JP7251890B2 - MOX fuel assembly - Google Patents

Description

Embodiments of the present invention relate to MOX fuel assemblies.

Fuel assemblies used in power generation light water reactors consist of multiple fuel rods containing nuclear material arranged in a lattice and bundled together. , the generated heat can be removed by introducing water into the interior of the fuel assembly. Light water reactors for power generation include boiling water reactors, which extract the heat inside fuel assemblies as steam to generate electricity, and pressurized water reactors, which extract high-temperature water and lead it to a heat exchanger called a steam generator to convert it into steam for power generation. .

Fuel assemblies used in light water reactors for power generation (both boiling water reactors and pressurized water reactors) include uranium fuel assemblies containing only uranium as the nuclear fission reaction material, and plutonium (Pu) as the nuclear material. There are MOX fuel assemblies containing uranium. MOX fuel assembly is an abbreviation for mixed oxide fuel, and includes plutonium oxide (PuO ₂ ), which is an oxide of nuclear material contained in the spent nuclear fuel of a nuclear reactor, and uranium oxide ( UO ₂ ) is mixed and used.

This MOX fuel assembly includes MOX fuel rods that are fuel rods containing Pu, uranium fuel rods that do not contain Pu and contain burnable poisons shown below, and burnable poison-free fuel rods Three types of uranium fuel rods are included. Some MOX fuel assemblies do not include burnable poison-free uranium fuel rods.

Among the above MOX fuel assemblies, a fuel assembly using two types of MOX fuel rods and uranium fuel rods containing burnable poison is called a discrete MOX fuel assembly. A fuel assembly in which uranium fuel rods containing no burnable poison are arranged exclusively around the corners of the fuel assembly or adjacent to the channel box is called an island-type MOX fuel assembly.

A burnable poison is a substance added to a fuel assembly to regulate the nuclear fission reaction of nuclear material. ₂ O ₃ ) are often used. Substances like this gadolinia are called burnable poisons because they have the property of being reduced by nuclear reactions. Burnable poisons are usually added only to fuel rods that do not contain Pu.

The amount of nuclear material per MOX fuel assembly is set so that the fuel assembly has the thermal energy to be generated per body. For example, in the case of a MOX fuel assembly for a boiling water reactor, when the energy generated per ton of nuclear fuel (hereinafter referred to as burnup) is 45 GWd/t on average, the average Pu enrichment of the assembly is about 6 wt%. be. In the case of MOX fuel assemblies for pressurized water reactors, it is higher, about 11 wt%.

The aggregate average Pu enrichment is the ratio of Pu contained in the aggregate to the total heavy metals. This is the average value for all heavy metals contained in fuel rods and MOX fuel rods. In addition, the aggregate average uranium 235 enrichment is the enrichment of uranium 235 for all of the uranium fuel rods that do not contain burnable poison, the uranium fuel rod that contains burnable poison, and the MOX fuel rods. Average value for uranium.

Further, Pu has several kinds of isotopes, among which two kinds of nuclides, Pu-239 and Pu-241, which are well fissile under the conditions of a light water reactor, are called fissile Pu. Examples of the ratio of fissile Pu contained in spent fuel to total Pu are about 58% for BWR and about 65% for PWR.

The ratio of fissionable Pu changes under various conditions. For example, the ratio of fissile Pu increases when the fuel is extracted at a lower burnup than when the ejected burnup is 45 GWd/t.

A nuclear reaction occurs in a light water reactor, and a fuel assembly that has reached a predetermined burnup and is removed from the light water reactor is called a spent fuel assembly (hereinafter, spent fuel). Spent fuel is reprocessed to extract the remaining nuclear material that has not undergone a nuclear reaction. In order to perform reprocessing, spent fuel must be stored for a certain period of time and cooled to keep it below the upper limit of decay heat for reprocessing (hereinafter referred to as the upper limit of decay heat). For example, four years or more of cooling is required to shear spent fuel of uranium fuel. There are also cases where the required cooling period is set at 15 years or more by lowering the upper limit of decay heat due to the safety requirements of the reprocessing process.

Spent fuel contains unstable nuclei, and decay heat is the heat generated by the decay and decay of these nuclei. Decay heat is due to fission products and actinides. Both components decay over time in the decay heat of uranium fuel and MOX fuel, respectively. There is almost no difference in the components of fission products between MOX fuel assemblies and uranium fuel assemblies, and it has been found that the decay heat component of nuclides called actinides is one order of magnitude larger in MOX fuel than in uranium fuel. there is Due to this difference, decay heat becomes actinide-dominated in the MOX fuel assembly 10 years after the spent fuel is removed from the reactor. On the other hand, uranium fuel assemblies become actinide-dominated several decades after they are removed from the reactor. Therefore, the cooling period required for MOX fuel to reach a certain decay heat is longer than that for uranium fuel.

Among the materials extracted as nuclear materials by reprocessing the fuel assemblies, uranium is called recovered uranium and remains at a higher enrichment than the natural uranium-235 concentration of about 0.7 wt %. The enrichment of this recovered uranium (hereinafter referred to as the recovered uranium 235 enrichment) is about 1 wt % regardless of differences in light water reactor plants and burnup. Since the recovered uranium-235 enrichment is generally lower than the enrichment used in light water reactor nuclear fuel, the recovered uranium may be re-enriched and used.

There are also methods for increasing the recovered uranium-235 enrichment. It is known that the recovered uranium-235 enrichment can be increased in a wide range by using an initial uranium enrichment higher than the initial uranium enrichment required for the desired burnup (Patent Document 1).

There is a concept of limiting the ratio of Pu enrichment and uranium-235 enrichment in MOX fuel assemblies. When fissile Pu (the sum of Pu-239 and Pu-241 does not include Pu-240 and 242) and uranium 235 are included in the MOX fuel, the weight ratio of uranium 235 to the weight ratio of fissile Pu is increased (see Patent Document 2). However, Patent Document 2 focuses on the ratio of the weight of uranium 235 to the weight of fissile Pu, and does not focus on not only the fissile Pu but also the aggregate average Pu enrichment. Moreover, Patent Document 2 relates to a pressurized water reactor, not to a boiling water reactor.

In addition, in the technology for improving the core thermal margin, the average uranium enrichment U of the region filled with burnable poison and the average fissile Pu enrichment E of the MOX fuel rods are set to U/E≧1 .3 (see Patent Document 3). However, this Patent Document 3 also focuses on the ratio of the weight of uranium 235 to the weight of fissile Pu, and does not focus on not only the fissile Pu but also the aggregate average Pu enrichment. In addition, Patent Document 3 aims to improve the thermal margin of the core, and does not aim to shorten the cooling period required to remove decay heat and improve Pu time utilization efficiency.

Japanese Patent Application Laid-Open No. 2005-30772 WO2014/088461A1 JP-A-10-68789

When reprocessing the MOX fuel assemblies used in the above-mentioned light water reactors, the required cooling period until the start of reprocessing is longer than that of uranium fuel assemblies because the proportion of actinides is higher than that of uranium fuel assemblies. do. The longer the cooling period before reprocessing, the longer the period in which Pu obtained by reprocessing can be recovered. The efficiency of time utilization deteriorates.

For example, if the aggregate average Pu enrichment is the same and the cooling period is doubled from 100 years to 200 years, the Pu time utilization efficiency will be halved. Similarly, when the cooling period is the same and the collective average Pu enrichment is halved, the temporal utilization efficiency is also halved. If the temporal utilization efficiency of Pu decreases, the next-generation MOX fuel assembly will need to use uranium as the nuclear material for that amount.

In the case where the aggregate average Pu enrichment is several times the aggregate average uranium enrichment, the amount of Pu that can be added to one aggregate can be increased, but the decay heat increases, so the Pu temporal utilization efficiency cannot necessarily be large. In the case where the aggregate average Pu enrichment is a fraction of the aggregate average uranium-235 enrichment, the decay heat can be reduced, but the amount that can be added to one aggregate is reduced, so the time utilization of Pu is reduced. This means that efficiency cannot be increased. In other words, the ratio of the aggregate average Pu enrichment to the aggregate average uranium-235 enrichment was conventionally inefficient in terms of Pu temporal utilization efficiency.

The present invention has been made to solve the above-mentioned problems, and aims to provide a MOX fuel assembly capable of shortening the cooling period necessary for removing decay heat and improving Pu time utilization efficiency. do.

A MOX fuel assembly of an embodiment is a MOX fuel assembly that is configured by regularly arranging and bundling a plurality of fuel rods and is used in a boiling water reactor. MOX fuel rod containing fuel elements containing both uranium 235 and plutonium as fissile materials in a fuel cladding, and fuel elements containing burnable poison and containing uranium 235 as fissile materials and not containing plutonium when not burned and a burnable poison-containing uranium fuel rod stored in a fuel cladding, and the ratio obtained by dividing the aggregate average plutonium enrichment by the aggregate average uranium-235 enrichment is set to 0.40 or more and 1.08 or less , Uranium-235 enrichment in some or all of said MOX fuel rods is greater than 1.5 wt% .

According to the present invention, it is possible to provide a MOX fuel assembly capable of shortening the cooling period necessary for removing decay heat and improving Pu time utilization efficiency.

The figure which shows the setting of the MOX fuel assembly of 1st Embodiment. The graph which shows transition of the decay heat in 1st Embodiment. 5 is a graph showing the relationship between Pu enrichment uranium-235 enrichment ratio, required years for cooling, and Pu time utilization efficiency in the first embodiment. 5 is a graph showing the relationship between Pu enrichment uranium-235 enrichment ratio, required years for cooling, and Pu time utilization efficiency in the first embodiment. 5 is a graph showing the relationship between Pu enrichment uranium-235 enrichment ratio, required years for cooling, and Pu time utilization efficiency in the first embodiment. 5 is a graph showing the relationship between Pu enrichment uranium-235 enrichment ratio, required years for cooling, and Pu time utilization efficiency in the first embodiment. The figure which shows the Pu time utilization efficiency in 1st Embodiment in each cooling age of uranium fuel. FIG. 4 is a diagram showing the cooling years in the first embodiment for each cooling age of uranium fuel; The figure which shows the setting of the MOX fuel assembly of 2nd Embodiment. The graph which shows transition of the decay heat in 2nd Embodiment. The figure which shows the setting of the MOX fuel assembly of 3rd Embodiment. The graph which shows transition of the decay heat in 3rd Embodiment. The figure which shows the setting of the MOX fuel assembly of 4th Embodiment. The figure which shows the setting of the MOX fuel assembly of 5th Embodiment. The graph which shows transition of the decay heat in 5th Embodiment. The graph which shows the relationship between Pu enrichment uranium-235 enrichment ratio, the number of years required for cooling, and Pu time utilization efficiency in 5th Embodiment. The figure which shows the setting of the MOX fuel assembly of 6th Embodiment. The graph which shows transition of the decay heat in 6th Embodiment. Fig. 3 is a cross-sectional view schematically showing the configuration of a 9x9 fuel assembly; The figure which shows the structure of a fuel rod typically. FIG. 2 is a plan view schematically showing the configuration of fuel assemblies and a core; FIG. 2 is a schematic plan view showing the arrangement of fuel rods in a uranium fuel assembly; Fig. 2 shows a uranium fuel assembly configuration; The graph which shows transition of the decay heat in a uranium fuel assembly. FIG. 2 is a schematic plan view showing an example of the arrangement of fuel rods in an MOX fuel assembly; FIG. 2 shows a conventional MOX fuel assembly setup; The graph which shows transition of the decay heat of the conventional MOX fuel assembly. FIG. 2 shows a conventional MOX fuel assembly setup; The graph which shows transition of the decay heat of the conventional MOX fuel assembly. FIG. 4 is a schematic plan view showing another example of the arrangement of fuel rods in the MOX fuel assembly;

Hereinafter, MOX fuel assemblies according to embodiments will be described with reference to the drawings.

First, the configuration of the fuel assembly will be described. As shown in FIGS. 19 and 20, both the uranium fuel assembly and the MOX fuel assembly 7 consist of fuel elements 1 called fuel pellets, which are made by sintering uranium dioxide in a cylindrical shape, and stacking them in multiple stages. It has fuel rods 3 that are housed in fuel cladding tubes 2 . The fuel assembly 7 has a fuel element 1, a burnable poison containing fuel rod 4 containing a burnable poison substance, and a water rod 5 having no fuel element inside and through which cooling water flows during operation. The fuel assembly 7 is configured by bundling them in a channel box 6, which is a square prism-shaped tube, in 9 rows and 9 columns. Eight of the fuel rods are part length fuel rods 9 whose total length is reduced to about 2/3. As shown in FIG. 21, a core 8 is constructed by regularly arranging these fuel assemblies 7 .

A portion of the fuel rods are burnable poison rods containing gadolinium oxide or gadolinia as a burnable poison in the fuel element. The fuel cladding is made of a zirconium alloy called Zircaloy-2. The type of light water reactor is a plant called an advanced boiling water reactor (ABWR), and the thermal output during rated operation is 3926 MW, the number of fuel assemblies per core is 872, and the amount of nuclear material per fuel is 3,926 MW. The metal weight is approximately 172 kg. Here, nuclear materials are uranium dioxide and Pu dioxide.

The arrangement of fuel assemblies 7 in the core 8 of an ABWR is as shown in FIG. In FIG. 21, one square corresponds to one fuel assembly. FIG. 22 shows the fuel rod arrangement of the uranium fuel assembly. A uranium fuel assembly includes uranium fuel rods (symbol U) and burnable poison rods (symbol G).

FIG. 23 shows the uranium-235 enrichment and average values for each type of fuel rod in the uranium fuel assembly. The average uranium-235 enrichment of a uranium fuel assembly is about 3.8 wt%. The enrichment of each uranium fuel rod and burnable poison rod is not necessarily the same, but the average value of the fuel assembly is about 3.8 wt%. This fuel assembly has a discharge burnup of 45 GWd/t and is removed from the reactor and cooled.

Fig. 24 shows the change in decay heat after one year from the removal of the uranium fuel assembly from the reactor. After 1000 years have passed, the decay heat of actinides etc. has progressed and the decay heat has decreased to 1/100 or less of that immediately after extraction.

Next, FIG. 25 shows an example of the fuel rod arrangement of the MOX fuel assembly. The fuel rod average Pu enrichment and uranium-235 enrichment for MOX fuel rods and burnable poison rods includes MOX fuel rods (symbol M) and burnable poison rods (symbol G). This MOX fuel assembly is a so-called discrete type and does not contain uranium fuel rods.

FIG. 26 shows an example of Pu enrichment and the like for each type of fuel rod in this case. As in the case of the uranium fuel assembly, the nuclide concentrations of the MOX fuel rods and burnable poison rods are not necessarily the same, but their average values are the values for each fuel rod category shown in FIG. In this example, the fissionable Pu isotope ratio is 58%. This fuel assembly has a discharge burnup of 45 GWd/t and is removed from the reactor and cooled.

FIG. 27 shows the change in decay heat after one year from the removal of this MOX fuel assembly from the reactor. After 1000 years have passed, the decay heat of actinides and the like has progressed, and the decay heat is attenuated to a few tenths of that immediately after extraction. Attenuation of decay heat is slower than that of uranium fuel assemblies. This is because the amount of nuclides to be converted is incomparably large.

Furthermore, FIG. 28 shows the Pu enrichment and the like for each type of fuel rod in the case of a discrete MOX fuel assembly with a fissile Pu isotope ratio of 80%.

FIG. 29 shows the change in decay heat time after one year from the removal of this MOX fuel assembly from the reactor. Decay heat is attenuated more than in the fissile Pu isotope ratio of 58%. This is because the amount of Pu contained in the MOX fuel assembly in the initial unburned state is less when the fissile Pu isotope ratio is 80%, so the amount of actinides contained in the removed fuel assembly is also reduced. It's for.

Another type of fuel rod arrangement for MOX fuel assemblies is shown in FIG. This MOX fuel assembly is a so-called island type and includes uranium fuel rods (symbol U). MOX fuel assemblies of this type are also contemplated. In addition, since the time change of decay heat is almost dominated by the average Pu enrichment of the initial assembly, even if there is a difference between the discrete type and the island type, the average Pu enrichment and uranium 235 enrichment of the aggregate If they are the same, the time change of the decay heat will be almost the same.

(First embodiment)
Next, the MOX fuel assembly of the first embodiment will be explained with reference to FIGS. 1 to 8. FIG.
FIG. 1 shows the Pu enrichment, the uranium-235 enrichment, and the average value of each fuel rod of the MOX fuel assembly of the first embodiment. The discharge burnup of this MOX fuel assembly is 45 GWd/t, and the structure of the MOX fuel assembly is the same as that of the uranium fuel assembly. The arrangement of fuel rods for each fuel rod type is the same as that shown in FIG. In FIG. 25, symbol M is the MOX fuel rod and G is the combustible poison rod, which are of the so-called discrete type.

The average Pu enrichment of this MOX fuel assembly is 2.5 wt%, the uranium-235 enrichment is 2.8 wt%, and the fissile Pu isotope ratio is 58%. The ratio is 0.89. The enrichment of uranium-235 contained in the MOX fuel rods was set at 2.3 wt%, which means that uranium enriched with natural uranium or recovered uranium with an enrichment exceeding 1.5 wt% must contain at least 1.5 wt% of natural uranium. uranium-235 enriched in

FIG. 2 shows the change in decay heat after one year from the removal of this MOX fuel assembly from the reactor. As compared with the decay heat of the existing MOX fuel shown in FIG. 27, it can be seen that there is a significant attenuation effect after 10 years. The effect of decay heat attenuation will be described in detail below.

Pu enrichment uranium 235 ratio (horizontal axis) obtained by dividing the aggregate average Pu enrichment by the aggregate average uranium 235 enrichment, Pu time utilization efficiency (solid line, left vertical axis), one year after removal from the reactor The graph in FIG. 3 shows the relationship between the number of years required for cooling (dashed line on the right vertical axis). The number of years required for cooling is the number of years in which decay heat becomes the same as when the uranium fuel assembly is cooled for 10 years. The required cooling period is greatly shortened to 16 years compared to 120 years for conventional MOX fuel, and the Pu time utilization efficiency is greatly increased to 2.85 times. Here, in the graph of FIG. 3, when the Pu enrichment uranium 235 ratio obtained by dividing the aggregate average Pu enrichment on the left end by the aggregate average uranium 235 enrichment is zero, the uranium fuel assembly, the right end Pu enrichment A conventional MOX fuel assembly shown in FIG.

Similarly, the graph in FIG. 4 shows the number of years required for cooling to reach the same decay heat as when the uranium fuel assembly is cooled for 15 years, and the graph in FIG. The graph in FIG. In either case, the number of years required for cooling is greatly shortened, and the Pu time utilization efficiency is greatly increased by 2.5 times or more.

The Pu time utilization efficiencies shown in the graphs of FIGS. 3 to 6 all have maximum values. This is because the degree uranium-235 ratio has a downwardly convex characteristic. Furthermore, the reason why the ratio of decay heat and Pu enrichment uranium 235 has a downward convex characteristic is that the neutron energy spectrum increases as the initial aggregate average Pu enrichment, which has a thermal neutron absorption cross section larger than uranium 235, is increased. By moving to the energy side and increasing the neutron flux in the resonance energy region, the amount of uranium-238 contained in all fuel rods during combustion that absorbs resonance energy neutrons increases, and the amount of nuclides converted to Pu-239 increases. This is because the greater the initial Pu enrichment, the more the minor actinide production increases beyond the initial value.

Fig. 7 summarizes the improvement effect of Pu time utilization efficiency, and Fig. 8 summarizes the required cooling years.
The above principle will be explained below. The components of decay heat are fission products and actinides as already explained. It is well known that the decay heat per unit weight of the fission product components is determined by the total amount of nuclear fission in the fuel assembly, and as a result, is almost exclusively governed by the burnup. That is, the amount of the decay heat fission product component per unit weight of the fuel assembly is almost unaffected by the history during combustion and the type of the fuel assembly.

On the other hand, the actinide decay heat per unit weight is affected by the initial Pu enrichment. Since the uranium fuel assembly does not contain Pu at the beginning, the actinide decay heat component is smaller than that of the MOX fuel assembly, and the decay heat of the whole is also smaller. For this reason, the decay heat of the MOX fuel assembly changes depending on the initial Pu enrichment.

On the other hand, in order to set the discharge burnup of the fuel assembly to a predetermined value, it is necessary to enrich the new fuel with required fissile nuclides. Conventional examples of this are shown in FIGS. 22 to 24 for uranium fuel assemblies and in FIGS. 25 to 29 for MOX fuel assemblies. In order to maximize the Pu time utilization efficiency in the MOX fuel assembly under the condition that the discharge burnup is set to a predetermined value in this way, the assembly average Pu enrichment under the required fissile nuclide concentration The ratio of the degree of uranium-235 to the aggregate-average uranium-235 enrichment should be set to an appropriate value.

In the first embodiment, an example of a discrete 9×9 array MOX fuel assembly for a boiling water reactor was shown. By setting the ratio of the assembly average Pu enrichment and the assembly average uranium 235 enrichment to a predetermined value, the island type MOX fuel assembly shown in FIG. It is clear that similar effects can be obtained with fuel assemblies, MOX fuel assemblies with a 10.times.10 array instead of a 9.times.9 array, and other types of MOX fuel assemblies.

Furthermore, although the uranium-235 in the MOX fuel rods is described as being mixed with recovered uranium, it is also possible to enrich natural uranium and use only uranium-235 without using recovered uranium. Also, an example in which the ratio of the Pu enrichment to the uranium-235 enrichment is set to 0.89 is shown, but this value can substantially maximize the Pu time utilization efficiency. For conventional MOX fuel assemblies, the Pu time utilization efficiency can be improved if the ratio of the assembly average Pu enrichment to the assembly average uranium 235 enrichment is less than 5.5 and greater than zero. Since such an aggregate average Pu enrichment and aggregate average uranium-235 enrichment are clearly effective, a combination may be used.

In addition, in the first embodiment, the aggregate average Pu enrichment can be reduced to 0.4 times that of the MOX fuel assembly of FIG. 26, which is a conventional example, so the absolute value of void reactivity is reduced. It also has the effect of improving safety performance in reactor operation, such as increasing the reactivity value of control rods.

As described above, according to the present embodiment, when the fissile Pu isotope ratio is 58%, the Pu enrichment uranium 235 ratio is set to a predetermined value, thereby significantly improving the Pu time utilization efficiency and cooling There is an advantage that the required number of years can be greatly shortened. In addition, by using uranium-235 with an enrichment degree of recovered uranium-235 exceeding 1.5%, recovered uranium can be used as it is without re-enrichment. 235 can be utilized, and there is an advantage that the production cost of fuel can be reduced compared to the case of using natural uranium enriched uranium 235.

(Second embodiment)
Next, a second embodiment will be described.
FIG. 9 shows the Pu enrichment, the uranium-235 enrichment, and the average value of each fuel rod of the MOX fuel assembly of the second embodiment. The discharge burnup of the MOX fuel assembly of this embodiment is 45 GWd/t, and the structure of the MOX fuel assembly is the same as that of the uranium fuel assembly. The aggregate average Pu enrichment is 1.3 wt%, the aggregate average uranium-235 enrichment is 3.2 wt%, the fissile Pu isotope ratio is 58%, the aggregate average Pu enrichment and the aggregate average uranium 235 enrichment The ratio divided by degrees is 0.40.

FIG. 10 shows the change in decay heat after one year from the removal of this MOX fuel assembly from the reactor. It takes 16 years for the decay heat of the uranium fuel assembly to reach 15 years after being removed from the reactor, and the cooling time is not much different from that of the uranium fuel assembly. The Pu time utilization efficiency is more than twice that of conventional MOX fuel assemblies.

In this second embodiment, by increasing the aggregate average uranium 235 enrichment and lowering the aggregate average Pu enrichment, the number of years required for cooling after removal from the reactor is further reduced from 23 years compared to the first embodiment. There is an advantage that it can be shortened.

(Third embodiment)
Next, a third embodiment will be described.
FIG. 11 shows the Pu enrichment, the uranium-235 enrichment, and the average value of each fuel rod of the MOX fuel assembly of the third embodiment. The discharge burnup of the MOX fuel assembly of this embodiment is 45 GWd/t, and the structure of the MOX fuel assembly is the same as that of the uranium fuel assembly. The aggregate average Pu enrichment is 2.8 wt%, the aggregate average uranium-235 enrichment is 2.6 wt%, the fissionable Pu isotope ratio is 58%, and the Pu enrichment uranium-235 enrichment ratio is 1.08. be.

FIG. 12 shows the change in decay heat after one year from the removal of this MOX fuel assembly from the reactor. In the uranium fuel assembly, decay heat corresponding to 15 years after removal from the reactor is reached in 28 years, and the Pu time utilization efficiency is 266%. In the third embodiment, there is an effect that the Pu time utilization efficiency can be improved more than in the second embodiment.

(Fourth embodiment)
Next, a fourth embodiment will be described.
The discharge burnup of the MOX fuel assembly of the fourth embodiment is 45 GWd/t, and the structure of the MOX fuel assembly is the same as that of the uranium fuel assembly. The region of the fuel rod type of the MOX fuel assembly of this embodiment is of a type called an island type, which is the same as that shown in FIG. 30 of the conventional example. FIG. 13 shows the Pu enrichment, the uranium-235 enrichment, and the average value of each fuel rod of the MOX fuel assembly of this embodiment. The aggregate average Pu enrichment is 2.5 wt%, the aggregate average uranium-235 enrichment is 2.8 wt%, the fissile Pu isotope ratio is 58%, and the Pu enrichment uranium-235 enrichment ratio is 0.89. be. It takes 23 years for the uranium fuel assembly to reach decay heat equivalent to 15 years after removal from the reactor, which is much shorter than the conventional MOX fuel assembly. Furthermore, Pu time utilization efficiency can be doubled or more than that of the conventional MOX fuel assembly.

(Fifth embodiment)
Next, a fifth embodiment will be described.
The discharge burnup of the MOX fuel assembly of the fifth embodiment is 45 GWd/t, and the structure of the MOX fuel assembly is the same as that of the uranium fuel assembly. The region of the fuel rod type of the MOX fuel assembly of this embodiment is the same as that shown in FIG. 25 showing the conventional example. FIG. 14 shows the Pu enrichment, the uranium-235 enrichment, and the average value of each fuel rod of the MOX fuel assembly of this embodiment. The aggregate average Pu enrichment is 1.4 wt%, the aggregate average uranium-235 enrichment is 3.1 wt%, the fissionable Pu isotope ratio is 80%, and the Pu enrichment uranium-235 enrichment ratio is 0.44. be.

Fig. 15 shows the change in decay heat after one year from the MOX fuel assembly of this embodiment being removed from the reactor. It takes 16 years for the uranium fuel assembly to reach the decay heat equivalent to 15 years after removal from the reactor, which is much shorter than the conventional MOX fuel assembly. Furthermore, the Pu time utilization efficiency is comparable to that of the conventional MOX fuel assembly.

Pu enrichment uranium 235 ratio (horizontal axis) obtained by dividing the aggregate average Pu enrichment by the aggregate average uranium 235 enrichment, Pu time utilization efficiency (solid line, left vertical axis), one year after removal from the reactor FIG. 16 shows the relationship between the number of years required for cooling (the dashed line on the right vertical axis). The Pu time utilization efficiency is maximized near 1.0 as in the case of the fissile Pu isotope ratio of 58%. This is due to the fact that the fissile Pu isotope ratio increases and the Pu enrichment decreases while the cooling time is also shortened.

(Sixth embodiment)
Next, a sixth embodiment will be described.
The discharge burnup of the MOX fuel assembly of the sixth embodiment is 45 GWd/t, and the structure of the MOX fuel assembly is the same as that of the uranium fuel assembly. The region of the fuel rod type of the MOX fuel assembly of this embodiment is the same as that shown in FIG. 25 showing the conventional example. FIG. 17 shows the fuel rod type regions of the MOX fuel assembly of this embodiment. The aggregate average Pu enrichment is 2.5 wt%, the aggregate average uranium-235 enrichment is 2.3 wt%, the fissionable Pu isotope ratio is 80%, and the Pu enrichment uranium-235 enrichment ratio is 1.05. be.

FIG. 18 shows the change in decay heat after one year from the removal of this MOX fuel assembly from the reactor. It takes 22 years for the uranium fuel assembly to reach decay heat equivalent to 15 years after removal from the reactor, and the Pu time utilization efficiency can be increased by about 40% compared to conventional MOX fuel assemblies.

As shown in the embodiment described above, it was shown that the fissile Pu isotope ratio is effective in the range of 58% to 80%. The isotope ratio is also within this range.

As described in each embodiment above, when the MOX fuel assembly used in the boiling water reactor is not burned, the ratio obtained by dividing the assembly average Pu enrichment by the assembly average uranium 235 enrichment is 0.40 By setting it to 1.08 or less, the cooling period required for decay heat can be shortened, and the Pu time utilization efficiency can be improved. In addition, using recovered uranium-235 that has not been re-enriched for MOX fuel rods with an enrichment of uranium-235 exceeding 1.5 wt % is also effective in reducing fuel assembly manufacturing costs.

Although several embodiments of the invention have been described above, these embodiments are presented by way of example and are not intended to limit the scope of the invention. These novel embodiments can be implemented in various other forms, and various omissions, replacements, and modifications can be made without departing from the scope of the invention. These embodiments and modifications thereof are included in the scope and gist of the invention, and are included in the scope of the invention described in the claims and equivalents thereof.

1 Fuel element 2 Fuel cladding tube 3 Fuel rod 4 Fuel rod containing burnable poison 5 Water rod 6 Channel box 7 Fuel assembly 9 …part-length fuel rods.

Claims (5)

A MOX fuel assembly configured by regularly arranging and bundling a plurality of fuel rods and used in a boiling water reactor,
The fuel rods are
MOX fuel rods containing fuel elements containing both uranium-235 and plutonium as fissile materials when unburned in fuel cladding;
a burnable poison containing uranium fuel rod containing a fuel element containing burnable poison when not burned, containing uranium 235 as a fissile material and not containing plutonium in a fuel cladding;
The ratio obtained by dividing the aggregate average plutonium enrichment by the aggregate average uranium-235 enrichment is set to 0.40 or more and 1.08 or less ,
A MOX fuel assembly , wherein the uranium-235 enrichment in some or all of said MOX fuel rods is greater than 1.5 wt% . The fuel rod further includes one or more uranium fuel rods containing fuel elements containing no burnable poison, uranium 235 as a fissile material, and no plutonium in a fuel cladding when not burned. The MOX fuel assembly according to claim 1, wherein 3. The MOX fuel assembly according to claim 1, wherein the discharge burnup is 45 GWd/t. 4. The MOX fuel assembly according to any one of claims 1 to 3, wherein recovered uranium that has not been re-enriched is used for the MOX fuel rods with a uranium-235 enrichment exceeding 1.5 wt%. 5. The MOX according to any one of claims 1 to 4 , wherein the total ratio of plutonium 239 and plutonium 241 among the plutonium isotopes contained in the MOX fuel rod is 0.58 or more and 0.80 or less. fuel assembly.

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