into the core 20A, and this core 20A The process is repeated until the composition of the TRU obtained by reprocessing the fuel assembly 41S, which is the spent fuel assembly taken out from the above, becomes substantially constant.

  In FIG. 42, the weight of the current reactor discharge TRU added to the new fuel assemblies with zero burnup loaded in the recycle cores of each recycle generation is shown as characteristic 54, and the spent fuel assemblies taken out from these recycle cores are shown in FIG. The weights of the TRUs contained in are shown in the characteristics 53 respectively. In FIG. 43, the weight ratio of Pu-239 contained in the TRUs in the fuel assemblies of zero burnup loaded in the recycling cores of the respective recycling generations is characteristic 55, and the spent fuel extracted from these recycling cores is shown as characteristic 55. The weight percentages of Pu-239 contained in the TRUs in the aggregate are shown in characteristics 56, respectively. Each blanket region is not filled with its mixed oxide fuel and does not contain TRUs.

  Each recycle generation core from the RG1 core to the RG10 core is loaded with a fuel assembly including the current reactor discharge TRU and a removal burnup of 65 GWd / t. In these recycled generation cores, the current reactor discharge TRU is fissioned. In this embodiment, in the cores of these recycle generations, the void coefficient is negative as shown by the characteristic 57 shown in FIG. 44, and it cannot occur as a cause event in the BWR as shown by the characteristic 58. Even if it is assumed that all the control rods do not operate in this state, no positive reactivity is input.

  In the present embodiment, each effect produced in the second embodiment can be obtained. According to the present embodiment, TRU can be efficiently extinguished even in a nuclear fuel material having a TRU composition different from that of the second embodiment.

  WSYang et al., A Metal Fuel Core Concept for 1000MWt Advanced Burner Reactor. In GLOBAL '07 Boise, USA, September, 2007, P.52, fission of TRU recovered by reprocessing spent nuclear fuel in LWR The concept of sodium-cooled ABR that is reduced in weight is described. Furthermore, this document shows that the TRU generated from the light water reactor can be confined in the light water reactor, the ABR and the fuel cycle facility by operating the light water reactor and the ABR in operation at the same time, and the TRU is stored outside the reactor. It also describes that the amount of long-lived radioactive waste can be greatly reduced.

  However, since ABR uses nuclear fuel material enriched in TRU for depleted uranium and Na as a coolant, the neutron energy in the core increases. For this reason, enriched TRUs are reduced by fission, and at the same time, many TRUs are newly made from U-238. In order to accommodate the total amount of TRUs from the currently operating light water reactor in the ABR, it is necessary to construct one ABR per light water reactor. Since ABR is expected to have a higher power generation cost than a light water reactor, there is a risk that the economic efficiency is impaired as compared with the case of operating only a light water reactor.

  Therefore, instead of enriching depleted uranium with TRU, a fuel assembly containing nuclear fuel material with TRU enriched in thorium that does not newly generate TRU is disclosed in JP 2008-215818 A, which has a low neutron energy in the core. By loading and operating the TRU extinguishing furnace, it is possible to prevent new generation of TRU and promote TRU fission efficiency. For this reason, it becomes possible to fission the TRUs for three light water reactors with one TRU extinguishing reactor of the present embodiment, eliminating the need for Na-cooled ABR with high power generation costs, and greatly improving the economic efficiency.

  The core of a light water reactor according to embodiment 12 which is another embodiment of the present invention will be described in detail below with reference to FIGS. 45 to 46, Tables 7 and 8. The core of the present embodiment is a core 20C similar to that of the seventh embodiment. The electric power currently being operated is 1350 MW, and 872 fuel assemblies loaded in the core and 74 fuel rods per fuel assembly are provided. It is a core of ABWR having. This core is the core of the TRU extinguishing reactor. The configuration of the present embodiment will be described with respect to parts different from the seventh embodiment, and the description of the same configuration as that of the seventh embodiment will be omitted.

The cross section of the core 20C of the present embodiment is the same as in FIG. 35, and the cross section of the fuel assembly 41M loaded on the core 20C is the same as in FIG. The core 20C is loaded with a fuel assembly 41M having a nuclear fuel material region height of 3.71 m. The PWR core is loaded with a fuel assembly containing low enriched uranium. TRU having the composition shown in Table 7 (hereinafter referred to as current reactor discharge TRU) obtained by reprocessing a spent fuel assembly with a burn-up burnup of 50 GWd / t taken out from the core of the PWR, and a mixed oxide of depleted uranium A core loaded with a fuel assembly having fuel and having zero burnup is a TRU first generation recycle core (hereinafter referred to as RG1 core). The current reactor discharge TRU having the composition shown in Table 7 is added to the TRU obtained by reprocessing the spent fuel assembly taken out from the RG1 core in an amount that makes the core critical. The core loaded with the non-burning fuel assembly having the mixed oxide fuel of TRU and deteriorated uranium obtained by this addition is the TRU second generation recycle core (hereinafter referred to as RG2 core). Thereafter, each time TRU recycling is repeated, the TRU obtained by reprocessing the spent fuel assemblies taken out from the recycle cores of each generation is added to the TRU obtained by the amount that makes the core critical. A zero burnup fuel assembly having a mixed TRU and depleted uranium mixed oxide fuel is loaded into the core.

  Table 8 shows the TRU composition contained in the zero burnup fuel assemblies and the composition of the TRU contained in the spent fuel assemblies taken out of the RG8 core in each of the recycle generation cores from the RG1 core to the RG8 core. . In FIG. 45, the weight of the current reactor discharge TRU added to the zero burnup fuel assemblies loaded in the recycle cores of each recycle generation from the RG1 core to the RG8 core is extracted from these recycle cores to the characteristic 60. The weights of TRUs contained in the spent fuel assemblies are shown in characteristics 59, respectively. In FIG. 46, the weight ratio of Pu-239 contained in the TRU in the fuel assemblies of zero burnup loaded in the recycle cores of the respective recycle generations from the RG1 core to the RG8 core is shown as a characteristic 61, and these recycle cores The weight ratios of Pu-239 contained in the TRUs in the spent fuel assemblies taken out from are shown in the characteristics 62, respectively.

  Each recycle generation core from the RG1 core to the RG8 core is loaded with a fuel assembly including the current reactor discharge TRU and a removal burnup of 45 GWd / t. In these recycled generation cores, the current reactor discharge TRU is fissioned. In this embodiment, in the cores of these recycle generations, the void coefficient is negative as shown by the characteristic 63 shown in FIG. 47 and cannot be caused as a cause event in the BWR as shown by the characteristic 64. Even when it is assumed that all control rods do not operate in this state, no positive reactivity is introduced up to the RG7 core. The RG8 core has a positive input reactivity of less than $ 1 but has a negative void coefficient sufficient for safety.

  In this embodiment, all spent fuel assemblies from the RG1 core to the RG7 core are reprocessed, and the discharged TRU is transferred to the core of the next recycling generation together with the current reactor discharge TRU. Accordingly, in the RG8 core from the RG1 core, only 208 bodies that are taken out from the RG8 core every year are left as spent fuel assemblies. On the other hand, in order to supply about 14 tons of TRU used from the RG1 core to the RG8 core, about 8000 spent fuel assemblies are converted into low enriched uranium fuel assemblies currently used in ABWR. It needs to be reprocessed. By using the ABWR of this embodiment in combination with the ABWR that uses the currently operating low enriched uranium, the number of spent fuel assemblies that remain is the case of operating only the ABWR that uses the low enriched uranium fuel. It can be greatly reduced to about 2.6%.

  The core of 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 FIGS. 48 to 50 and Table 6. FIG. The core of the present embodiment is a core 20C similar to that of the twelfth embodiment. The electric power currently being operated is 1350 MW, and 872 fuel assemblies loaded in the core and 74 fuel rods per fuel assembly are provided. It is a core of ABWR having. This core is the core of the TRU extinguishing reactor. The configuration of the present embodiment will be described with respect to differences from the twelfth embodiment, and the description of the same configuration as the twelfth embodiment will be omitted.

  In the core 20C of the present embodiment, a fuel assembly 41M having a nuclear fuel material region height of 3.71 m is loaded. The PWR core is loaded with a fuel assembly containing low enriched uranium. The nuclear fuel material of the core 20E is loaded. Further, a fuel assembly containing low enriched uranium is loaded in the core of ABWR. TRU (hereinafter referred to as current reactor discharge TRU) having the composition shown in Table 6 obtained by reprocessing a spent fuel assembly with a burn-up burnup of 45 GWd / t taken out from the core of ABWR, and a mixed oxide fuel of thorium A core loaded with a fuel assembly having zero burnup is a TRU first generation recycle core (hereinafter referred to as RG1 core). The current reactor discharge TRU having the composition shown in Table 6 is added to the TRU obtained by reprocessing the spent fuel assembly taken out from the RG1 core in an amount that makes the core critical. The core loaded with the zero burnup fuel assembly having the mixed oxide fuel of TRU and thorium obtained by this addition is the TRU second generation recycle core (hereinafter referred to as RG2 core). Thereafter, each time TRU recycling is repeated, the TRU obtained by reprocessing spent fuel assemblies taken out from each generation of the recycling core is added to the TRU obtained by the amount that makes the core critical. A zero burnup fuel assembly having a mixed TRU and thorium mixed oxide fuel is loaded into the core.

  In FIG. 48, the weight of the current reactor discharge TRU added to the fuel assemblies of zero burnup loaded in the recycle cores of each recycle generation is shown as characteristic 66 in the spent fuel assemblies taken out from these recycle cores. The weight of each TRU is shown in characteristic 65, respectively. In FIG. 49, the weight ratio of Pu-239 contained in the TRU in the fuel assembly of zero burnup loaded in the recycling core of each recycling generation has a characteristic 67, and the spent fuel taken out from these recycling cores The weight percentages of Pu-239 contained in the TRUs in the aggregate are shown in characteristic 68, respectively.

  Each recycle generation core from the RG1 core to the RG4 core is loaded with a fuel assembly including the current reactor discharge TRU and a removal burnup of 45 GWd / t. In these recycled generation cores, the current reactor discharge TRU is fissioned. In this embodiment, in the cores of these recycle generations, the void coefficient is negative up to RG3 as shown by the characteristic 69 shown in FIG. 50, and cannot occur as an initiating event in ABWR as shown by the characteristic 70 shown in FIG. Even when the entire core is in a 100% void state and all control rods are assumed not to operate, no positive reactivity is introduced until RG3.

  In this embodiment, the TRU multiple recycling after RG4 needs to solve the safety problem, but the amount of the current furnace discharge TRU of RG1 is larger than that of ABR and other embodiments. Assume that the RG1 core and ABWR using the current low enriched uranium are used together, assuming that the RG2 core will not be implemented for the time being until the reprocessing technology of the fuel assembly containing thorium is established. 208 spent fuel assemblies are taken out from the RG1 core every year. On the other hand, in order to supply about 5.7 tons of TRU used in the RG1 core, it is necessary to reprocess about 3200 spent fuel assemblies of low enriched uranium fuel assemblies currently used in ABWR. By using the ABWR of the RG1 core and the ABWR using the low enriched uranium currently in operation, the number of remaining spent fuel bodies is approximately 6.5 when operating only the ABWR using the low enriched uranium fuel. % Can be significantly reduced. This embodiment is one of the methods that can realize a significant reduction in the number of spent fuel assemblies simply by changing the fuel assemblies in the currently operating ABWR. When the spent fuel assembly currently considered as one of the options for TRU disposal is to be disposed of as it is, the mixed oxide fuel pellet of TRU and thorium is better than the mixed oxide fuel pellet of TRU and uranium. Since it is considered to be much more chemically stable, it is an effective method until a reprocessing technique for a fuel assembly containing thorium is established.

  DESCRIPTION OF SYMBOLS 1 ... Boiling water reactor, 2, 2A, 2B ... Plenum, 3, 3A, 3B ... Neutron absorption member, 4 ... Neutron absorber filling area, 5, 5A ... Upper blanket area, 6, 6A ... Upper fuel area, 7, 7A ... Inner blanket region, 8, 8A ... Lower fuel region, 9,9A ... Lower blanket region, 10 ... Upper reflector region, 11 ... Lower reflector region, 12, 16, 16A ... Nuclear fuel material region, 14 ... Upper tie plate, 15 ... Lower tie plate, 20, 20A ... Core, 25 ... Core shroud, 27 ... Reactor pressure vessel, 34 ... Fuel region, 35A, 35C ... First region, 35B, 35D ... Second region, 41 , 41A to 41N, 41Q to 41S ... fuel assembly, 42, 47 ... control rod, 44, 44A to 44N, 44P to 44S ... fuel rod.

Claims (14)

In a light water reactor core having a nuclear fuel material region in which a plurality of fuel assemblies having nuclear fuel materials containing a plurality of isotopes of transuranium nuclides are loaded and the nuclear fuel material is present,
The nuclear fuel material region has a height in a range of 20 cm to 250 cm;
The fuel assembly includes a lower fuel support member that supports lower end portions of a plurality of fuel rods that form the nuclear fuel material region therein, and an upper fuel support member that supports upper end portions of the plurality of fuel rods. A plenum is formed inside the fuel rod and above the nuclear fuel material region;
A neutron absorbing member containing either boron or hafnium is disposed below the upper fuel support member and above the nuclear fuel material region;
The neutron absorbing member is disposed between adjacent plenums of the fuel rods;
A core of a light water reactor, wherein the neutron absorbing member is attached to the upper fuel support member.   The core of a light water reactor according to claim 1, wherein the sum of the cross-sectional areas of all the neutron absorbing members is in the range of 10 to 50% of the cross-sectional area of the fuel assembly lattice.   The core of the light water reactor according to claim 1, wherein another neutron absorbing member is disposed below the nuclear fuel material region.   The core of the light water reactor according to claim 1, wherein a neutron absorber filling region is formed in the fuel rod below the nuclear fuel material region.   The core of the light water reactor according to claim 4, wherein a length of the neutron absorber filling region in an axial direction of the core is in a range of 10 mm to 150 mm.   The outer diameter of the fuel rod in the neutron absorber filling region is equal to or greater than the outer diameter of the fuel rod in the nuclear fuel material region, and in the neutron absorber filling region of the adjacent fuel rod. The core of the light water reactor according to claim 4 or 5, wherein a distance between outer surfaces of the cores is in a range in which 1.3 mm or more is formed.   The outer diameter of the plenum portion of the fuel rod is smaller than the outer diameter of the nuclear fuel material region portion of the fuel rod within a range of 3 mm or more, and the axial length of the core of the plenum The core of the light water reactor according to claim 1, wherein is in a range of 400 mm to 2500 mm.   The core of a light water reactor according to any one of claims 1 to 7, wherein a ratio of plutonium-239 in all of the super uranium nuclides included in the nuclear fuel material region is in a range of 40% to 60%.   The core of a light water reactor according to any one of claims 1 to 7, wherein a ratio of plutonium-239 in all the ultrauranium nuclides included in the nuclear fuel material region is in a range of 5% or more and less than 40%. A plurality of fuel rods; a lower fuel support member that supports lower end portions of the plurality of fuel rods; an upper fuel support member that supports upper end portions of the plurality of fuel rods; and one of boron and hafnium. With a neutron absorbing member,
The plurality of fuel rods are formed above a nuclear fuel material region having a height in a range of 20 cm to 250 cm, in which a nuclear fuel material containing a plurality of isotopes of super uranium nuclides is present. Form a plenum
The neutron absorbing member is disposed above the nuclear fuel material region and below the upper fuel support member;
The neutron absorbing member is disposed between adjacent plenums of the fuel rods;
A fuel assembly, wherein the neutron absorbing member is attached to the upper fuel support member.   The fuel assembly according to claim 10, wherein a neutron absorbing material filling region is formed in the fuel rod below the nuclear fuel material region.   The fuel assembly according to claim 11, wherein the length of the neutron absorber filling region in the axial direction of the fuel assembly is in a range of 10 mm to 150 mm.   The outer diameter of the fuel rod in the neutron absorber filling region is equal to or greater than the outer diameter of the fuel rod in the nuclear fuel material region, and in the neutron absorber filling region of the adjacent fuel rod. The fuel assembly according to claim 11 or 12, wherein an interval between the fuel assemblies is within a range in which 1.3 mm or more is formed.   An outer diameter of the plenum portion of the fuel rod is within a range of 3 mm or more smaller than an outer diameter of the nuclear fuel material region portion of the fuel rod, and the axial direction of the fuel assembly of the plenum is The fuel assembly according to claim 10, wherein the length is in a range of 400 mm to 2500 mm.

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