JP5225362B2 - Fuel assembly - Google Patents

Description

  The present invention relates to a fuel assembly, and more particularly, to a 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 nuclear fuel material loaded in the core of a light water reactor. To sequentially move between isotopes. In actinide nuclides, there are so-called odd nuclei that have large fission cross sections for resonance regions and thermal neutrons, and so-called even nuclei that fission only for fast neutrons. The abundance of each isotope of the nuclide changes greatly. This change in the existence ratio is known to depend on the neutron energy spectrum at the loading position of the fuel assembly in the core.

  Current light water reactors use low enriched uranium as nuclear fuel. However, since natural uranium resources are finite, fuel assemblies used in light water reactors are converted into depleted uranium, which is a residue during uranium enrichment, and transuranium nuclides (hereinafter referred to as TRU) extracted from spent fuel assemblies in light water reactors. Need to be replaced with a recyclable fuel assembly that uses enriched nuclear fuel material. The TRU is recycled as a useful resource for a fairly long period of time when a commercial furnace is expected to be required, during which time the amount of TRU needs to increase or remain almost constant. Patent Document 1 describes a technology for realizing a breeder reactor in which the amount of fissile Pu is increased or substantially constant through the burning of nuclear fuel in a light water reactor that currently occupies most of commercial reactors. Japanese Patent No. 3428150 and R.A. TAKEDA et al., Proc. of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems. The light water reactor that realized the breeder reactor described in GLOBAL '95 Versailles, France, September, 1995, P.938 has a plurality of fuel rods densely arranged in a triangular lattice, and has a plurality of hexagonal cross sections. The fuel assemblies are arranged in the core. The core of this light water reactor has a dense arrangement of fuel rods to reduce the amount of water around the fuel rods and increase the proportion of resonance and fast energy neutrons, while lowering the height of the mixed oxide fuel region of the TRU and The blanket area of depleted uranium is placed in the area, and the negative void coefficient, which is a safety standard, is maintained. 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. While ensuring economic efficiency, the growth ratio is set to 1 or more.

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

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

  R.TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems.GLOBAL '07 Boise, USA, September, 2007, P.1725 includes TRUs obtained by reprocessing spent nuclear fuel. Proposed light water breeder and TRU extinguishing reactors that recycle TRU while keeping the ratio of each isotope present in TRU almost constant in order to realize a multiple cycle that repeatedly recycles as new nuclear fuel . This light water breeder reactor has a reactor core loaded with a fuel assembly which is recycled with the TRU amount kept constant or increased, and which has a high burnup and increased proliferation resistance. The TRU extinguishing reactor is to collect TRU sequentially while reducing the TRU by fission to avoid TRU becoming a long-lived radioactive waste when the light water reactor finishes its mission, excluding the last one core It is a nuclear reactor that fissions all TRUs.

Japanese Patent No. 3428150

R. TAKEDA et al., Proc. Of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems. GLOBAL '95 Versailles, France, September, 1995, P.938 G. A. Ducat et al., “EVALUATION OF THE PARFAIT BLANKET CONCEPT FOR FAST BREEDER REACTORS”, MITNE-157, January, 1974 R.TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems.GLOBAL '07 Boise, USA, September, 2007, P.1725

  R. TAKEDA et al., Proc. Of International Conference on Evaluation of Emerging Nuclear Fuel Cycle Systems. In light water reactors that recycle TRUs as described in GLOBAL '95 Versailles, France, September, 1995, P.938, the amount of TRUs should be increased with sufficient safety margin to satisfy design standards such as abnormal transient changes and accidents. By keeping it constant and effectively using it as seed fuel and burning all the depleted uranium, long-term stable energy supply is realized, and when the fission reactor finishes its mission and the TRU is no longer needed, all TRUs are fissioned. This makes it possible to realize a recycling furnace that can prevent TRU from becoming a long-lived waste. On the other hand, in recent years, there has been a movement to make safety thinking more and more severe. For example, when the core flow rate suddenly drops for some reason, an accident outside the framework of the design standard such as a complex event that all control rods cannot be inserted (Anticipated Transient Without Scram) (ATWS): There is a need for a core with a higher safety potential that has sufficient safety margin to cope with (transients without scram).

The object of the present invention is to provide a safety margin even when a composite event occurs that exceeds the design standard in which all control rods do not operate when the core flow rate is significantly reduced for some reason without impairing the economics of the light water reactor. Another object is to provide a fuel assembly that can be further increased.

The feature of the present invention that achieves the above object is that when the fuel assembly loaded in the core of the light water reactor has a burnup of zero, the proportion of Pu-239 in all the transuranium nuclides contained in the nuclear fuel material is 40%. The sum of the height of the lower blanket region located at the lowermost position within the effective fuel length and the height of the upper blanket region located at the uppermost position within the effective fuel length is in the range of not less than 60% and not more than 60%. And the height of the lower blanket region is in the range of 1.6 times to 12 times the height of the upper blanket region.

  According to the present invention, in the light water reactor in which the core is loaded with the plurality of fuel assemblies described above, the core flow rate is suddenly reduced for some reason during the operation of the light water reactor, and all the control rods are A sufficient safety margin can be ensured even when a composite event occurs that exceeds the design criteria that does not work. When such a composite event occurs, the void rate in the core rises rapidly, and the boiling start point of the coolant flowing into the core from below the core in a slightly subcooled state shifts to the lower end side of the core, The power distribution in the core axis direction shifts to the lower end side of the core. For this reason, excess neutrons that shift to the lower end of the core are shifted to the lower end of the core by the neutron absorber in the neutron absorber filling region of the safety rod, the upper end of which is located near the lower end of the core. Excess neutrons can be absorbed. As a result, the capacity of the coolant that can be supplied from the emergency high-pressure core water injection system can automatically reduce the output to the power that can cool the fuel assembly in the core, even in complex events exceeding the design standard. Sufficient safety potential can be maintained. As described above, the present invention can improve the safety margin even when the above composite event occurs without impairing the economics of the light water reactor.

According to the present invention, even if a composite event occurs that exceeds the design standard in which all control rods do not operate when the core flow rate is significantly reduced for some reason without impairing the economics of the light water reactor, The output can be automatically reduced to an output capable of cooling the fuel assembly in the core by the capacity of the coolant that can be supplied from the core water injection system, and the safety margin can be further increased.

The average power distribution in the core axis direction during the rated power operation of the core with a fissile Pu multiplication ratio of 1.01 and the core flow rate decreased to 4 kt / h, which is the flow rate of coolant that can be supplied by the emergency high-pressure water injection system. It is explanatory drawing which shows the average output distribution of a core axis direction in the case. The average power distribution in the core axis direction during the rated power operation of the core with a fissionable Pu multiplication ratio of 1.01 and the core flow rate decreased to 4 kt / h, which is the flow rate of coolant that can be supplied by the emergency high-pressure water injection system. It is explanatory drawing which shows the void ratio distribution of the core height direction in the case. It is explanatory drawing which shows the thermal neutron flux distribution of the axial direction of the core of a fissile Pu multiplication ratio 1.01. It is explanatory drawing which shows the relationship between the core flow volume and output in the core of a fissile Pu multiplication ratio 1.01. It is explanatory drawing which shows the relationship between the void coefficient with respect to the height of the upper blanket area | region in the core of a fissile Pu multiplication ratio 1.01, and the ratio of a lower blanket area | region and an upper blanket area | region. It is explanatory drawing which shows the relationship of the sum of the height of the upper and lower blanket area | region with respect to the height of the upper blanket area | region in the core of a fissionable Pu multiplication ratio 1.01. The core power direction when the average power distribution in the core axis direction and the core flow rate during the rated power operation of the BWR core that extinguishes the TRU drops to 4 kt / h, which is the flow rate of coolant that can be supplied by the emergency high-pressure water injection system It is explanatory drawing which shows average output distribution. Core power axis direction when the average power distribution in the core axis direction during rated power operation of the BWR core that extinguishes TRU, and the core flow rate when the core flow rate drops to 4 kt / h, which is the flow rate of coolant that can be supplied by the emergency high-pressure water injection system It is explanatory drawing which shows the void ratio distribution. It is explanatory drawing which shows the thermal neutron flux distribution of the core axis direction of the BWR core which extinguishes TRU. It is explanatory drawing which shows the relationship between the core flow volume and output of a BWR core which extinguishes TRU. BRIEF DESCRIPTION OF THE DRAWINGS It is a longitudinal cross-sectional view of the light water reactor to which the core 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. It is explanatory drawing which showed the positional relationship of the axial direction of the fuel rod in a fuel assembly shown in FIG. 13, and a Y-shaped control rod. FIG. 12 is an explanatory diagram showing the arrangement of fuel assemblies in the equilibrium core state of the core shown in FIG. 11. 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 fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core of the core shown in FIG. It is explanatory drawing which showed arrangement | positioning of each fuel rod from which the enrichment degree of the fissile Pu in the cross section of the fuel assembly shown in FIG. 13 differs. It is explanatory drawing which shows the structure of the axial direction of the fuel rod in the fuel assembly shown in FIG. 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 the equilibrium core of a light water reactor according to a second embodiment. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core of the core shown in FIG. It is a cross-sectional view of the fuel assembly lattice in the core of the light water reactor according to embodiment 3, which is another embodiment of the present invention. FIG. 6 is an explanatory diagram showing the arrangement of fuel assemblies in the equilibrium core of a light water reactor according to a third embodiment. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core of the core shown in FIG. It is a cross-sectional view of the fuel assembly lattice in the core of the light water reactor according to embodiment 4, which is another embodiment of the present invention. FIG. 27 is an explanatory diagram showing an enrichment distribution of fissile Pu in the axial direction of a new fuel assembly loaded on the equilibrium core of the core shown in FIG. 26. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded into the balanced core of the light water reactor of Example 5 which is another Example of this invention. It is explanatory drawing which shows the enrichment distribution of the fissile Pu in the axial direction of the new fuel assembly loaded to the equilibrium core of the light water reactor of Example 6 which is another Example of this invention. It is explanatory drawing which showed the positional relationship of the height direction of the fuel rod in the fuel assembly of Example 6 which is another Example of this invention, and a Y-shaped control rod.

  Cooling water (coolant) that cools the fuel assemblies in the BWR core flows into the core from below as subcooled water of about 10 ° C., and is saturated with water and two phases of steam and water while cooling the fuel assemblies Become a flow. This cooling water is a two-phase flow having a void volume ratio of about 60 to 80% at the outlet of the core. Therefore, the distribution in the axial direction of the hydrogen atoms that greatly contribute to the deceleration of neutrons decreases from the bottom to the top of the core. For this reason, when a fuel assembly having a concentration distribution in one region in the axial direction is loaded on the core of the BWR, a large output peak is formed in the lower region of the core. If the coolant flow rate in the core decreases for any reason, the boiling start point of the coolant will move further downward than when the reactor is operating at the rated power and rated flow, and the power peak will increase further. Move down the core.

  The inventors fully considered the above-mentioned BWR core characteristics, and included Pu-239 in all TRUs that contain nuclear fuel material obtained by reprocessing and are included at the time of zero burnup. In the core of a light water reactor in which a plurality of fuel assemblies in the range of 40% to 60% is loaded, when the upper end of the neutron absorber filling region of the control rod is arranged near the lower end of the core, (1 ) The sum of the height thickness of the upper blanket area of the core and the height of the lower blanket area is in the range of 250 mm to 600 mm, and the height of the lower blanket area is 1.6 times to 12 times that of the upper blanket area. (2) The height of the lower blanket region is higher than that of the upper blanket region, and the height of the upper blanket is within the range of 30 mm to 105 mm. (3) The height of the lower blanket region is higher than that of the upper blanket region, and the height of the upper fuel region including Pu of the core is 10 mm or more and 25 mm or less than the height of the lower fuel region including Pu. It has been newly found that the safety margin can be further increased without impairing the economics of the light water reactor by applying any one of the configurations of increasing within the range. That is, by applying one of the configurations (1), (2), and (3), the coolant in the core is lost for some reason, and all the control rods are inserted into the core for some reason. Even when a composite event that exceeds the design criteria that would make it impossible to do so occurs, the void ratio in the core rapidly increases when the flow rate of the coolant supplied to the core (core flow rate) decreases rapidly. The power distribution in the axial direction of the core shifts to the lower end side of the core where the neutron absorber is disposed, and surplus neutrons in the core are automatically absorbed by the neutron absorber. For this reason, the reactor power is automatically reduced to the reactor power that can be cooled by the coolant flow rate supplied to the core by the emergency high-pressure core water injection system that is automatically started in an emergency. Thus, the inventors have newly found that by applying any one of the configurations (1), (2), and (3), it is possible to provide a core of a light water reactor in which the safety potential is further increased. Further, the increase in safety margin as described above is as follows: (4) The height of the lower blanket region is made higher than that of the upper blanket region, and the neutron absorbing material is disposed at a position where surplus neutrons generated at the time of the accident gather. The inventors have found that this can also be achieved.

  The safety margin can be further improved by combining several configurations of (1), (2), (3) and (4). 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), the combination of the other two configurations including (3), and the combination of the other two configurations including (4). . Further, when (4) is further combined with the combination of (1), (2) and (3), the safety margin is the largest among several combinations including the configurations of (1) to (4). growing.

  Furthermore, the inventors include nuclear fuel material obtained by reprocessing, and the proportion of Pu-239 in all TRUs included at the time of zero burnup is 5% or more and less than 40%. In a light water reactor core loaded with a plurality of fuel assemblies in the range, when the upper end of the neutron absorber filling region of the control rod is arranged near the lower end of the core, (5) the lower end of the fuel region in the core is the core And the height of the upper blanket region is in the range of 20 mm to 100 mm, and (6) the upper blanket region is provided and the height of the upper fuel region of the core By applying one of the configurations in which the height of the lower fuel region is higher than the height of the lower fuel region within the range of 10 mm or more and 25 mm or less, the performance and economy of the TRU multiple recycling design target are not impaired. Newly found that it is possible to further increase the overall margin. By combining the configuration of (6) with the configuration of (5), the safety margin is further improved compared to the core of a light water reactor having the configuration of (5) or the configuration of (6) alone.

  The present invention aims to improve the safety of a recycled light water reactor using nuclear fuel material containing TRU obtained by reprocessing. In the light water reactor disclosed in Japanese Patent No. 3428150, the present invention is considered to improve the performance as a breeding reactor (light water breeding reactor) and to dispose of it as long-lived radioactive waste when it becomes unnecessary. In the case where all TRUs other than one core TRU are fissioned while using the existing TRU as nuclear fuel material, safety is maintained even in the case of multiple accidents exceeding the design standard, and multiple TRU recycling is continued. It was made to make it possible.

  A light water reactor core with improved performance as a breeding reactor will be described. For example, a light water breeding reactor in which the remaining ratio of fissile Pu in BWR 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 the breeder reactor, the mass of hydrogen atoms forming the water used as the coolant in the light water reactor is smaller than that normally used as the coolant in the breeder reactor. Larger, it is necessary to reduce the ratio of coolant per unit volume of nuclear fuel material. 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 and the void coefficient, which is a safety index, deteriorates. . 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%.

  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 TRU from becoming a long-lived waste, when the nuclear fuel material is recycled in a state where the ratio of Pu-239 contained in all TRUs is 40% or more, the rate at which TRU decreases is slow. It takes too much time to consolidate 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%.

  Here, an outline of the parfait-type core will be described. The parfait-type core is a new fuel assembly to be loaded (with a burnup of zero). The lower blanket area, lower fuel area, internal blanket area, upper fuel area, and upper blanket area are arranged in this order from the lower end to the upper end. The fuel assembly arranged in is used. For this reason, also in the parfait-type core, a lower blanket region, a lower fuel region, an inner blanket region, an upper fuel region, and an upper blanket region are formed from the lower end portion toward the upper end portion. The lower fuel region and the upper fuel region contain TRU oxide fuel (or a mixed oxide fuel of TRU oxide and uranium oxide).

  The present invention is directed to the above-described recycle type light water reactor and the core of the light water reactor. In the following, the results of investigations by the inventors will be described.

  First, the examination results of the inventors for the core of the light water breeder reactor will be described below. An example of a light water breeder reactor core is a BWR core with a power ratio of 1350 MW and a breeding ratio of 1.01 with 720 fuel assemblies loaded in the core and 271 fuel rods per fuel assembly. I will explain.

  This BWR core is disclosed in Japanese Patent No. 3428150 and R.A. 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.C. TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems.The core described in GLOBAL '07 Boise, USA, September, 2007, P.1725 is subject to abnormal transient changes and accidents within the design criteria. However, when the core flow rate is suddenly reduced for some reason, and a complex event such as the failure of all control rods occurs, This is not always sufficient, and in some cases, TRU recycling may have to be stopped halfway. That is, it may be impossible to continue multiple recycling.

  In order to continue TRU recycling while maintaining a sufficient safety potential in the BWR core, it is necessary to maintain the void coefficient within a predetermined range. The inventors include a plurality of Pu-239 in the range of 40% or more and 60% or less in all TRUs containing the nuclear fuel material obtained by reprocessing and included at the time of zero burnup. We investigated safety margin improvement measures for light water reactor cores with lower and upper blanket regions loaded with fuel assemblies. As a result of this study, the inventors have found that when the core flow rate suddenly decreases for some reason, which is a function unique to BWR, the void ratio in the core increases rapidly, and becomes slightly sub-cooled and enters the core from below. As the boiling point of the coolant flowing into the core shifts to the lower end of the core, the axial power distribution of the core shifts to the lower end of the core, so multiple accidents can occur by placing neutron absorbing materials near the lower end of the core. It was newly found that sufficient safety potential can be maintained even at times, and as a result, multiple recycling of TRUs can be realized. Based on this knowledge, the inventors have a new knowledge that the safety potential can be increased while maintaining the proliferation ratio of TRU by any of the above-mentioned (1), (2), (3) and (4). It was found by. In the core of the target light water reactor, the control rod is inserted into the core from below.

In FIG. 1, the average power distribution in the axial direction of the core at the rated power operation of the core having a multiplication ratio of fissile Pu of 1.01 is represented by the characteristic 1, and the core flow rate is the coolant for the emergency high pressure core water injection system. The average power distribution in the core axis direction when the flow rate is reduced to 4 kt / h is represented by characteristic 2. In FIG. 2, the average void fraction distribution in the core axis direction corresponding to the characteristic 1 is represented by the characteristic 3, and the average void fraction distribution in the core axis direction corresponding to the characteristic 2 is represented by the characteristic 4. As the core flow rate suddenly drops from the rated 21 kton / h to 4 kton / h, the void ratio distribution increases rapidly from characteristic 3 to characteristic 4, and the boiling start point also moves to the lower end side of the core. Transition. For this reason, the power distribution in the core axis direction also shifts to the lower end side of the core, as shown by characteristics 1 to 2. If the core flow rate is extremely reduced so far, a large output peak may occur in the reflector (cooling water) at the bottom of the core, and a positive reactivity may be input to the core. Japanese Patent No. 3428150 and R.A. TAKED
A 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 the described core, a safety rod, which is a kind of control rod pulled out of the core during rated power operation, as is normally done in a core with a relatively low core height of 2 m or less. Is in a state of being pulled out to a position that does not affect the negative reactivity of the core (a position 20 cm to 30 cm below the lower end of the core). The thermal neutron flux distribution in the core axis direction in this state is indicated by characteristic 5 in FIG. Therefore, as described above, when the reactor is in operation, the upper end of the neutron absorbing material filling region of the safety rod drawn out below the lower end of the core is positioned at the lower end of the core. When it decreases, it absorbs the extra neutrons that shift to the bottom of the core. The thermal neutron flux distribution in the core axis direction at this time is indicated by characteristic 6 in FIG. However, if the upper end of the safety rod is positioned at the lower end of the core, the reactivity of the core will decrease.

Increasing the height of the fuel region can be considered as a measure to solve the lack of reactivity.
However, in this measure, the volume ratio of the blanket region per unit volume of the fuel region is lowered, the fissile Pu multiplication ratio is lowered, and the design target of the fissile Pu multiplication ratio is not satisfied. In order to increase the fissile Pu multiplication ratio, it is necessary to further increase the heights of the upper and lower bracket regions in the core. An increase in the height of these regions results in a decrease in the neutron leakage rate in the core axis direction and worsens the void coefficient, which is an important safety indicator.

Loaded with multiple fuel assemblies that contain nuclear fuel material obtained by reprocessing and have a Pu-239 ratio of 40% to 60% in all TRUs at the time of zero burnup As a result of the investigations conducted by the inventors on the core of a light water reactor, when the neutron absorbing material is arranged near the lower end of the core, the height of the upper blanket region is reduced and the height of the lower blanket region is increased. By doing so, it was found that the void coefficient can be avoided.
In the light water breeding reactor, the vicinity of the lower end of the core described above refers to a position between the lower end of the core and a position lower than the lower end, for example, 5 mm, and this lower blanket is formed when a lower blanket region is formed in the core. The area is also included near the lower end. The height of the lower blanket area is higher than that of the upper blanket area, that is, the sum of the heights of the upper and lower blanket areas is 250 mm or more, and the height of the lower blanket area is that of the upper blanket area. By making it 1.6 times or more, as shown in FIG. 4, when the core flow rate is significantly reduced for some reason while satisfying all the constraints and maintaining the growth ratio of 1.01, Even in the event of a combined event exceeding the design standard of not operating, the output can be automatically reduced to the output that can cool the fuel assembly in the core with the capacity of the coolant that can be supplied to the core from the emergency high-pressure core water injection system . For this reason, there is a fuel assembly that contains nuclear fuel material obtained by reprocessing and has a ratio of Pu-239 in all TRUs included at the time of zero burnup within a range of 40% to 60%. The safety margin can be improved in the core of the loaded light water reactor. When the sum of the respective heights of the upper and lower blanket regions exceeds 600 mm, or when the height of the lower blanket region is greater than 12 times that of the upper blanket region, the spent fuel assembly removed from the core The ratio of Pu-239 contained in all TRUs in the nuclear fuel material present in the body is higher than that contained in the new fuel assembly having zero burnup. For this reason, when the core flow rate is increased in order to keep these values at the same level, the pressure loss of the core exceeds the design standard, so that the structural design of the fuel assembly becomes difficult. Accordingly, the sum of the heights of the upper and lower blanket regions is included in the range of 250 mm to 600 mm.

FIG. 5 shows the change in the void coefficient when the height of the upper blanket region is changed in the core of a light water reactor having a fissile Pu multiplication ratio of 1.01, as a characteristic 31, and the lower part of the upper blanket region with respect to the height. The ratio of the height of the blanket area is indicated by the characteristic 32. As shown in FIG. 5, the height of the upper blanket region is 105 mm or less, the height of the lower blanket region is 1.6 times or more the height of the upper blanket region, and the void coefficient is −2 × 10 ⁻⁴ Δk. It turned out to be more negative than / k /% void. When the negative absolute value of the void coefficient is increased, the core flow rate is greatly reduced, that is, the void ratio in the core is greatly increased, and the composite event exceeds the design standard where all control rods do not operate. However, the output automatically decreases to the output at which the fuel assembly can be cooled by the coolant supply capacity of the emergency high pressure core water injection system. When the height of the upper blanket region is less than 30 mm, the output of the fuel pellet located near the upper end of the upper blanket region is greatly affected by the thermal neutron flux in the upper reflector and exceeds the design standard. End up. Therefore, the height of the upper blanket region is set within a range of 30 mm or more and 105 mm or less.

  In FIG. 6, the sum of the respective heights of the upper and lower blanket regions when the height of the upper blanket region is changed is indicated by a characteristic 33. It was found that by making the height of the upper blanket region 105 mm or less, the sum of the heights of the upper and lower blanket regions becomes 250 mm or more. Further, when the reactor is operated with the upper end of the neutron absorbing material filling region of the safety rod positioned in the vicinity of the lower end of the core, there is a possibility that the boron-10 that is the neutron absorbing material included in the safety rod is lost earlier. . For this reason, in some cases, it is also effective to place pellets containing neutron absorbing materials such as boron, gadolinia, Dy, Sm, and Eu below the lower blanket region in the fuel rods included in each fuel assembly. It is.

  The height of the lower blanket region is set higher than that of the upper blanket region, and the height of the upper fuel region including the core TRU is set higher than the height of the lower fuel region including the TRU by 10 to 25 mm. ing. By making the height of the upper fuel region 10 mm or more higher than the height of the lower fuel region, the safety margin of the core can be improved even when the above-mentioned combined event occurs. When the height of the upper fuel region is higher than the height of the lower fuel region by more than 25 mm, the output of the upper fuel region becomes too high and exceeds the output design standard.

Next, R.D. 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 will be described. A description will be given of another BWR core, which is an example of a TRU extinguishing core, in which an electric output is 1350 MW and 720 fuel assemblies having 397 fuel rods per fuel assembly are loaded.

When TRU recycling is repeated for the purpose of reducing TRU, that is, the proportion of Pu-239 in all TRUs that contain nuclear fuel material obtained by reprocessing and are included at the time of zero burnup is 5% or more When loading fuel assemblies within the range of less than 40% into the core is repeated every operation cycle, leakage from the core occurs due to the neutron moderating effect of the hydrogen atoms that form water in the reflector region below the core. The generated fast neutrons are decelerated and a large output peak of thermal neutrons is generated. In order to prevent a problem in which the output of each fuel pellet located near the lower end of the lower fuel region of each adjacent fuel assembly in the core due to neutron flow continuity exceeds the design standard value, R.D. TAKEDA et al., Proc. Of International Conference on Advanced Nuclear Fuel Cycles and Systems.GLOBAL '07 Boise, USA, September, 2007,
In the TRU extinguishing furnace described in P.1725, a lower blanket region having a height of about 20 mm is attached below the lower fuel region. By applying any one of the above-mentioned methods (5) and (6) for increasing the safety potential found by the inventors, the upper end of the neutron absorber filling region of the control rod is positioned near the lower end of the core. Since it is possible to suppress the generation of thermal neutron output peaks in the reflector (cooling water) near its lower end, it is not necessary to provide a lower blanket region at the bottom of the core. That is, the lower end of the fuel region, particularly the lower fuel region, coincides with the position of the lower end of the core. In the TRU extinguishing reactor, the vicinity of the lower end of the core where the upper end of the neutron absorbing material filling region of the control rod is positioned means between the lower end of the core and a position lower than the lower end, for example, 5 mm.

  In FIG. 7, the average power distribution in the core axis direction during rated power operation is indicated by the characteristic 11, and the core flow rate is reduced to 4 kt / h, which is the flow rate of cooling water that can be supplied by the emergency high pressure core water injection system. The average power distribution in the direction of the core axis at is indicated by a characteristic 12. In FIG. 8, the average void ratio distribution in the core axis direction corresponding to the characteristic 11 is indicated by a characteristic 13, and the average void ratio distribution in the core axis direction corresponding to the characteristic 12 is indicated by a characteristic 14. As the core flow rate rapidly decreases from the rated 20 kton / h to 4 kton / h, the average void fraction distribution in the core axis direction increases rapidly from characteristic 13 to characteristic 14 shown in FIG. At the same time, the boiling start point shifts to the lower end side of the core, whereby the power distribution in the core axis direction is also shifted from the characteristic 11 shown in FIG. If the core flow rate is extremely reduced so far, a large output peak may occur in the reflector below the core, and positive reactivity may be input to the core. The safety rod pulled out of the core during rated power operation has the effect of introducing negative reactivity into the core, as is usually done in cores with a relatively low core height of 2 m or less. It waits in the state pulled out to the position (30 cm lower position from the lower end of a core) which does not give. In the core in which the neutron absorber filling region of the safety rod is not disposed near the lower end of the core and the lower blanket region having a height of 20 mm is provided, the thermal neutron flux distribution in the core axis direction is shown by the characteristic 15 in FIG. Yes. Therefore, when the core flow rate suddenly decreases when the upper end of the neutron absorber filling region of the safety rod pulled out below the lower end of the core during operation of the reactor is positioned at the lower end of the core, Excess neutrons that shift downward can be absorbed by the safety rod. The thermal neutron flux distribution in the core axis direction at this time is indicated by a characteristic 16 in FIG. As shown in FIG. 10, the capacity of the coolant that can be supplied from the emergency core injection system to the core even at the time of a composite event exceeding the design standard in which all control rods do not operate when the core flow rate is significantly reduced for some reason. Thus, the output can be automatically reduced to an output that can cool the fuel assembly in the core. For this reason, there is a fuel assembly that contains nuclear fuel material obtained by reprocessing and has a ratio of Pu-239 in all TRUs included at the time of zero burnup within a range of 5% or more and less than 40%. The safety margin can be improved in the core of the loaded light water reactor.

  In the core of the TRU extinguishing reactor, the safety margin of the core can be improved by setting the height of the upper blanket region to 100 mm or less. However, when the height of the upper blanket region is less than 20 mm, as in the case of the light water breeder reactor, it is greatly influenced by the thermal neutron flux in the upper reflector and located near the upper end of the upper blanket region. The output of fuel pellets exceeds the design standard. Therefore, the height of the upper blanket region is set within a range of 20 mm or more and 100 mm or less.

  In the core of the TRU extinguishing reactor, the reason why the height of the upper fuel region is higher than the height of the lower fuel region within the range of 10 mm or more and 25 mm or less is because these values are set in the light water breeding reactor The same.

  Hereinafter, embodiments of the present invention will be described in detail 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 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 power scale to which this embodiment can be applied.

  An outline of a BWR, which is a light water reactor for an electric output of 1350 MW, to which the core 20 of the present embodiment is applied will be described with reference to FIG. In the BWR 19, a reactor core 20, a steam separator 21, and a steam dryer 22 are disposed in a reactor pressure vessel 27. The core 20 is a parfait-type core and is surrounded by a core shroud 25 in the reactor pressure vessel 27. A plurality of control rods 2 are arranged at positions where they can be inserted into the core 20. These control rods 2 are inserted into the core 20 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 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 19 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 1 are loaded in the core 20. Three fuel assemblies 1 are provided with Y-shaped control rods 2 at a ratio of one, and 223 control rods 2 are arranged. About 1/6 of the 223 control rods 2 are control rods (output adjustment control rods) that adjust the output of the reactor by being inserted into or pulled out from the core 20 of the operating BWR 19, and the rest About 5/6 is a control rod 2 (hereinafter referred to as a safety rod) that is inserted into the core 20 when the nuclear reactor is stopped while being pulled out of the core 20 of the operating BWR 19. The fuel assembly 1 is divided into five areas of an active fuel length portion from the upper end to the lower end: 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. They are formed sequentially (see FIG. 17). A core 20 loaded with a plurality of fuel assemblies 1 has 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 from the upper end toward the lower end. Five regions of an internal blanket region 7A formed by 7, a lower fuel region 8A formed by the lower fuel region 8 and a lower blanket region 9A formed by the lower blanket region 9 are sequentially arranged. The regions 5A, 6A, 7A, 8A, and 9A are located at the same positions as the regions 5, 6, 7, 8, and 9 of the fuel assembly 1 in the axial direction of the core 20.

  In the fuel assembly 1, as shown in FIG. 13, 271 fuel rods 3 having a diameter of 10.1 mm are arranged in a regular triangular lattice in a channel box 4 which is a hexagonal cylinder. The cross section of the fuel assembly 1 has a hexagonal shape, and a gap between the plurality of fuel rods 3 included in the fuel assembly 1 is 1.3 mm. A plurality of fuel pellets (not shown) made of nuclear fuel material are arranged in the cladding tube 36 of each fuel rod 3 so as to be aligned in the axial direction. Nine fuel rods 3 are arranged in the outermost fuel rod row. As shown in FIG. 14, the fuel rod 3 is formed using a nuclear fuel material obtained by reprocessing in a cladding tube 36 whose lower end is sealed with a lower end plug 33 and whose upper end is sealed with an upper end plug 35. Filled with multiple fuel pellets. The effective fuel length 14 is an area where these fuel pellets are filled. A gas plenum 34 is formed between the upper end of the active fuel length 14 and the upper end plug 35 in a sealed cladding 36. In order to increase the effective fuel length 14 of each fuel rod 3, the five regions of the upper blanket region 5, the upper fuel region 6, the inner blanket region 7, the lower fuel region 8 and the lower blanket region 9 are sequentially moved from the upper end to the lower end. Forming.

The control rod 2 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 absorber rods filled with B ₄ C as a neutron absorber, and is arranged around the tie rods with an interval of 120 degrees. The control rod 2 is provided with a follower portion 16 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 core 20. In the control rod 2, the neutron absorber filling region 15 formed by the neutron absorber filled in each neutron absorber rod exists below the follower portion 16 (see FIG. 14).

  When the BWR 19 is operated at the rated output, the safety rod, which is the control rod 2 that is fully pulled out, is positioned such that the upper end of the neutron absorber filling region 15 is located at the lower end of the effective fuel length 14 of the fuel rod 3. The core 20 is pulled out (see FIG. 14).

  When the BWR 19 is in operation, the coolant in the downcomer is pressurized by the rotation of the internal pump (coolant supply device) 26 and supplied to the core 20. The coolant supplied into the core 20 is guided into each fuel assembly 1 and heated by heat generated by the fission of the fissile material, and a part thereof becomes steam. The coolant in the gas-liquid two-phase flow state is led from the core 20 to the steam / water separator 21 to separate the steam. The water is further removed from the separated steam by the steam dryer 22. The steam from which moisture has been removed is supplied to the turbine (not shown) through the main steam pipe 23, and the turbine is rotated. A generator (not shown) connected to the turbine is rotated to generate electric power. The steam discharged from the turbine is condensed in a condenser (not shown) to become condensed water. The condensed water (feed water) is guided into the reactor pressure vessel 27 through the feed water pipe 24. The liquid coolant separated by the steam separator 22 is mixed with the above-mentioned water supply in the downcomer and is pressurized by the internal pump 26 again. The rated flow rate of BWR19 is 22 kton / h.

  The arrangement of the fuel assembly 1 in the core 20 in the state of an equilibrium core will be described with reference to FIG. The fuel assembly 1E of the fifth operation cycle having the longest residence time in the reactor is arranged in the outermost peripheral region of the core having a low neutron importance. A fuel assembly 1A having the highest neutron infinite multiplication factor and a stay in the reactor in the first cycle is loaded in the outer region located inside thereof, and the power distribution in the radial direction of the core is flattened. I am trying. The fuel assemblies 1B, 1C, and 1D of the second, third, and fourth cycles are arranged in the core inner region. With this arrangement, the power distribution in the core inner region is flattened. The fuel assemblies 1A, 1B, 1C, 1D, and 1E are the fuel assemblies 1 shown in FIG. 13 and FIGS. 17 and 18 described later, respectively. Lower tie plates (not shown) of these fuel assemblies are supported by a plurality of fuel support fittings (not shown) provided on a core support plate (not shown) arranged at the lower end of the core 20. The A coolant passage for guiding the coolant to the fuel assembly is formed in the fuel support fitting, and an orifice (not shown) installed in the fuel support fitting is disposed at the inlet of the coolant passage. The core 20 is formed with two regions of the outermost peripheral region 6 and the inner region 1 located inside thereof in the radial direction (see FIG. 16). The diameter of the orifice located in the outermost peripheral area 6 where the output of the fuel assembly 1 is small is smaller than the diameter of the orifice located in the inner area 7. As shown in FIG. 17, the fuel assembly 1 has an active fuel length portion, from the upper end toward the lower end, and as shown in FIG. 17, the upper blanket region 5, the upper fuel region 6, the inner blanket region 7, the lower fuel region 8 and the lower blanket Five regions of region 9 are sequentially formed. The height of each region is as follows. The height of the upper blanket region 5 (upper blanket region 5A) is 70 mm, the height of the upper fuel region 6 (upper fuel region 6A) is 241 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 225 mm, and the height of the lower blanket region 9 (lower blanket region 9) is 280 mm. When the fuel assembly 1 was a new fuel assembly with zero burnup, all the fuel rods 3 of the fuel assembly 1 were filled with depleted uranium oxide pellets in three blanket regions. The upper fuel region 6 and the lower fuel region 8 are filled with a mixed oxide fuel in which deteriorated uranium is mixed at a ratio of 172 when the weight of TRU is 100 on average. The weight ratio of fissile Pu to the total weight of TRU and depleted uranium in the mixed oxide fuel, that is, the average fissile Pu enrichment is 18 wt%. The TRU is a material extracted from the nuclear fuel material contained in the spent fuel assembly 1 by reprocessing. Each blanket region is not filled with the mixed oxide fuel. In addition, instead of depleted uranium, depleted uranium oxide pellets recovered from spent fuel assemblies may be used for each blanket region. The fuel assembly 1 includes five types of fuel rods 3 shown in FIG. These fuel rods 3 are fuel rods 3A to 3E. The fuel rods 3A to 3E are arranged in the fuel assembly 1 as shown in FIG. The mixed oxide fuel filled in each of the upper fuel region 6 and the lower fuel region 8 of each of the fuel rods 3A to 3E is rich in fissionable Pu in the fuel rod 3A in the state of a new fuel assembly (zero burnup). The enrichment is 10.7 wt%, the fuel rod 3B has an enrichment of 13.5 wt%, the fuel rod 3C has an enrichment of 16.8 wt%, the fuel rod 3D has an enrichment of 18.2 wt%, and The enrichment of the fuel rod 3E is 19.5 wt%. The average fissile Pu enrichment in the upper and lower fuel regions 6, 8 is both 18 wt%. Although there is no TRU in each blanket region of each fuel rod 3, each mixed oxide fuel in the upper fuel region 6 and lower fuel region 8 of each fuel rod 3 is shown in Table 1 in a state where the burnup is zero. Contains TRU in composition. The fuel assembly 1 is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 44 wt%. Table 1 shows that after the fuel assembly 1 is removed from the core 20, it stays outside the reactor for a total of 3 years, 2 years at the fuel storage pool and fuel reprocessing facility, and 1 year at the fuel production facility. It is the composition of TRUs present in the nuclear fuel material contained in the new fuel assembly loaded into the core obtained by reprocessing the nuclear fuel material in the fuel assembly. The nuclear fuel material obtained by reprocessing and present in the new fuel assembly 1 includes a plurality of TRU isotopes of TRUs shown in Table 1. In this embodiment, the composition of TRU present in the fuel assembly taken out from the core when the operation in the operation cycle is completed, and the fuel assembly newly loaded in the core in a state where the operation in the operation cycle can be started. TRU multiple recycling is realized in which the composition of TRU present in the body is substantially constant.

According to this embodiment, the sum of the height of the upper blanket region and the height of the lower blanket region is 350 mm and the height of the lower blanket region is four times the height of the upper blanket region. By positioning the upper end of the neutron absorber filling region 15 of the safety rod, which is the control rod 2 that is fully pulled out at the time, at the lower end of the effective fuel length 14 of the fuel rod 3 (lower end of the core 20) (see FIG. 14), even if the core flow rate suddenly drops for some reason during operation of the BWR19, and a composite event occurs that exceeds the design criteria in which all control rods do not operate, a sufficient safety margin can be secured. it can. When such a composite event occurs, the void rate in the core rises rapidly, and the boiling start point of the coolant flowing into the core from below the core in a slightly subcooled state shifts to the lower end side of the core, The power distribution in the core axis direction shifts to the lower end side of the core. Therefore, the upper end absorbs excess neutrons shifted to the lower end of the core by B ₄ C in the neutron absorber filling region 15 located at the lower end of the core, that is, the lower end of the lower blanket region 9A. Can do. As a result, in this embodiment, the output can be automatically reduced to an output capable of cooling the fuel assembly 1 in the core 20 with the capacity of the coolant that can be supplied from the emergency high-pressure core water injection system. Sufficient safety potential can be maintained even during complex events exceeding. Such a present Example can improve the safety margin even when the above-mentioned complex event occurs without impairing the economics of the light water breeding reactor, which is a light water reactor.

  In this example, the height of the lower blanket area is higher than the height of the upper blanket area, and the height of the upper blanket area is 70 mm, which is 105 mm or less. The safety margin of the core can be further improved. In this embodiment, the height of the lower blanket region is higher than the height of the upper blanket region, and the height of the upper fuel region is 16 mm higher than that of the lower fuel region. It is possible to further improve the safety margin of the core when this occurs.

  The core in the case where the upper end of the neutron absorber filling region 15 of the safety rod, which is the control rod 2 that is fully pulled out during the rated output operation of the BWR 19, is positioned at the lower end of the effective fuel length 14 of the fuel rod 3. As a measure for suppressing the decrease in reactivity, the height of the upper fuel region 6 is set to 241 mm, and the height of the lower fuel region 8 is set to 225 mm. In order to keep the growth ratio 1.01 while minimizing the influence on the void coefficient, the height of the upper blanket 5 is 70 mm, and the height of the lower blanket 9 is 1.6 times that of the upper blanket 5. The thickness was 280 mm.

  According to this embodiment, when all the constraints are satisfied and the growth ratio is maintained at 1.01, a composite event occurs that exceeds the design standard in which all control rods do not operate when the core flow rate is significantly reduced for some reason. However, the output can be automatically reduced to an output capable of cooling the fuel assembly by the capacity of the coolant that can be supplied to the core from the emergency high-pressure core water injection system. For this reason, the safety margin of BWR19 which is a light water breeding reactor can be improved (refer FIG. 4).

In the BWR 19 to which the core 20 is applied, which generates the same electric power 1350 MW as the ABWR, using the reactor pressure vessel 27 having the same size as that of the current ABWR, the upper fuel region excluding the upper blanket region 5A and the lower blanket region 9A 6A, the burnup degree of the core region including the lower fuel region 8A and the internal blanket region 7A is higher than that of the light water breeding reactor described in Japanese Patent No. 3428150, in which the burnup is 45 GWd / t can do. The take-off burnup of the core region of the core 20 is 53 GWd / t, and the take-off burnup of the core 20 including the upper blanket region 5A and the lower blanket region 9A is 45 GWd / t. According to this example, MCPR is 1.3, the void coefficient is −3 × 10 ⁻⁴ Δk / k /% void, and R.I. TAKEDA et al., Proc. More than the void coefficient of light water breeder reactors described in GLOBAL '07 Boise, USA, September, 2007, P.1725 -2 × 10 ⁻⁵ Δk / k /% void of International Conference on Advanced Nuclear Fuel Cycles and Systems. The absolute value increases by an order of magnitude. Furthermore, according to the present embodiment, a growth ratio of 1.01 can be realized with the ratio of each isotope of TRU kept substantially constant as described above.

  In the present embodiment, the upper end of the neutron absorber filling region 15 of the safety rod that is fully pulled out is positioned at the lower end of the effective fuel length 14 of the fuel rod 3 (lower end of the lower blanket region 9A) (see FIG. 14). Instead, below the effective fuel length 14 of each fuel rod 3 included in each fuel assembly, boron, gadolinia, Dy, Sm. Even if the pellet 21 containing a neutron absorbing substance such as Eu is arranged, the same effect can be obtained (see FIG. 19).

  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 a configuration in which the fuel assembly 1 is replaced with the fuel assembly 1K shown in FIGS. 20 and 22 in the first embodiment, and other configurations are the same as those in the first embodiment. . The configuration of the present embodiment will be described with respect to differences from the first embodiment, and the description of the same configuration as the first embodiment will be omitted. The core 20A is also a parfait type core. The light water reactor to which the core 20A is applied is the BWR 19 shown in FIG. 11 in which the core 20 is replaced with the core 20A. The BWR 19 to which the core 20A is applied has the same configuration as the BWR 19 to which the core of the first embodiment is applied except for the core 20. The core 20A is a core applied to a TRU extinguishing reactor.

  In a fuel assembly 1K (see FIG. 20) arranged in the core 20A, 397 fuel rods 3K having a diameter of 7.2 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3K is 2.2 mm, and 11 fuel rods 3K are arranged in the outermost fuel rod row. As shown in FIG. 21 in the state of the equilibrium core, the reactor core 20A is provided with fuel assemblies 1A to 1D having different operating cycle numbers. The fuel assembly 1D in the fourth cycle is disposed in the outermost peripheral region of the core. The fuel assemblies 1A for the first cycle are arranged in the outer region of the core, and the fuel assemblies 1B, 1C, 1D of the second to fourth cycles are dispersedly arranged in the inner region of the core. Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1B are annularly arranged. In such a core 20A, the power distribution in the radial direction is further flattened. Each of the fuel assemblies 1A to 1E shown in FIG. 21 is a fuel assembly 1K.

  The fuel assembly 1K has a configuration in which the fuel effective length portion is divided into four regions, and the lower blanket is removed from the fuel assembly 1 (see FIG. 22). The height of the upper blanket region 5 is 30 mm, the height of the upper fuel region 6 is 228 mm, the height of the inner blanket region 7 is 560 mm, and the height of the lower fuel region 8 is 215 mm. When the fuel assembly 1K is a new fuel assembly with zero burnup, all the fuel rods 3K of the fuel assembly 1K have two blanket regions filled with depleted uranium oxide pellets, the upper fuel region 6 and the lower fuel region 1K. The fuel region 8 is filled with TRU oxide fuel. This TRU oxide fuel has a fissile Pu enrichment of 13.0 wt%. The TRU used for the fuel assembly 1K is obtained by reprocessing the nuclear fuel material of the spent fuel assembly. Each blanket region is not filled with the mixed oxide fuel. Each TRU fuel present in the upper fuel region 6 and the lower fuel region 8 includes TRUs having the compositions shown in Table 2. The fuel assembly 1L has a burnup of zero, and the ratio of Pu-239 in all TRUs is 8.5 wt%. The core 20A is formed by 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 8A. The lower fuel regions 8A are sequentially arranged from the upper end toward the lower end. In the core 20A, the lower end of the lower fuel region 8A coincides with the position of the lower end of the core 20A, and the lower blanket region is not formed.

Similarly to FIG. 14 of the first embodiment, in this embodiment, the safety rod, which is the control rod 2 that is fully pulled out when the BWR 19 is operated at the rated output, has neutron absorption filled with B ₄ C. The material filling region 15 is pulled out from the core 20A so that the upper end of the material filling region 15 is positioned at the lower end of the effective fuel length of the fuel rod 3K. The control rod 2 is provided with a follower portion 16 made of carbon, which is a substance having a lower deceleration ability than light water, above the neutron absorber filling region 15.

According to the present embodiment in which the height of the upper blanket region is 30 mm which is 100 mm or less and the lower end of the lower fuel region coincides with the lower end of the core 20A and no lower blanket region is provided, the entire blanket state is obtained. The upper end of the neutron absorbing material filling region 15 of the plurality of safety rods is located at the lower end of the effective fuel length 14 of the fuel rod 3, that is, the lower end of the lower fuel region 8A (see FIG. 14), thereby being a TRU extinguishing furnace. When the core flow rate suddenly drops for some reason during the operation of the BWR19 and a composite event occurs that exceeds the design standard in which all control rods do not operate, the void ratio in the core 20A rises rapidly and is slightly subcooled. Thus, the boiling start point of the coolant flowing from below the core 20A shifts to the lower end side of the core 20A, and the power distribution in the core axis direction also shifts to the lower end side of the core. That. At this time, excess neutrons shifted to the lower end side of the core can be absorbed by B ₄ C existing in each neutron absorber filling region 15 whose upper end is located at the lower end of the lower fuel region 8A. As a result, the output can be automatically reduced to an output at which the fuel assembly can be cooled by the capacity of the coolant that can be supplied from the emergency high-pressure core water injection system to the core 20A. For this reason, even when a composite event exceeding the design standard occurs, a sufficient safety potential can be maintained in the TRU extinguishing furnace. Such a present Example can improve a safety margin even when said composite event arises, without impairing the economical efficiency of the TRU extinction reactor which is a light water reactor.

  This embodiment has an upper blanket region, and the height of the upper fuel region is 13 mm higher than that of the lower fuel region, which is 13 mm higher. Therefore, the safety margin of the core is further improved when the above composite event occurs. Can be made.

  Core reactivity when the upper end of the neutron absorber filling region 15 of the safety rod that is fully pulled out during operation of the BWR 19 is positioned at the lower end of the effective fuel length of the fuel rod 3 (lower end of the lower fuel region 8A) While taking measures against the decrease, the height of the upper blanket region 5 was set to 30 mm and the height of the upper fuel region 6 was set 13 mm higher than the height of the lower fuel region 8 so that the influence on the void coefficient was minimized.

  According to this embodiment in which the height of the upper blanket region is 30 mm which is 100 mm or less, and the lower end of the lower fuel region coincides with the lower end of the core 20A and no lower blanket region is provided, all constraints are satisfied. However, the capacity of the coolant that can be supplied to the core 20A from the emergency high-pressure core water injection system even when a composite event occurs that exceeds the design standard in which all control rods do not operate when the core flow rate drops significantly for some reason Thus, the output can be automatically reduced to an output that can cool the fuel assembly (see FIG. 10). For this reason, even if such a composite event occurs, the safety margin of the core 20A can be improved. The core 20A can reduce the amount of TRU contained in the fuel assembly 1K than when the burnup is zero.

In the BWR 19 to which the core 20A is applied that generates the same electrical output 1350 MW as the ABWR by using the reactor pressure vessel having the same size as that of the current ABWR, it is possible to achieve the take-off burnup 65 GWd / t of the core 20A. According to this embodiment, MCPR is 1.3, the void coefficient is −4 × 10 ⁻⁴ Δk / k /% void, and R.I. TAKEDA et al., Proc. More than the void coefficient of the TRU extinguishing furnace described in GLOBAL '07 Boise, USA, September, 2007, P.1725 -2 × 10 ⁻⁵ Δk / k /% void of International Conference on Advanced Nuclear Fuel Cycles and Systems. The absolute value increases by an order of magnitude. Furthermore, according to the present embodiment, it is possible to reduce the TRU while realizing the TRU isotope ratio retention.

  Also in the present embodiment, as in the first embodiment, the upper end of the neutron absorber filling region 15 of the safety rod that is fully drawn is the lower end of the effective fuel length 14 of the fuel rod 3 (the lower end of the lower fuel region 8A). (See FIG. 14), instead of the effective fuel length 14 of each fuel rod 3 included in each fuel assembly, boron, gadolinia, Dy, Sm. Even if the pellet 21 containing a neutron absorbing substance such as Eu is arranged, the same effect can be obtained (see FIG. 19).

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

  The core 20B of the light water reactor of the present embodiment has a configuration in which the fuel assembly 1K is replaced with the fuel assembly 1L described in FIGS. 24 and 25 in the core 20A of the second embodiment. Is the same. The light water reactor to which the core 20B is applied is a BWR 19 shown in FIG. 11 in which the core 20 is replaced with the core 20B. The BWR 19 to which the core 20B is applied has the same configuration as the BWR 19 to which the core of the first embodiment is applied except for the core 20. The core 20B is a core applied to a TRU extinguishing reactor. The configuration of the present embodiment will be described with respect to parts different from the second embodiment, and the description of the same configuration as the second embodiment will be omitted.

  In the fuel assembly 1L (see FIG. 24) used in this embodiment, 397 fuel rods 3L having a diameter of 7.6 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3L is 1.8 mm, and 11 fuel rods 3L are arranged in the outermost fuel rod row. As shown in FIG. 23 in the state of the equilibrium core, the reactor core 20B is provided with fuel assemblies 1A to 1D having different numbers of operating cycles. The fuel assembly 1D in the fourth cycle is disposed in the outermost peripheral region of the core. The fuel assemblies 1A for the first cycle are arranged in the outer region of the core, and the fuel assemblies 1B, 1C, 1D of the second to fourth cycles are dispersedly arranged in the inner region of the core. Between the core inner region and the core outer region, there is an intermediate region in which a plurality of fuel assemblies 1B are annularly arranged. In the core 20B, the power distribution in the radial direction is further flattened. Each of the fuel assemblies 1A to 1E shown in FIG. 23 is a fuel assembly 1L.

  Like the fuel assembly 1K, the fuel assembly 1L divides the effective fuel length into four regions (see FIG. 25), the upper blanket region 5 has a height of 50 mm, and the upper fuel region 6 Is 183 mm, the inner blanket region 7 is 560 mm, and the lower fuel region 8 is 173 mm. When the fuel assembly 1L is a new fuel assembly having zero burnup, all the fuel rods 3L of the fuel assembly 1L are filled with oxide pellets of depleted uranium in two blanket regions, and an upper fuel region 6 and a lower fuel assembly 1L. The fuel region 8 is filled with TRU oxide fuel. The enrichment level of fissile Pu in this TRU fuel is 18.0 wt%. Each blanket region is not filled with the mixed oxide fuel. Each TRU oxide fuel present in the upper fuel region 6 and the lower fuel region 8 contains TRUs having the compositions shown in Table 3. This TRU is a material obtained by reprocessing the nuclear fuel material of the spent fuel assembly. The fuel assembly 1L has a burnup of zero, and the ratio of Pu-239 in all TRUs is 13.4 wt%. The core 20B is also formed by 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 8A. The lower fuel regions 8A are sequentially arranged from the upper end toward the lower end. In the core 20B, the lower end of the lower fuel region 8A coincides with the position of the lower end of the core 20B, and the lower blanket region is not formed.

  Similarly to FIG. 14 of the first embodiment, this embodiment also has a neutron absorber filling region 15 of each safety rod (a part of the control rod 2) that is fully pulled out when the BWR 19 is operated at the rated output. Is located at the lower end of the effective fuel length of the fuel rod 3L (lower end of the lower fuel region 8A). The Y-shaped control rod 2 is provided with a follower portion 16 made of carbon, which is a substance having a lower deceleration ability than light water, above the neutron absorber filling region 15.

According to this embodiment in which the height of the upper blanket region is 50 mm, which is 100 mm or less, and the lower end of the lower fuel region coincides with the lower end of the core 20B and no lower blanket region is provided, the entire blanket region is brought out. By placing the upper end of the neutron absorbing material filling region 15 of each safety rod at the lower end of the effective fuel length 14 of the fuel rod 3, that is, the lower end of the lower fuel region 8A (see FIG. 14), the BWR 19 which is a TRU extinguishing furnace When a composite event that exceeds the design standard where all the control rods do not operate occurs due to some reason during operation, the void ratio in the core 20B suddenly increases and is slightly subcooled. The boiling start point of the coolant flowing in from below the core 20B moves to the lower end side of the core 20B, and the power distribution in the core axis direction shifts to the lower end side of the core 20B. To. At this time, excess neutrons shifted to the lower end side of the core 20B can be absorbed by B ₄ C existing in each neutron absorber filling region 15 whose upper end is located at the lower end of the lower fuel region 8A. As a result, the output can be automatically reduced to an output at which the fuel assembly can be cooled by the capacity of the coolant that can be supplied from the emergency high-pressure core water injection system to the core 20B. For this reason, even when a composite event exceeding the design standard occurs, a sufficient safety potential can be maintained in the TRU extinguishing furnace. Such a present Example can improve a safety margin, even when said composite event arises, without impairing the economical efficiency of a TRU extinction furnace.

  Since the present embodiment has an upper blanket region and the height of the upper fuel region is 10 mm higher than that of the lower fuel region, it is possible to further improve the safety margin of the core when the above composite event occurs. .

  Core reactivity when the upper end of the neutron absorber filling region 15 of the safety rod that is fully pulled out during operation of the BWR 19 is positioned at the lower end of the effective fuel length of the fuel rod 3 (lower end of the lower fuel region 8A) While taking measures against the decrease, the height of the upper blanket region 5 was set to 50 mm and the height of the upper fuel region 6 was set to be 10 mm higher than the height of the lower fuel region 8 so that the influence on the void coefficient was minimized.

  According to the present embodiment, even when a composite event exceeding the design standard occurs in which all control rods do not operate when the core flow rate is significantly reduced for some reason while satisfying all the constraints, the emergency high pressure core injection system The output can be automatically reduced to an output capable of cooling the fuel assembly with the capacity of the coolant that can be supplied to the core 20B. For this reason, even if such a composite event occurs, the safety margin of the core 20B can be improved. The core 20B can reduce the amount of TRU contained in the fuel assembly 1L than when the burnup is zero.

In the BWR 19 to which the core 20B is applied that generates the same electrical output 1350 MW as the ABWR using a reactor pressure vessel having the same size as the current ABWR, it is possible to achieve a take-off burnup 65 GWd / t of the core 20C. In this embodiment, the void coefficient is −6 × 10 ⁻⁴ ΔΔk / k /% void, the MCPR is 1.3, the ratio retention of TRU isotopes can be realized, and TRU can be reduced.

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

  The core 20C of the light water reactor according to the present embodiment has a configuration in which the fuel assembly 1K is replaced with the fuel assembly 1M shown in FIGS. 26 and 27 in the core 20A of the second embodiment. The same. The light water reactor to which the core 20C is applied is the BWR 19 shown in FIG. 11 in which the core 20 is replaced with the core 20C. The BWR 19 to which the core 20C is applied has the same configuration as the BWR 19 to which the core of the first embodiment is applied except for the core 20. The core 20C is a core applied to a TRU extinguishing reactor. The configuration of the present embodiment will be described with respect to parts different from the second embodiment, and the description of the same configuration as the second embodiment will be omitted.

  In the fuel assembly 1M (see FIG. 26) used in this embodiment, 397 fuel rods 3M having a diameter of 7.1 mm are arranged in a regular triangular lattice in the channel box 4. The gap between the fuel rods 3M is 2.3 mm, and 11 fuel rods 3M are arranged in the outermost fuel rod row. The arrangement of the fuel assemblies in the balanced core of the present embodiment is the same as that of FIG.

  Like the fuel assembly 1K, the fuel assembly 1M divides the effective fuel length into four regions (see FIG. 27), the upper blanket region 5 has a height of 30 mm, and the upper fuel region 6 The height of the inner blanket region 7 is 560 mm, and the height of the lower fuel region 8 is 227 mm. When the fuel assembly 1M is a new fuel assembly having zero burnup, all the fuel rods 3M of the fuel assembly 1M are filled with two uranium oxide pellets in the two blanket regions, and the upper fuel region 6 and the lower fuel region 1M. The fuel region 8 is filled with TRU oxide fuel. This TRU oxide fuel has a fissile Pu enrichment of 11.6 wt%. Each blanket region is not filled with the mixed oxide fuel. Each TRU fuel present in the upper fuel region 6 and the lower fuel region 8 includes TRUs having the compositions shown in Table 4. This TRU is a material obtained by reprocessing the nuclear fuel material of the spent fuel assembly. The fuel assembly 1M is in the state of a new fuel assembly, and the ratio of Pu-239 in all TRUs is 7.0 wt%. The core 20C is also formed by 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 8A. The lower fuel regions 8A are sequentially arranged from the upper end toward the lower end. In the core 20C, the lower end of the lower fuel region 8A coincides with the position of the lower end of the core 20C, and the lower blanket region is not formed.

  Similarly to FIG. 14 of the first embodiment, in this embodiment, the neutron absorption of each safety rod (a part of the Y-shaped control rod 2) which is fully pulled out when the BWR 19 is operated at the rated output. The upper end of the material filling area 15 is located at the lower end of the effective fuel length of the fuel rod 3M (lower end of the lower fuel area 8A). The control rod 2 is provided with a follower portion 16 made of carbon, which is a substance having a lower deceleration ability than light water, above the neutron absorber filling region 15.

According to the present embodiment in which the height of the upper blanket region is 30 mm which is 100 mm or less and the lower end of the lower fuel region coincides with the lower end of the core 20C and no lower blanket region is provided, the entire blanket state is obtained. By placing the upper end of the neutron absorbing material filling region 15 of each safety rod at the lower end of the effective fuel length 14 of the fuel rod 3, that is, the lower end of the lower fuel region 8A (see FIG. 14), the BWR 19 which is a TRU extinguishing furnace When a composite event that exceeds the design standard in which all the control rods do not operate occurs for some reason during operation, the void ratio in the core 20C increases rapidly and is slightly subcooled. The boiling start point of the coolant flowing from below the core 20C moves to the lower end side of the core 20C, and the power distribution in the core axis direction shifts to the lower end side of the core 20C. To. At this time, excess neutrons shifted to the lower end side of the core 20C can be absorbed by B ₄ C existing in each neutron absorber filling region 15 whose upper end is located at the lower end of the lower fuel region 8A. As a result, the output can be automatically reduced to an output that can cool the fuel assembly by the capacity of the coolant that can be supplied from the emergency high-pressure core water injection system to the core 20C. For this reason, even when a composite event exceeding the design standard occurs, a sufficient safety potential can be maintained in the TRU extinguishing furnace. Such a present Example can improve a safety margin, even when said composite event arises, without impairing the economical efficiency of a TRU extinction furnace.

  This embodiment has an upper blanket region, and the height of the upper fuel region is 13 mm higher than that of the lower fuel region, which is 13 mm higher. Therefore, the safety margin of the core is further improved when the above composite event occurs. Can be made.

  Core reactivity when the upper end of the neutron absorber filling region 15 of the safety rod that is fully pulled out during operation of the BWR 19 is positioned at the lower end of the effective fuel length of the fuel rod 3 (lower end of the lower fuel region 8A) While taking measures against the decrease, the height of the upper blanket region 5 was set to 30 mm and the height of the upper fuel region 6 was set 13 mm higher than the height of the lower fuel region 8 so that the influence on the void coefficient was minimized.

  According to the present embodiment, even when a composite event exceeding the design standard occurs in which all control rods do not operate when the core flow rate is significantly reduced for some reason while satisfying all the constraints, the emergency high pressure core injection system The output can be automatically reduced to an output that can cool the fuel assembly with the capacity of the coolant that can be supplied to the core 20C. For this reason, even if such a composite event occurs, the safety margin of the core 20C can be improved. The core 20C can reduce the amount of TRU contained in the fuel assembly 1M compared to when the burnup is zero.

A take-off burnup of 65 GWd / t can be achieved in the BWR 19 to which the reactor core 20C is applied, which generates the same electric power 1350 MW as the ABWR, using a reactor pressure vessel having the same size as the current ABWR. In this embodiment, the void coefficient is −3 × 10 ⁻⁴ Δk / k /% void, the MCPR is 1.3, the ratio retention of TRU isotopes can be realized, and TRU can be reduced.

  The core of the light water reactor according to embodiment 5, which is another embodiment of the present invention, will be described in detail below with reference to FIG. The core of the light water reactor of the present embodiment has the configuration shown in FIG. 28 for the fuel assembly 1 loaded in the core 20 of the first embodiment, and other configurations are the same as those of the first embodiment.

  As shown in FIG. 28, the fuel assembly is arranged in the fuel effective length portion from the upper end toward the lower end, and as shown in FIG. 28, the upper blanket region 5, the upper fuel region 6, the inner blanket region 7, the lower fuel region 8 and the lower blanket region. Nine five regions are sequentially formed. The height of each region is as follows. The height of the upper blanket region 5 is 105 mm, the height of the upper fuel region 6 is 248 mm, the height of the inner blanket region 7 is 520 mm, the height of the lower fuel region 8 is 232 mm, and the lower blanket The height of the region 9 is 280 mm. The core 20 loaded with the plurality of fuel assemblies 1 in which the respective regions shown in FIG. 28 are loaded also includes the upper blanket region 5A formed by the upper blanket region 5, the upper fuel region 6A formed by the upper fuel region 6, An inner blanket region 7A formed by the inner blanket region 7, a lower fuel region 8A formed by the lower fuel region 8A, and a lower blanket region 9A formed by the lower blanket region 9 are sequentially arranged from the upper end to the lower end. Yes.

  The core of the present embodiment can also obtain each effect produced in the first embodiment.

In the present embodiment, which considers only safety within the design standard, the reactor core of the present embodiment, which generates the same electrical output 1350 MW as that of ABWR, is applied using a reactor pressure vessel that is approximately the same size as the current ABWR. In the BWR 19, higher burnup can be realized than in the first embodiment, and a takeout burnup of 66 GWd / t in the core region and a takeout burnup of 55 GWd / t in the core including the upper and lower blanket regions can be realized. In this example, the void coefficient is −5 × 10 ⁻⁵ Δk / k /% void, the MCPR is 1.3, and the ratio of each isotope of TRU is kept substantially constant as described above. In this state, a growth ratio of 1.01 can be realized.

  The core of a light water reactor according to embodiment 6, which is another embodiment of the present invention, will be described in detail below with reference to FIGS. 29 and 30. FIG.

  In the core of the present embodiment, the fuel assembly 1 loaded in the core 20 of the first embodiment is configured as shown in FIG. 29, and other configurations are the same as those in the first embodiment. In this embodiment, when the light water reactor to which the core of this embodiment is applied is operated at the rated power, the safety rod (part of the Y-shaped control rod 2) filled in the neutron absorber is fully pulled out. The upper end of the region 15 is located at 1/5 of the height of the lower blanket region 9 from the lower end of the lower blanket region 9 (see FIG. 30).

  As shown in FIG. 29, this fuel assembly has an active fuel length portion, from the upper end toward the lower end, as shown in FIG. 29. The upper blanket region 5, the upper fuel region 6, the inner blanket region 7, the lower fuel region 8, and the lower blanket Five regions of region 9 are sequentially formed. The height of each region is as follows. The height of the upper blanket region 5 is 60 mm, the height of the upper fuel region 6 is 235 mm, the height of the inner blanket region 7 is 450 mm, the height of the lower fuel region 8 is 219 mm, and the lower blanket The height of the region 9 is 280 mm. Similarly to the core 20, the core of this embodiment also 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 formed in the fuel assembly shown in FIG. 29. 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 formed at the same axial position.

  The core of the present embodiment can also obtain each effect produced in the first embodiment.

In the BWR 19 to which the core of the present embodiment is applied that generates the same electric power 1350 MW as that of the ABWR using the reactor pressure vessel 27 having the same size as that of the current ABWR, the burn-up burnup 54 GWd / t in the core region is applied. A take-off burnup 45 GWd / t of the core 20 including the upper and lower blanket regions can be realized. Further, in this example, the void coefficient is −3 × 10 ⁻⁴ Δk / k /% void, the MCPR is 1.3, and the ratio of each isotope of TRU is substantially constant as described above. In this state, a growth ratio of 1.01 can be realized.

  1, 1K, 1L, 1M ... Fuel assembly, 2 ... Control rod, 3 ... Fuel rod, 5, 5A ... Upper blanket region, 6, 6A ... Upper fuel region, 7, 7A ... Internal blanket region, 8, 8A ... Lower fuel region, 9, 9A ... lower blanket region, 14 ... active fuel length, 15 ... neutron absorber filling region, 19 ... BWR, 20, 20A, 20B, 20C ... core.

Claims (4)

A nuclear fuel material including a plurality of isotopes of a transuranium nuclide, and is disposed between an upper blanket region, a lower blanket region, and the upper blanket region and the lower blanket within an effective fuel length filled with the nuclear fuel material. In the fuel assembly forming the fuel region containing the above-mentioned super uranium nuclide,
When the burnup is zero, the ratio of Pu-239 in all the above-mentioned transuranium nuclides contained in the nuclear fuel material is in the range of 40% or more and 60% or less, and is located at the lowest position within the effective fuel length. The sum of the height of the lower blanket region and the height of the upper blanket region located at the uppermost position within the effective fuel length is in the range of 250 mm to 600 mm, and the height of the lower blanket region is the upper blanket A fuel assembly characterized by being in the range of 1.6 times to 12 times the height of the region.   The fuel assembly according to claim 1, wherein a height of the upper blanket region is in a range of 30 mm to 105 mm. The fuel assembly includes an inner blanket region, an upper fuel region that is disposed between the upper blanket region and the inner blanket, and includes the super uranium nuclide, and a region between the upper blanket region and the inner blanket region and the lower blanket. A lower fuel region that is the fuel region and includes the transuranium nuclide,
The fuel assembly according to claim 1 or 2, wherein a height of the upper fuel region is higher than a height of the lower fuel region within a range of 10 mm to 25 mm.   The fuel assembly according to any one of claims 1 to 3, wherein a nuclear fuel material including a neutron absorber is disposed below the lower blanket region.

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