Light water reactor fuel assembly and method of nuclear fuel cycle utilization

Abstract

Provided is a light water reactor uranium fuel assembly capable of reducing the heat generation amount of both Am-241 and Cm-244, which are causes of glassification, by a light water reactor alone without using a fast neutron reactor, and capable of reducing the amount of glassification. A light water reactor uranium fuel assembly for a nuclear fuel cycle in which an americium isotope is extracted at the time of reprocessing of used fuel and the extracted americium isotope is added to the nuclear fuel cycle, the weight ratio W (wt% unit) of the added americium 241 with respect to the weight of the fuel heavy metal being W < -0.006e with respect to the average uranium 235 enrichment e (wt% unit) of the fuel assembly 2 +0.12e-0.43 (enrichment 5 wt% or more), W < -0.000356e+0.00357 (enrichment 4.2 wt% or more and less than 5.0 wt%).

Description

TECHNICAL FIELD

[0001] Embodiments of the present application relate to a light water reactor uranium fuel assembly and a method for use in a nuclear fuel cycle. BACKGROUND

[0002] A fuel assembly used in a light water reactor for power generation (hereinafter, a light water reactor) is configured by arranging a plurality of fuel rods containing a nuclear material in a lattice shape and bundling them, and is configured so that, at the time of output operation of the nuclear reactor, the fuel rods generate heat by nuclear fission reaction of the nuclear material inside, and the generated heat can be removed by cooling water guided to the inside of the fuel assembly.

[0003] The light water reactor has a boiling water type light water reactor and a pressurized water type light water reactor, the boiling water type light water reactor taking out heat inside the fuel assembly as steam and generating power, and the pressurized water type light water reactor taking out heat inside the fuel assembly as high-temperature water and generating power after becoming steam by being introduced into a heat exchanger such as a steam generator. The former is called a BWR (Boiled Water Reactor), and the latter is called a PWR (Pressurized Water Reactor).

[0004] Among the fuel assemblies used in the light water reactor (both BWR and PWR), there are a uranium fuel assembly containing only uranium as a nuclear material that generates a nuclear fission reaction, and a MOX (Mixed Oxide Fuel) fuel assembly containing plutonium (hereinafter, Pu) and uranium as nuclear materials.

[0005] The uranium fuel assembly contains uranium 234, uranium 235, uranium 238, which are isotopes of the element uranium, and in some of the fuel assemblies, uranium 236 (hereinafter, U-234, U-235, U-238, U-236, respectively) is also contained. Furthermore, among the uranium fuel assemblies, there are two types of uranium fuel rods, one containing a burnable poison and the other not containing a burnable poison.

[0006] A burnable poison is a substance added in the fuel assembly in order to adjust the nuclear fission reaction of the nuclear material, and for example, in the fuel assembly used in a boiling water type nuclear reactor, gadolinium oxide (Gd2O3) as an oxide of gadolinium is sometimes used, but in addition to gadolinium oxide, substances having the property that the neutron absorption ability decays due to burning are also known.

[0007] A fuel assembly after emitting a prescribed energy in a light water reactor is called spent fuel (hereinafter, SF). SF contains Pu, and on the other hand, in addition to fission products (hereinafter, FP) and residual U-235, U-238 and the like, there are nuclides having an atomic number greater than Pu or a group of nuclides collectively called minor actinides (hereinafter, MA) including neptunium 237 (hereinafter, Np-237). In addition, as MA, in addition to the above, there are nuclides which are nucleophilic to Pu and uranium isotopes.

[0008] As to Pu, since it is a useful nuclear fission material, in Japan, France, a nuclear fuel cycle is performed in which Pu taken out by performing a process of taking out Pu from SF (hereinafter, reprocessing) and residual uranium are reused as mixed oxide fuel (hereinafter, MOX fuel).

[0009] FP and MA are unstable nuclides which undergo β-decay or α-decay and the like, and SF which heats due to these decays needs to be stored and cooled for a certain period to be able to perform reprocessing. After cooling, reprocessing is performed to take out Pu and uranium. MA and FP are contained in the residue as waste.

[0010] FP and MA contained in the waste solution after reprocessing also heat, and are liquid which has a high level of radioactivity and heats, and therefore are called high level radioactive waste (hereinafter, HLW). HLW is mixed into glass and solidified (hereinafter, vitrification) after reprocessing in order to finally stably perform a process of being buried into a stratum. Specific nuclides among MA and FP remaining in the residue can be chemically separated. For example, separation of americium, curium and other MA nuclides in reprocessing can be performed.

[0011] If the decay heat of SF increases due to the influence of burnup and the like, the cooling period until reprocessing increases, and the period of being stored in a storage facility is prolonged. Since a certain amount of SF is discharged from a light water reactor under the conditions of the light water reactor and the like, the cooling period becomes long, and when the number of bodies supplied from the storage facility to reprocessing decreases, the increase rate of the number of stored SF bodies increases, and the storage surplus decreases.

[0012] Similarly, if the decay heat of HLW per unit volume increases, there is an upper limit to the heat generation per glass weight of a vitrification body, and therefore the volume of the vitrification body increases to increase the amount of waste. As described above, the decay heat has an influence on the storage surplus of SF and the burial capacity surplus of the vitrification body.

[0013] As for the latter, in view of the excess capacity of the glassified body in the disposal, a technique of extracting specific MA nuclides by nuclide in the reprocessing process and adding a part of the nuclides to a light water reactor or a fast neutron reactor to transmute into single-life nuclides is known as partitioning and transmutation technique (hereinafter referred to as P&T). It is known that MA nuclides as the object of P&T are americium 241 and curium 244 (hereinafter referred to as Am-241 and Cm-244, respectively) which are large in heat generation contribution in HLW.

[0014] As a heat generation component in MA contained in SF, plutonium 238 also occupies a large proportion, but since it is removed as Pu in the reprocessing, it is contained only in a trace amount in HLW, and the contribution to heat generation is to the extent that can be ignored.

[0015] As a result, in a light water reactor, the component of MA decay heat in SF which is taken out from the reactor and has passed 10 to 100 years is roughly Am-241 and Cm-244. As for the time change thereof, it is known that as for Am-241, the decay heat increases with time, and Cm-244 shows a decreasing characteristic, and in the example of uranium fuel of the existing light water reactor, the decay heat becomes the same degree in about 20 years of cooling.

[0016] Am-241 is generated due to beta decay of plutonium 241 (hereinafter, Pu-241) with a half-life of 14.4 years to Am-241 with the increase of the time elapsed. That is, the longer the cooling period until the reprocessing, the more the decay heat component of Am-241 increases.

[0017] As one of the P&T techniques, a method of adding Am-241 to the fuel of a light water reactor to perform decay destruction is known. An example of addition to VVER fuel as a form of PWR is also known. In this example, the thermal neutron region of Am-241 has a large capture cross section, and therefore, as the burnup of the fuel assembly progresses, transmutation progresses, and Am-241 decreases.

[0018] On the other hand, Am-241 after transmutation further repeatedly performs transmutation based on thermal neutron absorption and decay to reach Cm-244. That is, in the transmutation of Am-241 in a light water reactor, there is a relationship in which the decay heat of Am-241 decreases, and on the other hand, the decay heat of Cm-244 increases.

[0019] There is also an example of adding MA to the fuel of a BWR to perform decay. In this example, all of the four nuclides of Np-237, Am-241, Am-243, and Cm-244 in MA are simultaneously added to UO2 in a certain ratio only in the fuel for MA burning, and are added homogeneously, and Am nuclides are not added alone. Furthermore, as an example of this evaluation, only the case where the uranium weight ratio of MA is 5 wt% is shown. In addition, in this example, the enrichment of U-235 of the uranium fuel rod is a standard value, and the enrichment is not increased by the addition of MA.

[0020] On the other hand, as an example of reducing MA generated by a light water reactor by a method other than P&T, there is an example of increasing the enrichment of U-235 to a value required to achieve a burnup degree, and reducing the amount of TRU generated. For example, in a case where the burnup degree is set to 45 GWd / t, and the average uranium 235 enrichment needs to be 3.8 wt%, by setting it to 10 wt%, an example in which the total weight ratio of MA in SF is reduced compared to fuel in which the reference enrichment is not increased is also known.

[0021] In this example, the amount of Pu-241 generated and the amount of Cm-244 generated are each greatly reduced compared to fuel in which the reference enrichment is not increased.

[0022] Before the used fuel is reprocessed, it needs to be cooled to the level of decay heat accepted by the reprocessing system, and under the latest reference of the reprocessing plant in Japan, uranium fuel needs to be cooled for 15 years or more to reach shearing.

[0023] As such a document, there is Japanese Patent Publication No. 2017-32408 (hereinafter referred to as Patent Document 1) of Japan. SUMMARY

[0024] High-level radioactive waste, after being vitrified, is semi-permanently stored by being buried in the ground for ground disposal, but a ground disposal site is required in proportion to the amount of vitrified body. A site suitable for a ground disposal site needs to be a site where ground changes are less long-term, and the available sites are limited. If the vitrified body also increases in the future along with the amount of SF generated, the disposal site will be full, and disposal will not be possible. As a result, reprocessing will not be possible, and power generation itself can not be possible. As such, the current nuclear fuel cycle in which high-level radioactive waste is vitrified and ground disposal is performed is considered to have no long-term sustainability, and becomes a large technical problem in terms of the continuation of nuclear power generation.

[0025] In view of such a technical problem, it is considered to reduce high-level radioactive waste by P&T. However, the P&T technology is premised on a transmutation reactor other than a light water reactor such as a fast neutron reactor, and thus there is a technical problem that a very long period is required until practical use and glassified bodies are continuously produced until practical use. Therefore, in order to solve this problem, it is necessary to reduce the generation of high-level waste and glassified bodies in a light water reactor without being premised on the practical use of a fast neutron reactor.

[0026] Among the range known in the past, there is a method of reducing Am-241 by recycling Am-241, which is a part of high-level radioactive waste, in a light water reactor. However, in the method of recycling Am-241 in a light water reactor, there is a problem that Cm-244 increases, and the total amount of decay heat of MA cannot be reduced in addition to originally generated Cm-244. In addition, there is a uranium fuel in which the enrichment of U-235 in a light water reactor fuel is increased to a necessary burnup or more, and the generation of MA including Cm-244 is reduced, but there is a technical problem that Am-241 cannot be reduced.

[0027] The present application was made in order to solve the above-described technical problem, and has an object to provide a light water reactor uranium fuel assembly and a method for using a nuclear fuel cycle, which can reduce the heat generation amount of both Am-241 and Cm-244, which are causes of the generation of glassified bodies, by only a light water reactor without using a fast neutron reactor, and can reduce the amount of glassified bodies.

[0028] The light water reactor uranium fuel assembly of the embodiment is characterized in that the weight ratio W (wt% unit) of added Am-241 to the weight of fuel heavy metals is in the range of W < -0.006e + 0.12e - 0.43 (enrichment of 5 wt% or more), W < -0.000356e + 0.00357 (enrichment of 4.2 wt% or more and less than 5.0 wt%) with respect to the average U-235 enrichment e (wt% unit) of the fuel assembly. 2 +0.12e - 0.43 (enrichment of 5 wt% or more), W < -0.000356e + 0.00357 (enrichment of 4.2 wt% or more and less than 5.0 wt%) with respect to the average U-235 enrichment e (wt% unit) of the fuel assembly.

[0029] Effects of Invention

[0030] According to the present application, it is possible to provide a light water reactor uranium fuel assembly and a method for using a nuclear fuel cycle, which can reduce the heat generation amount of both Am-241 and Cm-244, which are causes of the generation of glassified bodies, by only a light water reactor without using a fast neutron reactor, and can reduce the amount of glassified bodies. BRIEF DESCRIPTION OF DRAWINGS

[0031] Figure 1 is a graph showing the configuration position of the uranium fuel rod of the fuel assembly of the embodiment, the burnable poison rod.

[0032] Figure 2 is a graph showing the enrichment of uranium 235, the concentration of the burnable poison, the concentration of Am-241 of the embodiment.

[0033] Figure 3 is a graph showing the progress of the decay heat of the embodiment.

[0034] Figure 4 is a graph showing the progress of the decay heat of Comparative Example 1.

[0035] Figure 5 is a graph showing the progress of the decay heat of Comparative Example 2.

[0036] Figure 6 is a graph showing the progress of the decay heat of Comparative Example 3.

[0037] Figure 7 is a graph showing the progress of the decay heat in the case where reprocessing is performed 40 years after the withdrawal.

[0038] Figure 8 is a graph showing the progress of the change in the weight ratio of Am-241 of the embodiment.

[0039] Figure 9 is a graph showing the progress of the change in the weight ratio of Am-241 of the embodiment after the withdrawal.

[0040] Figure 10 is a graph showing the comparison of the decay heat in the case where reprocessing is performed 40 years after the withdrawal.

[0041] Figure 11 is a graph showing the comparison of the decay heat in the case where reprocessing is performed 20 years after the withdrawal.

[0042] Figure 12 is a graph showing the comparison of the decay heat in the case where reprocessing is performed 20 years after the withdrawal.

[0043] Figure 13 is a graph showing a mathematical expression for limiting the range of the concentration of added Am-241.

[0044] Figure 14 is a graph showing the difference in the degree of burn of the weight ratio in the burn of Pu-241.

[0045] Figure 15 is a graph showing the progress of the decay heat at the enrichment of 10 wt% of uranium 235 at each concentration of Am-241.

[0046] Figure 16 is a graph showing the passage of decay heat at a uranium 235 enrichment of 7.5 wt% at each Am-241 concentration.

[0047] Figure 17 is a graph showing the passage of decay heat at a uranium 235 enrichment of 5.0 wt% at each Am-241 concentration.

[0048] Figure 18 is a graph showing the passage of decay heat at a uranium 235 enrichment of 3.8 wt% at each Am-241 concentration.

[0049] Figure 19 is a graph showing the upper and lower limits of Am-241 addition.

[0050] Figure 20 is a graph showing a structural example of a horizontal cross section of a fuel assembly.

[0051] Figure 21 is a graph showing a structural example of a fuel rod.

[0052] Figure 22 is a graph showing a structural example of a fuel assembly and a control rod in a nuclear reactor.

[0053] BRIEF DESCRIPTION OF DRAWINGS

[0054] 1... fuel assembly, 2... fuel cladding tube, 3... fuel rod, 4... fuel rod containing a burnable poison, 5... water rod, 6... channel box, 7... fuel assembly, 8... core, 9... partial long fuel rod. DETAILED DESCRIPTION

[0055] Hereinafter, a light water reactor uranium fuel assembly and a method for using a nuclear fuel cycle according to an embodiment will be described with reference to the accompanying drawings.

[0056] In this embodiment, by increasing the U-235 enrichment to a necessary level of combustion, americium nuclides, including Am-241, generated from the reprocessing of uranium fuel (which reduces the formation of MA, including Cm-244) and uranium fuel SF, are added to the uranium fuel at a specific concentration range. The added Am-241 undergoes nuclide transformation during fuel combustion, resulting in a decrease in its concentration. Conversely, the Cm-244 concentration increases due to the nuclide transformation of Am-241. At this point, the Am-241 addition concentration condition can be determined such that it exceeds the range of Cm-244 reduction caused by using uranium fuel with increased U-235 enrichment, relative to the increase in Cm-244 decay heat caused by the addition of Am-241. In this way, the americium generated without increasing the heat of Cm-244 is used in the fuel addition, reducing the amount of Am-241 contained in the HLW used in the addition. As a result, the decay heat of Am-241 from HLW is reduced in the amount used in fuel additives, and the total decay heat of HLW, combined with the decay heat of Cm-244, is reduced, thereby enabling a reduction in the amount of glass-cured body produced.

[0057] First, examples of the structures of fuel assemblies, fuel rods, and reactor cores used in the past will be explained. For example... Figure 20 , Figure 21 , Figure 22 As shown, the reactor includes: fuel elements 1, which are sintered into cylindrical shapes using uranium dioxide, known as fuel particles; fuel rods 3, which are stacked in multiple layers and housed in fuel cladding tubes 2; fuel rods 4 containing combustible poison substances in fuel elements 1; waterlods 5, which contain no fuel elements and are internally supplied with cooling water during operation; a fuel assembly 7, which consists of fuel rods bundled in 9 rows and 9 columns and housed in channel boxes 6, which are shaped like quadrangular prisms; and a reactor core 8, which is formed by regularly arranging the fuel assembly 7. Additionally, the fuel assembly 7 contains multiple partially long fuel rods 9.

[0058] A portion of the fuel rods 3 contains gadolinium oxide, i.e., gadolinium oxide, as a burnable poison in the fuel element 1, and the average concentration thereof is about 4%. The fuel cladding tube 2 is made of a zirconium alloy called zircalloy or zircaloy. The light water reactor is a plant of a type called an advanced boiling water reactor (ABWR), and has a thermal output of 3926 MW at a rated operation, a number of fuel assemblies per core of 872, and a weight of uranium metal per 1 fuel of 172 kg. The thermal output of the core 8 is 100% of the rated, and the average burnup of the fuel element 1 during the operation of 1 cycle is about 45 GWd / t.

[0059] Next, the structure of the embodiment will be shown in Figure 1 、 Figure 2 . The fuel assembly 7 is configured with 9 rows and 9 columns of fuel rods 4, which are composed of uranium fuel rods U and burnable poison-containing fuel rods G containing gadolinium oxide as a burnable poison material. The number of each fuel rod is 32 of the uranium fuel rods U and 42 of the burnable poison-containing fuel rods G, for a total of 74. The average enrichment of uranium 235 of the uranium fuel rods U and the burnable poison-containing fuel rods G is 10 wt%. The burnable poison concentration is 8 wt% as shown in Figure 2 . Am-241 is added at a weight ratio of 0.35 wt% in the uranium fuel rods U, and is not added in the burnable poison-containing fuel rods G. The average weight ratio of Am-241 of all the fuel rods is 0.15 wt%. In addition, in the burnable poison-containing fuel rods G, the amount of addition of the burnable poison must be strictly managed, and therefore Am-241 is not added in the present embodiment. However, it is also possible to add Am-241 to the burnable poison-containing fuel rods G as well.

[0060] Next, the effects and principles of the embodiment will be shown by comparing with the prior art. Figure 3 The time variation of the decay heat of Am-241, Cm-244, and the total thereof will be shown in the case where the uranium fuel of the embodiment is burned at a burnup of 45 GWd / t, removed from the core, and then cooled.

[0061] The results of the comparative example in the case where the same burnup condition is burned and cooled in the prior art will be shown in Figure 4 、 Figure 5 、 Figure 6 .

[0062] Figure 4is the result of the first comparative example in the case where the uranium enrichment and the burnup are the same, and only Am-241 is not added to the fuel assembly 7. Compared with the first comparative example, the Am-241 and Cm-244 and the total decay heat increase in the period of approximately 70 years after the initial cooling. This is because, as a result of the addition of Am-241 in the example, the Am-241 burnup and the Cm-244 generated by the nuclear transmutation of the added Am-241 increase compared with the first comparative example. The example is the result of the increase in the decay heat, but as described later, since the Am-241 component can be removed from the decay heat after the reprocessing to form only Cm-244, the decay heat after the reprocessing of the example is greatly reduced.

[0063] Figure 5 is a graph showing the result of the second comparative example in which the average uranium 235 enrichment is set to 3.8wt%, and the burnup is the same as the example, 45 GWd / t. In addition, Am-241 is not added. When the second comparative example is compared with the example, the decay heat of Am-241 and Cm-244 and the total decay heat are approximately the same. This is the result of the mutual cancellation of the decrease in the amount of decay of Am-241 due to the decrease in the generation of Pu-241 caused by the increase in the enrichment without increasing the burnup in the example, and the increase in the amount of nuclear transmutation of Cm-244 by the added Am-241 in the new fuel. As a result, the example is the result of the decay heat being equivalent to that of the second comparative example, but the decay heat after the reprocessing can be greatly reduced in the example due to the above reason.

[0064] Figure 6 is a graph showing the result of the third comparative example in which the average uranium 235 enrichment is set to 3.8wt%, and the burnup is the same as the example, 45 GWd / t. In addition, Am-241 is added at the same concentration of 0.15wt% as the example. If the third comparative example is compared with the example, the decay heat of Cm-244 is greatly reduced, and the decay heat of Am-241 is somewhat reduced, and the total decay heat is greatly reduced particularly in the initial cooling period. This is the result of the effect of the reduction in the decay heat of Am-241 and Cm-244 caused by the increase in the enrichment without increasing the burnup in the example and the addition of Am-241 to the new fuel approximately canceling each other, and, in contrast, in the third comparative example, both the decay heat of Cm-244 and Am-241 become large by the Am-241 added to the new fuel.

[0065] In Figure 7 is a graph showing the total decay heat of Am-241 and Cm-244 when the SF of the example is removed from the core and reprocessed at the time of 40 years after the reprocessing, and the HLW glass containing Am-241 and Cm-244 is solidified. Figure 7the graph. The solid line 1 in the graph is the second comparative example, and the broken line 2 is the first comparative example. In the present embodiment, Am-241 is added to the uranium fuel in the entire amount after reprocessing, so that Am-241 is not left on the HLW side, and thus the entire Am-241 decay heat is excluded from the total of the decay heat of the HLW after reprocessing.

[0066] The reason for this is as follows. Figure 8 indicates the change in the burnup of the Am-241 weight ratio of the present embodiment, Figure 9 indicates the change in the Am-241 weight ratio when the used fuel is taken out at a burnup of 45 GWd / t and cooled. Initially, Am-241 is 0.15 wt%, but by burning it becomes about 0.03 wt% at a burnup of 45 GWd / t, and thereafter increases by applying the decay component of Pu-241 by cooling. Thereafter, Am-241 decreases after the maximum value of 0.11 wt%. That is, even if the addition of Am-241 at 0.15 wt% is performed, at the time of taking out it is lower than 0.15 wt%, so if the Am-241 generated by burning is used for the entire amount of the addition of Am-241 to the fuel, Am-241 can not be left on the HLW side. That is, if Am-241 is repeatedly added to the uranium fuel in an appropriate concentration range, Am-241 generated in the burning of the fuel can be enclosed in the core of the light water reactor using uranium fuel at all times. In contrast, the comparative examples are all not a method of adding Am-241 to the uranium fuel, the decay heat of Am-241 is included in the decay heat of the HLW, and accordingly the decay heat of the HLW increases, and the amount of the glass solidification body increases. In addition, if the addition of Am-241 at 0.15 wt% is repeatedly performed, the amount of Am-241 remaining in the SF gradually decreases, but if Am-241 from, for example, 4 to 5 SFs is used collectively, the addition of Am-241 at 0.15 wt% can be performed at all times, and the application of the present application is always possible.

[0067] In Figure 10 the graph. The solid line 1 in the graph is the second comparative example, and the broken line 2 is the first comparative example. In the present embodiment, Am-241 is added to the uranium fuel in the entire amount after reprocessing, so that Am-241 is not left on the HLW side, and thus the entire Am-241 decay heat is excluded from the total of the decay heat of the HLW after reprocessing. Figure 10The middle FORSETI (registered trademark) indicates a light water reactor uranium fuel assembly using a surplus concentration of concentrated uranium. In this embodiment (3 in the figure), since Am-241 can be excluded after reprocessing, the residual decay heat has only a component from Cm-244, and can be greatly reduced to 11% compared to the second comparative example. The glassified body limits the HLW heat generation amount therein below a certain temperature, and thus the amount of the glassified body is proportionally reduced by the reduction of the decay heat of the HLW. In addition, in this example, a case where FP is not included in the glassified body is shown, but it is clear that even if FP is present, the glassified body can be reduced by this embodiment.

[0068] In Figure 11 , the cooling time variation of the total value of the decay heat of Am-241 and Cm-244 at the time of reprocessing after 20 years from the time when SF is taken out of the core of this embodiment (3 in the figure) is shown. The point line 3 in Figure 7 is shown together with the first comparative example (dotted line 2 in the figure), the second comparative example (solid line 1 in the figure). The same effect as Figure 7 can be obtained by this embodiment, but since the reprocessing period is early, Cm-244 decays less after SF is taken out, and thus the decay heat amount is increased compared to the example of Figure 7 . Figure 12 The value of the decay heat is compared in , but is greatly reduced to 23% compared to the second comparative example. Even if the cooling period until reprocessing is shortened like this, according to this embodiment, the glassified body can be greatly reduced.

[0069] Figure 13 Next, the range of the concentration of Am-241 to be added, which is also included in this embodiment, is shown in . In the case where the concentration of Am-241 is W (wt%), and the enrichment of uranium 235 is e (wt%), the following inequalities, formulae <1> W <-0.006e 2 +0.12e-0.43 (enrichment 5 wt% or more), formula <2> W <-0.000356e+0.00357 (enrichment 4.2 wt% or more and less than 5.0 wt%) are applied to determine the upper limit. The upper limit of the addition concentration W of Am-241 to the uranium fuel is limited to the range in the case where the following advantages are obtained.

[0070] The upper limit is a value set in such a way that the decay heat of Am-241 and Cm-241 in SF does not exceed the decay heat of the fuel of the reference enrichment 3.8 wt%. In addition, the decay heat in SF also includes a component from FP, but since the FP decay heat has almost no influence on the composition of the fuel and the degree of burning, the above inflection is established as it is even if Am-241 is added.

[0071] Further, a certain condition is considered to be set for the range of the burnup to which the present embodiment can be applied. The use Figure 14 This is explained. Figure 14 The Pu-241 weight fraction in the burnup at the burnup range of 30 to 60 GWd / t is shown in the case where the uranium 235 enrichment is changed to 3.8, 5.0, 7.5, and 10.0 wt% with the added weight fraction of Am-241 being 0.15 wt%. The Pu-241 decays with a half-life of 14.4 years and is transmuted to Am-241 during the cooling process after the SF is taken out. If this case is considered where the reprocessing is performed after the SF is taken out and more than 15 years have passed, and further, if the case is considered where the weight fraction of Am-241 immediately after the SF is taken out is one tenth of the amount of Pu-241, the amount of Pu-241 at the time of the SF taken out becomes the upper limit of the Am-241 in the SF at the time of the reprocessing. Therefore, the weight fraction of Am-241 in the SF at the time of the reprocessing is independent of the cooling period, Figure 14 The Pu-241 weight fraction shown becomes the upper limit. The weight fraction of Pu-241 in the range of 60 GWd / t or less is 0.14 wt%, which is the upper limit, and therefore the weight fraction of Am-241 in the HLW is also not more than 0.14 wt% in this burnup range.

[0072] As in the present embodiment, in the case where Am-241 is added at 0.15 wt%, if it is in the above-mentioned burnup range, the Am-241 at the time of the addition can exceed the weight fraction of Am-241 in the HLW during any cooling period. This means that it is possible to achieve the complete confinement of Am-241 in the uranium fuel without transfer to the HLW side. However, if the condition of the complete confinement of Am-241 in the uranium fuel is not attached, the weight fraction of Am-241 added to the uranium fuel can be determined within the range of <Equation 1> and <Equation 2>.

[0073] The use Figure 15 , Figure 16 , Figure 17 , Figure 18 The basis for setting the upper limits of <Equation 1> and <Equation 2> is explained. First, the Figure 15The decay heat of the second comparative example (uranium 235 enrichment 3.8 wt%) was compared with the decay heat of the case where the added concentration (wt% unit) of Am-241 in the fuel assembly of the example in which the average uranium 235 enrichment was 10 wt% and the burnup was 45 GWd / t was changed to 0.0, 0.05, 0.10, 0.15, and 0.20, and the total value of the decay heat of the sum of Am-241 and Cm-244 was taken as a reference. In the case where the enrichment was 10 wt%, the total decay heat of Am-241 and Cm-244 increased approximately in proportion to the weight ratio of Am-241. The added concentration of Am-241 formed in such a manner that the decay heat was lower than that of the reference fuel was 0.17 wt% when the graph was read. That is, in the case where the average uranium 235 enrichment was 10 wt%, if the added concentration of Am-241 was not less than 0.17 wt%, the decay heat exceeded that of the second comparative example taken as a reference.

[0074] Figure 16 , Figure 17 , Figure 18 The same study as Figure 15 indicates that the uranium 235 enrichment (wt% unit) was changed to 7.5, 5.0, and 3.8, respectively. In the same way as Figure 14 , the upper limit concentration of Am-241 with respect to the enrichment was 0.135, 0.02, and 0.0. Of course, in the case where the enrichment was 3.8 wt%, Am-241 could not be added. Furthermore, although not shown, if linear interpolation is performed according to the size relationship of Figure 17 and Figure 18 , in the case where the enrichment was 4.2 wt%, the upper limit was 0.056 wt%.

[0075] Figure 19 The upper limit value of the added concentration of Am-241 obtained in Figure 15 , Figure 16 , Figure 17 , Figure 18 is summarized in a table, and <Equation 1> and <Equation 2> are equations in which the values are mathematized.

[0076] The above describes several embodiments of the present application, but these embodiments are presented as examples and are not intended to limit the scope of the application. These new embodiments can be implemented in other various ways, and various omissions, substitutions, and changes can be made within the scope of the gist of the application. These embodiments and modifications thereof are included in the scope or gist of the application, and are included in the scope of the application and equivalents thereof recited in the claims.

Claims

1. A light water reactor fuel assembly used for a nuclear fuel cycle, which is constructed by bundling a plurality of fuel rods, comprising a uranium fuel rod containing fissile uranium and americium extracted from reprocessing of used fuel and not containing a poison, and a poison rod containing uranium and a poison. The addition concentration W of americium 241 to the uranium fuel is: When the uranium enrichment e is 5wt% or more, less than -0.006e 2 +0.12e - 0.43, when the uranium enrichment is greater than or equal to 4.2wt% and less than 5.0wt%, less than -0.000356e+0.00357, where, wt% is a unit of the addition concentration W of americium 241 to the uranium fuel.

2. A method for operating a nuclear fuel cycle, which is a method for operating a nuclear fuel cycle in which a used fuel assembly is taken out and reprocessed, The method for operating the nuclear fuel cycle is characterized in that, americium is extracted from the used fuel assembly at the time of reprocessing and added to a light water reactor fuel assembly which is constructed by bundling a plurality of fuel rods, comprising a uranium fuel rod containing fissile uranium and americium extracted from reprocessing of used fuel and not containing a poison, and a poison rod containing uranium and a poison, The addition concentration W of americium 241 to the uranium fuel is: When the uranium enrichment is 5wt% or more, less than -0.006e 2 +0.12e-0.43, less than -0.000356e+0.00357 when the uranium enrichment is 4.2 wt% or more and less than 5.0 wt%, wherein wt% is a unit of the addition concentration W of americium 241 to the uranium fuel.

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