JP5931353B2 - Spent nuclear fuel reprocessing method and spent nuclear fuel reprocessing facility - Google Patents

Description

  The present invention relates to a spent nuclear fuel reprocessing method and spent nuclear fuel reprocessing facility for reprocessing spent nuclear fuel used in a nuclear reactor.

  Conventionally, spent nuclear fuel (hereinafter referred to as LWR spent nuclear fuel) used in a light water reactor (LWR) is reprocessed at a reprocessing facility for LWR. On the other hand, when reprocessing spent nuclear fuel (hereinafter referred to as FBR spent nuclear fuel) used in a fast breeder reactor (FBR), it is performed at a reprocessing facility for FBR.

  As a reprocessing method using such a reprocessing facility, an FBR reprocessing facility for reprocessing FBR spent nuclear fuel, a recovery residual liquid discharged from the FBR reprocessing facility, and an LWR spent nuclear fuel are used. A method using an LWR reprocessing facility for reprocessing is known (for example, see Patent Document 1). In this reprocessing method, the recovery residual liquid discharged from the FBR reprocessing facility can be reprocessed by the LWR reprocessing facility, so that the FBR reprocessing facility is economically light. Can do.

Japanese Patent Laid-Open No. 11-287890

  By the way, from the viewpoint of land acquisition and construction cost of a spent nuclear fuel reprocessing facility, it is desired to reprocess the LWR spent nuclear fuel and the FBR spent nuclear fuel in a single reprocessing facility. When the spent nuclear fuel is reprocessed in a single reprocessing facility, a plurality of different types of spent nuclear fuel are input into the reprocessing facility. At this time, the fissile rate of the fissile radioactive isotopes (plutonium 239, plutonium 241 and the like) of the nuclear fuel material (plutonium and the like) contained in the spent nuclear fuel is different depending on the type of the spent nuclear fuel. The fissile rate is the ratio of fissile radioactive isotopes in the nuclear fuel material. In particular, the ratio of plutonium 239 and plutonium 241 in the total plutonium is called the Pu fissile rate.

  A loading fuel manufactured using new nuclear fuel generated after reprocessing (for example, MOX fuel) needs to have the Pu fissile rate as the target Pu fissile rate in response to a request from the nuclear power generation facility. New nuclear fuel generated after reprocessing also needs to have the target Pu fissile rate. However, since spent nuclear fuel with a different Pu fissile rate is input to the reprocessing facility, depending on the type of spent nuclear fuel that is input to the reprocessing facility, the Pu fissile rate of the new nuclear fuel is set as the target Pu fissile rate. It may be difficult to do.

  Therefore, in the present invention, when a plurality of different spent nuclear fuels are reprocessed in a single reprocessing facility, the fissile ratio of the fissionable radioisotope of the nuclear fuel material contained in the new nuclear fuel after the reprocessing, It is an object of the present invention to provide a spent nuclear fuel reprocessing method and a spent nuclear fuel reprocessing facility capable of achieving a target fissile rate.

  The spent nuclear fuel reprocessing method of the present invention is a spent nuclear fuel reprocessing method in which a plurality of different spent nuclear fuels are reprocessed in a single reprocessing facility to generate new nuclear fuel, A mixing step of mixing a high fissil nuclear fuel with a high fissile rate and a low fissile nuclear fuel with a low fissile rate so that the fissile rate of the fissionable radioisotope contained in the nuclear fuel material is the target fissile rate. It is characterized by.

  According to this configuration, in the mixing step, the high fissil nuclear fuel and the low fissil nuclear fuel can be mixed. Therefore, the fission rate of the fissionable radioactive isotope of the radioactive material contained in the new nuclear fuel after the reprocessing is increased. It can be the target fissile rate. Thereby, the desired fuel for loading according to a request | requirement can be manufactured by using the new nuclear fuel after reprocessing. The nuclear fuel materials are uranium and plutonium, and the fissile radioisotopes are uranium 233, uranium 235, plutonium 239, plutonium 241 and americium 241 and the like.

  In this case, the fissile radioactive isotope preferably contains at least one of plutonium 239 and plutonium 241.

  According to this configuration, when a fuel for loading such as a MOX fuel used in a light water reactor or a MOX fuel used in a fast breeder reactor is manufactured using a new nuclear fuel, the plutonium mixed with the MOX fuel is used. The Pu fissile rate of plutonium 239 and plutonium 241 which are fissile radioisotopes can be set as the target Pu fissile rate. Thereby, the new MOX fuel obtained by reprocessing a spent nuclear fuel can be manufactured suitably.

  In this case, it is preferable that the enrichment degree of the nuclear fuel material contained in the loading fuel manufactured using the new nuclear fuel is equal to or less than the upper limit enrichment degree.

  According to this configuration, since the enrichment of the nuclear fuel material contained in the loading fuel manufactured using the newly generated nuclear fuel can be made lower than the upper limit enrichment, the reprocessing of the loading fuel can be performed. It can be easy.

  In this case, the nuclear fuel material is preferably plutonium, and the upper limit enrichment is preferably 30%.

  According to this configuration, the plutonium enrichment (hereinafter referred to as Pu enrichment) in the loading fuel can be set to 30% or less after setting the target fissile rate. Thereby, for example, when the loading fuel is dissolved using nitric acid, since the Pu enrichment is 30% or less, the total amount of plutonium contained in the loading fuel can be suitably dissolved. The Pu enrichment is the ratio of plutonium to all uranium and plutonium contained in the nuclear fuel.

  In this case, the spent nuclear fuel contained in the spent fuel assembly is exposed by mechanical treatment, and the spent nuclear fuel after mechanical treatment is dissolved using nitric acid, A dissolving step, a refining step for removing the insoluble residue in the fuel solution, a measuring step for measuring the fuel solution from which the insoluble residue has been removed, and a solution containing the nuclear fuel material from the measured fuel solution. A separation step for separation, a purification step for purifying the solution after separation, an enrichment adjustment step for adjusting the enrichment of the nuclear fuel material in the solution by concentrating the solution after purification, and after the adjustment of the enrichment A denitration step of denitrating the fuel solution to generate a new nuclear fuel containing nuclear fuel material, and a fuel production step of producing a loading fuel from the denitrated nuclear fuel, the mixing step is a mechanical treatment step And dissolution process, dissolution process and clarification Between the clarification step and the weighing step, between the weighing step and the separation step, between the separation step and the purification step, between the purification step and the enrichment adjustment step, and the enrichment adjustment step. It is preferable to be performed between the denitration process and between the denitration process and the fuel production process.

  According to this configuration, since the mixing step can be performed between any of the steps, the versatility of the reprocessing method can be improved.

  A spent nuclear fuel reprocessing facility according to the present invention is a spent nuclear fuel reprocessing facility that reprocesses a plurality of different spent nuclear fuels to generate new nuclear fuel, and is intended for fission of nuclear fuel material contained in the nuclear fuel. It is characterized by comprising a mixing device for mixing a high fissile nuclear fuel with a high fissile rate and a low fissile nuclear fuel with a low fissile rate so that the fissile rate of the active radioisotope becomes the target fissile rate.

  According to this configuration, since the mixing device can mix the high fissil nuclear fuel and the low fissil nuclear fuel, the fissile ratio of the fissionable radioisotope of the nuclear fuel material contained in the new nuclear fuel after reprocessing is obtained. It can be the target fissile rate. Thereby, the desired fuel for loading according to a request | requirement can be manufactured by using the new nuclear fuel after reprocessing.

  In this case, the mixing device may include one or more buffer units that respectively store a plurality of types of spent nuclear fuel having different fissile rates, and a mixing unit that mixes the spent nuclear fuel supplied from each buffer unit. preferable.

  According to this configuration, a plurality of types of spent nuclear fuel are respectively stored in one or more buffer units, and the amount of spent nuclear fuel supplied from each buffer unit to the mixing unit is adjusted, whereby the mixed nuclear fuel is mixed. The fissile rate of spent nuclear fuel can be accurately adjusted to the target fissile rate.

  In this case, it is preferable that the plurality of types of spent nuclear fuel stored in one or more buffer units are sequentially supplied to the mixing unit one by one.

  According to this configuration, since multiple types of spent nuclear fuel can be sequentially supplied to the mixing unit one by one, it is easy to adjust the fissile rate of the spent nuclear fuel after mixing to the target fissile rate. it can. In addition, it is preferable to supply in order from a high fissile nuclear fuel among several types of spent nuclear fuel.

  In this case, it is preferable that the mixing device has a buffer mixing unit that mixes and stores a plurality of types of spent nuclear fuel having different fissile ratios.

  According to this configuration, the high fissile nuclear fuel and the low fissile nuclear fuel can be mixed with a simple apparatus configuration without separately providing the buffer unit and the mixing unit.

  In this case, the spent nuclear fuel contained in the spent fuel assembly is exposed by mechanical treatment, and the spent nuclear fuel after mechanical treatment is dissolved using nitric acid, A dissolving device that removes the insoluble residue in the fuel solution, a weighing device that measures the fuel solution from which the insoluble residue has been removed, and a solution containing nuclear fuel material from the fuel solution after weighing. A separation device for separation, a purification device for purifying the solution after separation, an enrichment adjusting device for concentrating the purified solution to adjust the enrichment of nuclear fuel material in the solution, and after the enrichment adjustment A denitration device that denitrates the solution to generate a new nuclear fuel containing nuclear fuel material, and a fuel production device that produces fuel for loading from the nuclear fuel after denitration, the mixing device includes a mechanical processing device, a melting device Equipment, clarification equipment, weighing equipment, separation equipment , Purifier, enrichment adjusting device, denitrification device, among the fuel production apparatus, it is preferably provided on any device.

  According to this structure, since the mixing apparatus can be provided in any apparatus, the versatility of a reprocessing facility can be improved.

  According to the spent nuclear fuel reprocessing method and spent nuclear fuel reprocessing facility of the present invention, when a plurality of different spent nuclear fuels are reprocessed in a single reprocessing facility, a new nuclear fuel after the reprocessing is used. Since the fissile rate of the fissionable radioisotope of the nuclear fuel material contained in the target fuel can be set as the target fissile rate, a desired loading fuel can be manufactured according to the demand.

FIG. 1 is a schematic configuration diagram of a spent nuclear fuel reprocessing facility according to the first embodiment. FIG. 2 is a schematic configuration diagram around the weighing device in the spent nuclear fuel reprocessing facility according to the first embodiment. FIG. 3 is a flowchart regarding a method for reprocessing spent nuclear fuel. FIG. 4 is a schematic configuration diagram around the weighing device in the spent nuclear fuel reprocessing facility according to the first modification. FIG. 5 is a schematic configuration diagram around a separation device in a spent nuclear fuel reprocessing facility according to the second embodiment. FIG. 6 is a schematic configuration diagram around the enrichment adjusting device in the spent nuclear fuel reprocessing facility according to the third embodiment. FIG. 7 is a schematic configuration diagram around a fuel manufacturing apparatus in a spent nuclear fuel reprocessing facility according to a fourth embodiment.

  Hereinafter, a spent nuclear fuel reprocessing method and a spent nuclear fuel reprocessing facility according to the present invention will be described with reference to the accompanying drawings. The present invention is not limited to the following examples. In addition, constituent elements in the following embodiments include those that can be easily replaced by those skilled in the art or those that are substantially the same.

  The spent nuclear fuel reprocessing facility according to the present embodiment extracts a new nuclear fuel by extracting and reprocessing nuclear fuel materials such as uranium (U) and plutonium (Pu) contained in a plurality of different spent nuclear fuels. U products and U / Pu products are generated.

  FIG. 1 is a schematic configuration diagram of a spent nuclear fuel reprocessing facility according to the first embodiment. The reprocessing facility 1 is a single facility capable of reprocessing a plurality of different spent nuclear fuels. As different types of spent nuclear fuel, uranium fuel was used in boiling water reactor (BWR), BWR spent nuclear fuel, uranium fuel was used in pressurized water reactor (PWR) PWR spent nuclear fuel, B-MOX spent nuclear fuel using MOX fuel in boiling water reactor, P-MOX spent nuclear fuel using MOX fuel in pressurized water reactor, uranium fuel or MOX fuel in fast breeder reactor There are used FBR spent nuclear fuel.

  The multiple types of spent nuclear fuel differ in the fissile ratio of the fissile radioisotope of the nuclear fuel material contained in the spent nuclear fuel and the enrichment of the nuclear fuel material. Here, nuclear fuel materials include uranium and plutonium, and fissile radioisotopes include uranium 233, uranium 235, plutonium 239, plutonium 241 and americium 241. In the following description, the case where it is applied to plutonium as a nuclear fuel material will be described.

The spent nuclear fuel is different in the plutonium fissile ratio (Pu fissile ratio) in the spent nuclear fuel and the plutonium enrichment (Pu enrichment) in the nuclear fuel. Pu fissile rate is an index representing the ratio of fissile radioisotope to the total amount of plutonium. Plutonium fissionable isotopes are plutonium 239 ( ²³⁹ Pu) and plutonium 241 ( ²⁴¹ Pu). For this reason, the Pu fissile rate is obtained by “( ²³⁹ P + ²⁴¹ Pu) / Pu”. The Pu enrichment is an index representing the ratio of the content of plutonium to the total amount of uranium and plutonium. Therefore, the Pu enrichment is obtained by “Pu / (U + Pu)”. An index obtained by multiplying the Pu fissile rate and the Pu enrichment is the Pu fissile enrichment, and the Pu fissile enrichment is obtained by “( ²³⁹ P + ²⁴¹ Pu) / (U + Pu)”.

  Here, the loading fuel (for example, the MOX fuel) manufactured using the new nuclear fuel generated by the reprocessing is based on the request from the nuclear power generation facility side for the Pu fissile rate of the MOX fuel and the target Pu fissile rate. In addition, it is necessary to set the Pu enrichment of the MOX fuel to the target Pu enrichment. In other words, the Pu fissile enrichment of the MOX fuel needs to be the target Pu fissile enrichment.

  On the other hand, a plurality of types of spent nuclear fuels have different Pu fissile rates, and in particular, B-MOX spent nuclear fuel is the spent nuclear fuel with the lowest Pu fissile rate. Therefore, for example, if the MOX fuel manufactured using new nuclear fuel obtained by reprocessing the B-MOX spent nuclear fuel is the target Pu fissile rate, the Pu enrichment exceeds the upper limit enrichment, It is difficult to reprocess. For this reason, the reprocessing facility 1 of Example 1 has the following configuration. Hereinafter, with reference to FIG. 1, a reprocessing facility 1 for reprocessing a plurality of types of spent nuclear fuel described above will be described.

  As shown in FIG. 1, the reprocessing facility 1 includes a mechanical processing device (for example, a shearing device) 5, a dissolution device 6, a clarification device 7, a weighing device 8, a separation device 9, and a purification device 10. The enrichment adjusting device 11, the denitration device 12, the fuel production device 13, and the mixing device are provided.

  The components of the spent nuclear fuel supplied to the mechanical processing device 5 are estimated in advance. The estimated component of the spent nuclear fuel can be calculated based on data such as the component before use and the burnup during use.

  The mechanical processing device 5 exposes the spent nuclear fuel contained in the spent fuel assembly by mechanical processing. At this time, since waste gas is generated when the spent nuclear fuel is mechanically processed by the mechanical processing device 5, the mechanical processing device 5 sends out the generated waste gas to a waste gas processing device (not shown). The mechanical processing device 5 supplies the spent nuclear fuel after the mechanical processing toward the melting device 6.

  The melting device 6 dissolves the spent nuclear fuel by putting the spent nuclear fuel after mechanical treatment into the melting tank storing nitric acid. The spent nuclear fuel dissolved in nitric acid becomes a fuel solution, and the dissolving device 6 supplies the fuel solution to the refining device 7.

  The clarification device 7 is composed of, for example, a centrifuge. The clarification device 7 removes the insoluble residue contained in the fuel solution that does not dissolve in nitric acid, and supplies the fuel solution from which the insoluble residue has been removed to the metering device 8. The removed insoluble residue is supplied to a waste storage tank (not shown).

  The metering device 8 receives the fuel solution for measuring the components in the fuel solution. The components in the fuel solution include, for example, nitric acid concentration, uranium amount, isotope ratio in uranium, plutonium amount, isotope ratio in plutonium, and the like. Then, from the components in the fuel solution, indices such as U enrichment, U fissile ratio, U fissile enrichment, Pu enrichment, Pu fissile ratio, Pu fissile enrichment, and the like that are actually measured values are derived. The metering device 8 supplies the fuel solution after metering to the separation device 9.

  The separation device 9 separates the uranium solution and the uranium / plutonium solution as solutions from the fuel solution. Specifically, the separation device 9 inputs an extraction solvent into the fuel solution to generate a uranium / plutonium loaded solvent that is an extraction solvent containing uranium and plutonium. Subsequently, the separation device 9 inputs a Pu back-extraction solvent into the uranium / plutonium loading solvent to extract a uranium / plutonium solution (hereinafter referred to as a U / Pu solution) that is a solution containing uranium and plutonium. Thereafter, the separation device 9 inputs the U back-extraction solvent into the uranium-loaded solvent that is the uranium / plutonium-loaded solvent after plutonium extraction, and extracts a uranium solution (hereinafter referred to as a U solution) that contains uranium. .

  The purifier 10 purifies the U solution and the U / Pu solution into a uranium nitrate solution (hereinafter referred to as a nitric acid U solution) and a uranium nitrate / plutonium solution (hereinafter referred to as a nitric acid U / Pu solution). The purification device 10 supplies the purified nitric acid U solution and the nitric acid U / Pu solution to the enrichment adjusting device 11.

  The enrichment adjusting device 11 adjusts the Pu enrichment by mixing the purified nitric acid U solution and the nitric acid U / Pu solution after heating and concentrating. Specifically, the enrichment adjusting device 11 adjusts the Pu enrichment of the nitric acid U / Pu solution by mixing the nitric acid U solution with the concentrated nitric acid U / Pu solution. The enrichment adjusting device 11 supplies the nitric acid U solution and the enriched nitric acid U / Pu solution to the denitration device 12.

  The denitration device 12 denitrates the nitric acid U solution and the nitric acid U / Pu solution after the enrichment is adjusted to obtain uranium oxide and uranium / plutonium oxide. The obtained uranium oxide and uranium / plutonium oxide are sent to the fuel production apparatus 13.

  The fuel production apparatus 13 stores uranium oxide (U product) and uranium / plutonium oxide (U / Pu product), and appropriately mixes the stored uranium oxide and uranium / plutonium oxide to load fuel. Is manufacturing.

  The mixing device mixes a plurality of kinds of spent nuclear fuels having different Pu fissile rates so that the Pu fissile rate after mixing becomes the target Pu fissile rate. Although details will be described later, the mixing device is provided integrally with the weighing device 8. Spent nuclear fuel includes dissolved fuel solution, separated U solution and U / Pu solution, purified nitric acid U solution and nitric acid U / Pu solution, denitrated uranium oxide and uranium / plutonium oxide It means to include things.

  Next, the metering device 8 provided integrally with the mixing device will be described in detail with reference to FIG. FIG. 2 is a schematic configuration diagram around the weighing device in the spent nuclear fuel reprocessing facility according to the first embodiment. As shown in FIG. 2, the weighing device 8 includes a weighing tank 21, a plurality of buffer tanks (buffer units) 22, and an adjustment tank (mixing unit) 23.

  The measuring tank 21 is connected to the clarification device 7 and stores the fuel solution supplied from the clarification device 7 in order to measure the components of the fuel solution. The fuel solution after measurement is supplied toward the buffer tank 22 or the adjustment tank 23. The plurality of buffer tanks 22 are respectively connected to the measurement tank 21 and temporarily store the fuel solution after measurement. In the first embodiment, a plurality of buffer tanks 22 are provided, but one or more buffer tanks 22 may be provided. The fuel solution stored in the buffer tank 22 is supplied toward the adjustment tank 23. The plurality of buffer tanks 22 can store fuel solutions having different Pu fissile rates. The adjustment tank 23 is connected to the measurement tank 21 and the buffer tank 22 and stores the fuel solution supplied from the measurement tank 21 and the buffer tank 22. The adjustment tank 23 is mixed by flowing a plurality of types of fuel solution having different Pu fissile rates, and the Pu fissile rate after mixing is set as the target Pu fissile rate.

  Next, with reference to FIG. 3, a series of flows regarding a spent nuclear fuel reprocessing method performed in the spent nuclear fuel reprocessing facility configured as described above will be described. FIG. 3 is a flowchart regarding a method for reprocessing spent nuclear fuel.

  When the fuel assembly containing the spent nuclear fuel whose component has been estimated in advance is supplied to the mechanical processing device 5, the mechanical processing device 5 converts the spent nuclear fuel contained in the supplied fuel assembly into the machine The mechanical processing step S <b> 1 to be exposed by the general processing is executed, and the spent nuclear fuel after the mechanical processing is supplied to the melting device 6. When the spent nuclear fuel is supplied to the melting device 6, the melting device 6 executes a melting step S2 for melting the supplied used nuclear fuel using nitric acid, and uses the dissolved spent nuclear fuel as a fuel solution. Supply toward the clarification device 7. When the fuel solution is supplied to the clarification device 7, the clarification device 7 executes a clarification step S <b> 3 to remove insoluble residues contained in the fuel solution, and supplies the fuel solution after clarification to the metering device 8. .

  When the fuel solution is supplied to the metering device 8, the metering device 8 causes the fuel solution to flow into the metering tank 21 and executes a metering step S4 in which the components of the fuel solution accumulated in the metering tank 21 are metered. . Thereby, the error of the component of the fuel solution estimated before the measurement and the component of the fuel solution after the measurement is corrected. The fuel solution after measurement is supplied to the buffer tank 22. Thereafter, the process is repeated from S1 to S4 for the kind of spent nuclear fuel to be reprocessed. Thereby, different types of fuel solution are stored in the plurality of buffer tanks 22.

  When a predetermined type of spent nuclear fuel becomes a fuel solution and is stored in the plurality of buffer tanks 22 and the measuring tanks 21, the measuring device 8 is configured so that the measured Pu fissile rate and the Pu fissile rate become the target Pu fissile rate. Based on the enrichment, fuel solutions having different Pu fissile ratios and Pu enrichments are supplied to the adjustment tank 23 one by one in order. That is, the metering device 8 performs a mixing step S5 in which the fuel solution having a high Pu fissile rate (high Pu fissile nuclear fuel) and the fuel solution having a low Pu fissile rate (low Pu fissile nuclear fuel) are mixed in the adjustment tank 23. Run. In addition, when supplying several types of fuel solution to the adjustment tank 23, it is preferable to supply in order from a fuel solution with a high Pu fissile rate. Thereby, the measuring device 8 can obtain the fuel solution having the target Pu fissile rate. The metering device 8 supplies the fuel solution collected in the adjustment tank 23 toward the separation device 9.

  When the fuel solution is supplied to the separation device 9, the separation device 9 executes a separation step S6 for extracting uranium and plutonium in the fuel solution as a U / Pu solution and a U solution. When the U / Pu solution and the U solution are extracted, the separation device 9 supplies the U / Pu solution and the U solution to the purification device 10.

  When the U / Pu solution and the U solution are supplied to the purification device 10, the purification device 10 executes a purification step S7 for purifying the U solution and the U / Pu solution into a nitric acid U solution and a nitric acid U / Pu solution. Thereafter, the purification device 10 supplies the purified nitric acid U solution and the nitric acid U / Pu solution to the enrichment adjusting device 11. When the nitric acid U solution and the nitric acid U / Pu solution are supplied to the enrichment adjusting device 11, the enrichment adjusting device 11 mixes the solution after heating and concentrating the nitric acid U solution and the nitric acid U / Pu solution. Then, the enrichment adjustment step S8 for adjusting the Pu enrichment is performed. Thereafter, the enrichment adjusting device 11 supplies the nitric acid U solution and the enriched nitric acid U / Pu solution toward the denitration device 12.

  When the nitric acid U solution and the enriched nitric acid U / Pu solution are supplied to the denitration device 12, the denitration device 12 denitrates the nitric acid U solution and the nitric acid U / Pu solution to obtain uranium oxide and uranium / A denitration step S9 for generating plutonium oxide is performed. Thereafter, the denitration device 12 supplies uranium oxide and uranium / plutonium oxide to the fuel production device 13. When uranium oxide and uranium / plutonium oxide are supplied to the fuel production apparatus 13, the fuel production apparatus 13 temporarily stores the uranium oxide and uranium / plutonium oxide in a storage container, and stores the stored uranium oxide and uranium. / After appropriately mixing plutonium oxide, a fuel production step S10 for producing a loading fuel is performed.

  As described above, according to the configuration of the first embodiment, since the high Pu fissile nuclear fuel and the low Pu fissile nuclear fuel can be mixed in the metering device 8 in which the mixing device is integrated, the fuel solution after mixing is mixed. The Pu fissile rate can be set as the target Pu fissile rate. As a result, the reprocessing facility 1 equipped with the metering device 8 can produce a desired loading fuel according to demand from the fuel solution (nuclear fuel) having the target Pu fissile rate.

  Moreover, according to the structure of Example 1, since the several buffer tank 22 was provided, since one buffer tank 22 can be allocated with respect to one type of fuel solution (spent nuclear fuel) to reprocess, The volume of the fuel solution stored in the buffer tank 22 can be suppressed.

  In the reprocessing facility 1 of the first embodiment, the mechanical processing device 5, the dissolving device 6 and the clarification device 7 are provided as one system, but the mechanical processing device 5, the dissolving device 6 and the clarification device 7 are provided. A plurality of systems may be provided to form a plurality of systems, and the fuel solution reprocessed in the plurality of systems may be allowed to flow into the plurality of buffer tanks 22 via the measuring tank 21.

  In the reprocessing facility 1 of the first embodiment, a plurality of types of fuel solutions are mixed in the adjustment tank 23. If it is not necessary to mix them, the adjustment is performed from the measurement tank 21 without flowing into the buffer tank 22. You may supply to the separation apparatus 9 after supplying to the tank 23. FIG.

  Moreover, in the reprocessing facility 1 of Example 1, after making it flow into the some buffer tank 22, it was mixed in the adjustment tank 23, However, Not only this but the structure shown in FIG. 4 may be sufficient. FIG. 4 is a schematic configuration diagram around the weighing device in the spent nuclear fuel reprocessing facility 1 according to the first modification. The weighing device 8 of the reprocessing facility 1 according to the modification 1 has a configuration in which the plurality of buffer tanks 22 are omitted, and the adjustment tank (buffer mixing unit) 23 also functions as the buffer tank 22. Here, the spent nuclear fuel reprocessing method performed in the spent nuclear fuel reprocessing facility 1 according to Modification 1 will be described.

  In the reprocessing method according to the modified example 1, when one type of spent nuclear fuel is executed from the mechanical processing step S1 to the weighing step S4 among the plurality of types of spent nuclear fuel to be reprocessed, the weighing device 8 One type of fuel solution accumulated in the measuring tank 21 is supplied to the adjustment tank 23. Thereafter, when another kind of spent nuclear fuel is executed from the mechanical processing step S1 to the weighing step S4, the weighing device 8 converts the other kind of fuel solution accumulated in the weighing tank 21 to 1 It supplies to the adjustment tank 23 where the seed | species fuel solution accumulated. This process is repeated for the type of spent nuclear fuel to be reprocessed.

  As described above, even in the configuration according to the modified example 1, since the high Pu fissile nuclear fuel and the low Pu fissile nuclear fuel can be mixed in the measuring device 8, the Pu fissile ratio of the fuel solution after mixing is set to the target Pu. It can be a fissile rate. Moreover, according to the structure which concerns on the modification 1, since the several buffer tank 22 can be omitted, a reprocessing facility can be made into a simple structure. In this case, it is preferable to manage the supply amount of the fuel solution supplied to the adjustment tank 23 in advance so that the fuel solution after mixing has the target Pu fissile rate.

  Next, the spent nuclear fuel reprocessing facility 31 and the reprocessing method according to the second embodiment will be described with reference to FIG. FIG. 5 is a schematic configuration diagram around a separation device in a spent nuclear fuel reprocessing facility according to the second embodiment. In the second embodiment, only different parts will be described in order to avoid redundant description. In the spent nuclear fuel reprocessing facility 1 according to the first embodiment, the mixing device and the metering device 8 are integrated. However, in the spent nuclear fuel reprocessing facility 31 according to the second embodiment, the mixing device and the separation device 35 are combined. Is united. Hereinafter, the separation device 35 provided integrally with the mixing device will be described in detail with reference to FIG.

  The separation device 35 includes a supply tank 41, a co-decontamination apparatus 42, a waste liquid receiving tank 43, a Pu extraction apparatus 44, a U / Pu solution receiving tank 45, a U extraction apparatus 46, a U solution receiving tank 47, and a plurality of A buffer tank (buffer section) 48 and an adjustment tank (mixing section) 49 are provided.

  The supply tank 41 is connected to the measuring device 8 and stores the fuel solution supplied from the measuring device 8. The fuel solution stored in the supply tank 41 is supplied toward the co-decontamination device 42.

  The co-decontamination apparatus 42 generates an uranium / plutonium loaded solvent, which is an extraction solvent containing uranium and plutonium, by adding an extraction solvent to the fuel solution, while supplying a cleaning solution to the fuel solution. Isolate impurities. That is, the fuel solution is separated in the co-decontamination apparatus 42 into a uranium / plutonium loading solvent and a cleaning liquid containing impurities other than uranium and plutonium. The co-decontamination device 42 supplies the uranium / plutonium loaded solvent toward the Pu extraction device 44 and supplies the cleaning liquid containing impurities toward the waste liquid receiving tank 43. The waste liquid receiving tank 43 stores a cleaning liquid containing impurities separated in the co-decontamination apparatus 42.

  The Pu extraction device 44 inputs the Pu back-extraction solvent into the uranium / plutonium loading solvent supplied from the co-decontamination device 42 and separates it into the U / Pu solution and the uranium loading solvent after the plutonium extraction. Then, the Pu extraction device 44 supplies the U / Pu solution toward the U / Pu solution receiving tank 45, and supplies the uranium loaded solvent toward the U extraction device 46. The U / Pu solution receiving tank 45 stores the U / Pu solution separated in the Pu extraction device 44 and is connected to a plurality of buffer tanks 48.

  The U extraction device 46 puts the U back-extraction solvent into the uranium loaded solvent supplied from the Pu extraction device 44 and separates it into the U solution and the extraction solvent after uranium extraction. Then, the U extraction device 46 supplies the U solution toward the U solution receiving tank 47. The U solution receiving tank 47 stores the U solution separated in the U extraction device 46.

  The plurality of buffer tanks 48 are respectively connected to the U / Pu solution receiving tank 45 and temporarily store the U / Pu solution. In the second embodiment, as in the first embodiment, one or more buffer tanks 48 may be used. The U / Pu solution stored in the buffer tank 48 is supplied toward the adjustment tank 49. The plurality of buffer tanks 48 can store U / Pu solutions having different Pu fissile ratios. The adjustment tank 49 is connected to the U / Pu solution receiving tank 45 and each buffer tank 48 and stores the U / Pu solution supplied from the U / Pu solution receiving tank 45 and each buffer tank 48. A plurality of types of U / Pu solutions having different Pu fissile ratios are mixed into the adjustment tank 49 and mixed, and the Pu fissile ratio after mixing is set as the target Pu fissile ratio.

  The U solution stored in the U solution receiving tank 47 and the U / Pu solution stored in the adjustment tank 49 are supplied toward the purifier 10. The purifier 10 has a U purifier 10a for purifying the U solution and a U / Pu purifier 10b for purifying the U / Pu solution. The U purifier 10a purifies the U solution into a nitric acid U solution. In the / Pu purification apparatus 10b, the U / Pu solution is purified to a nitric acid U / Pu solution.

  As for the spent nuclear fuel reprocessing method performed in the spent nuclear fuel reprocessing facility 31 configured as described above, the mixing step S5 shown in FIG. 3 is after the separation step S6. More specifically, after performing the steps from the mechanical processing step to the metering step, when the fuel solution after measurement is supplied to the separation device 35, the separation device 35 supplies the fuel solution to the supply tank. The fuel solution accumulated in the supply tank 41 is supplied toward the co-decontamination device 42. The co-decontamination device 42 separates the fuel solution into a uranium / plutonium loading solvent and a cleaning solution containing impurities. Thereafter, the co-decontamination device 42 supplies the uranium / plutonium loaded solvent to the Pu extraction device 44.

  The Pu extraction device 44 executes a separation step of separating the supplied uranium / plutonium loading solvent into a U / Pu solution and a uranium loading solvent. Thereafter, the Pu extraction device 44 supplies the U / Pu solution to the U / Pu solution receiving tank 45. When the U / Pu solution is supplied to the U / Pu solution receiving tank 45, the separation device 35 supplies the U / Pu solution from the U / Pu solution receiving tank 45 toward the buffer tank 48. Thereafter, the steps from the mechanical treatment step to the separation step are repeated for the kind of spent nuclear fuel to be reprocessed. Accordingly, different types of U / Pu solutions are stored in the U / Pu solution receiving tank 45 and the plurality of buffer tanks 48.

  When different types of U / Pu solutions are stored in the plurality of buffer tanks 48 and the U / Pu solution receiving tanks 45, the separation device 35 causes the Pu fissile ratio and the Pu enrichment degree after measurement so that the target Pu fissile ratio is obtained. Based on the above, U / Pu solutions having different Pu fissile rates and Pu enrichments are supplied to the adjustment tank 49 one by one in order. That is, the separation device 35 mixes the U / Pu solution having a high Pu fissile rate (high Pu fissile nuclear fuel) and the U / Pu solution having a low Pu fissile rate (low Pu fissile nuclear fuel) in the adjustment tank 49. Execute. Thereby, the separation device 35 can obtain the U / Pu solution having the target Pu fissile rate. Then, the separation device 35 supplies the U / Pu solution accumulated in the adjustment tank 49 toward the purification device 10.

  As described above, according to the configuration of the second embodiment, the high Pu fissile nuclear fuel and the low Pu fissile nuclear fuel can be mixed in the separation apparatus 35 in which the mixing apparatus is integrated. The Pu fissile rate of the solution can be the target Pu fissile rate. As a result, the reprocessing facility 31 including the separation device 35 can manufacture a desired loading fuel according to the demand. In the second embodiment, the configuration of the first modification of the first embodiment can be applied.

  Next, the spent nuclear fuel reprocessing facility 51 and the reprocessing method according to the third embodiment will be described with reference to FIG. FIG. 6 is a schematic configuration diagram around the enrichment adjusting device in the spent nuclear fuel reprocessing facility according to the third embodiment. In the third embodiment, only different parts will be described in order to avoid redundant description. In the spent nuclear fuel reprocessing facility 1 according to the first embodiment, the mixing device and the metering device 8 are integrated. However, in the spent nuclear fuel reprocessing facility 51 according to the third embodiment, the mixing device and the enrichment adjustment are adjusted. The device 55 is integrated. Hereinafter, the enrichment adjusting device 55 provided integrally with the mixing device will be specifically described with reference to FIG.

  The enrichment adjusting device 55 includes a U concentrator 61, a U solution receiving tank 62, a U dispensing tank 63, a U / Pu concentrating apparatus 64, a U / Pu solution receiving tank 65, a U / Pu dispensing tank 66, An adjustment tank (mixing unit) 67 and a plurality of buffer tanks (buffer units) 68 are provided.

  The U concentrator 61 heats and concentrates the nitric acid U solution supplied from the U purifier 10 a of the purifier 10. The U concentrator 61 supplies the concentrated nitric acid U solution toward the U solution receiving tank 62. The U solution receiving tank 62 stores the concentrated nitric acid U solution, and is connected to the U dispensing tank 63. The enrichment adjusting device 55 supplies the nitric acid U solution accumulated in the U solution receiving tank 62 toward the U dispensing tank 63. The U dispensing tank 63 dispenses the supplied nitric acid U solution toward the denitration device 12.

  The U / Pu concentrator 64 heats and concentrates the nitric acid U / Pu solution supplied from the U / Pu purifier 10b of the purifier 10. The U / Pu concentrator 64 supplies the concentrated nitric acid U / Pu solution toward the U / Pu solution receiving tank 65 and the plurality of buffer tanks 68. The U / Pu solution receiving tank 65 stores the concentrated nitric acid U / Pu solution, and is connected to the adjustment tank 67.

  The plurality of buffer tanks 68 are respectively connected to the U / Pu concentrating device 64 and temporarily store the nitric acid U / Pu solution. In the third embodiment, as in the first embodiment, one or more buffer tanks 68 may be used. The nitric acid U / Pu solution stored in each buffer tank 68 is supplied toward the adjustment tank 67. The nitric acid U / Pu solutions stored in the plurality of buffer tanks 68 have different Pu fissile rates.

  The adjustment tank 67 is connected to the U solution receiving tank 62, the U / Pu solution receiving tank 65, and each buffer tank 68, and from the nitric acid U solution supplied from the U solution receiving tank 62, the U / Pu solution receiving tank 65, and each buffer tank 68. The supplied nitric acid U / Pu solution is stored. In the adjustment tank 67, the nitric acid U solution and a plurality of different nitric acid U / Pu solutions having different Pu fissile rates are mixed, so that the Pu fissile rate of the mixed nitric acid U / Pu solution is set as the target Pu fissile rate. . At this time, the Pu enrichment of the mixed nitric acid U / Pu solution is also set as the target Pu enrichment.

  The nitric acid U / Pu solution stored in the adjustment tank 67 is supplied toward the U / Pu discharge tank 66. The U / Pu discharge tank 66 discharges the supplied nitric acid U / Pu solution toward the denitration apparatus 12.

  The nitric acid U solution dispensed from the U dispensing tank 63 and the nitric acid U / Pu solution dispensed from the U / Pu dispensing tank 66 are supplied toward the denitration apparatus 12. The denitration device 12 has a U denitration device 12a for denitrating the nitric acid U solution and a U / Pu denitration device 12b for denitrating the nitric acid U / Pu solution. In the U denitration device 12a, the nitric acid U solution is uranium oxide, In the U / Pu denitration apparatus 12b, the nitric acid U / Pu solution is made uranium / plutonium oxide.

  As for the spent nuclear fuel reprocessing method performed in the spent nuclear fuel reprocessing facility 51 configured as described above, the mixing step S5 shown in FIG. 3 is after the enrichment adjusting step S8. More specifically, after performing the steps from the mechanical treatment step to the purification step, the purified nitric acid U solution and nitric acid U / Pu solution are supplied to the enrichment adjusting device 55. Then, the enrichment adjusting device 55 concentrates the nitric acid U solution in the U concentrating device 61, supplies the concentrated nitric acid U solution toward the U solution receiving tank 62, and the U / Pu concentrating device 64 supplies the nitric acid U / The Pu solution is concentrated, and the concentrated nitric acid U / Pu solution is supplied toward the U / Pu solution receiving tank 65 or the buffer tank 68. Thereafter, the steps from the mechanical treatment step to the enrichment adjustment step are repeated for the kind of spent nuclear fuel to be reprocessed. Thereby, different types of nitric acid U / Pu solutions are stored in the U / Pu solution receiving tank 65 and the plurality of buffer tanks 68.

  When different types of nitric acid U / Pu solutions are stored in the plurality of buffer tanks 68 and the U / Pu solution receiving tanks 65, the enrichment adjusting device 55 measures the target Pu fissile rate and the target Pu enrichment. Based on the subsequent Pu fissile rate and Pu enrichment, nitric acid U / Pu solutions having different Pu fissile rates and Pu enrichment are fed one by one toward the adjustment tank 67 one by one. Further, the enrichment degree adjusting device 55 supplies the nitric acid U solution based on the Pu enrichment after measurement so as to achieve the target Pu enrichment. In other words, the enrichment adjusting device 55 is configured to adjust the nitric acid U / Pu solution (high Pu fissile nuclear fuel) with a high Pu fissile rate and the nitric acid U / Pu solution (low Pu fissile nuclear fuel) with a low Pu fissile rate into the adjusting tank 67. A mixing step of mixing is performed. Thereby, the enrichment degree adjusting device 55 can obtain the nitric acid U / Pu solution having the target Pu fissile rate and the target Pu enrichment level. The enrichment adjusting device 55 supplies the nitric acid U / Pu solution accumulated in the adjusting tank 67 toward the denitration device 12.

  As described above, according to the configuration of the third embodiment, in the enrichment adjusting device 55 in which the mixing device is integrated, the high Pu fissile nuclear fuel and the low Pu fissile nuclear fuel can be mixed. The Pu fissile rate of the nitric acid U / Pu solution can be set as the target Pu fissile rate. In addition, since the enrichment adjusting device 55 can also adjust the Pu enrichment during mixing, the Pu enrichment of the nitric acid U / Pu solution after mixing can be set as the target Pu enrichment. . As a result, the reprocessing facility 51 provided with the enrichment adjusting device 55 can produce a desired loading fuel according to the demand. In the third embodiment, the configuration of the first modification of the first embodiment can also be applied.

  Next, a spent nuclear fuel reprocessing facility 71 and a reprocessing method according to the fourth embodiment will be described with reference to FIG. FIG. 7 is a schematic configuration diagram around a fuel manufacturing apparatus in a spent nuclear fuel reprocessing facility according to a fourth embodiment. In the fourth embodiment, only different parts will be described to avoid redundant description. In the spent nuclear fuel reprocessing facility 1 according to the first embodiment, the mixing device and the metering device 8 are integrated. However, in the spent nuclear fuel reprocessing facility 71 according to the fourth embodiment, the mixing device and the fuel manufacturing device 75 are combined. Together. Hereinafter, the fuel production device 75 provided integrally with the mixing device will be described in detail with reference to FIG.

  The fuel manufacturing apparatus 75 includes a plurality of U storage containers (buffer units) 81, a plurality of U / Pu storage containers (buffer units) 82, a mixing unit 83, and a nuclear fuel manufacturing unit 84.

  The plurality of U storage containers 81 store uranium oxide supplied from the U denitration device 12a of the denitration device 12 for each type of spent nuclear fuel. The plurality of U / Pu storage containers 82 stores uranium / plutonium oxide supplied from the U / Pu denitration device 12b of the denitration device 12 for each type of spent nuclear fuel.

  The mixing unit 83 mixes uranium oxide stored in the plurality of U storage containers 81 and uranium / plutonium oxide stored in the plurality of U / Pu storage containers 82. In the mixing unit 83, uranium oxide and uranium / plutonium oxides having a plurality of different Pu fissile ratios are mixed, so that the Pu fissile ratio of the uranium / plutonium oxide after mixing is set as the target Pu fissile ratio. . At this time, the Pu enrichment of the uranium / plutonium oxide after mixing is also set as the target Pu enrichment.

  The nuclear fuel production unit 84 produces a fuel for loading such as fuel pellets from the mixed uranium / plutonium oxide.

  In the spent nuclear fuel reprocessing method performed in the spent nuclear fuel reprocessing facility 71 configured as described above, the mixing step S5 shown in FIG. 3 is performed after the denitration step S9. More specifically, the uranium oxide and the uranium / plutonium oxide after denitration are supplied to the fuel production apparatus 75 after performing the steps from the mechanical treatment process to the denitration process.

  The fuel production device 75 removes uranium / plutonium oxide from the U / Pu storage container 82 based on the measured Pu fissile rate and the Pu enrichment so as to achieve the target Pu fissile rate and the target Pu enrichment. While supplying toward the mixing part 83, uranium oxide is supplied toward the mixing part 83 from the U storage container 81. That is, the fuel production device 75 mixes uranium / plutonium oxide having a high Pu fissile rate (high Pu fissile nuclear fuel) and uranium / plutonium oxide having a low Pu fissile rate (low Pu fissile nuclear fuel) in the mixing unit 83. A mixing step is performed. Thereby, the fuel manufacturing apparatus 75 can obtain the uranium / plutonium oxide having the target Pu fissile rate and the target Pu enrichment. Then, the fuel production device 75 supplies the uranium / plutonium oxide accumulated in the mixing unit 83 to the nuclear fuel production unit 84, and the nuclear fuel production unit 84 produces a loading fuel such as MOX fuel.

  As described above, according to the configuration of the fourth embodiment, the high Pu fissile nuclear fuel and the low Pu fissile nuclear fuel can be mixed in the fuel manufacturing apparatus 75 in which the mixing apparatus is integrated. The Pu fissile rate of the plutonium oxide can be set as the target Pu fissile rate. In addition, since the fuel enrichment device 75 can also adjust the Pu enrichment at the time of mixing, the Pu enrichment of the uranium / plutonium oxide after the mixing can be set as the target Pu enrichment. Thereby, the reprocessing facility 71 equipped with the fuel production device 75 can produce a desired loading fuel according to the demand.

  In Examples 1 to 4, the mixing device is provided integrally with the metering device 8, the separation device 35, the enrichment degree adjusting device 55, and the fuel production device 75, so that the mixing step is performed in the metering step, the separation step, and the wealthy device. It was performed after the chemical conversion adjustment step and the denitration step. However, the present invention is not limited to this configuration, and it may be provided integrally with other devices such as the dissolving device 6 and the clarifying device 7, and the mixing step may be performed after the dissolving step and the clarifying step.

  When the mixing device is provided integrally with the melting device 6, the mixing step may be performed in the melting tank of the melting device 6.

  In the configurations of the first to fourth embodiments, the Pu fissile rate is mixed so as to become the target Pu fissile rate. However, the present invention is not limited to this configuration, and the uranium fissile rate (U fissile rate) becomes the target U fissile rate. May be mixed.

DESCRIPTION OF SYMBOLS 1 Reprocessing facility 5 Mechanical processing apparatus 6 Dissolution apparatus 7 Clarification apparatus 8 Weighing apparatus 9 Separation apparatus 10 Purification apparatus 11 Enrichment degree adjustment apparatus 12 Denitration apparatus 13 Fuel production apparatus 21 Measurement tank 22 Buffer tank 23 Adjustment tank 31 Reprocessing facility (Example 2)
35 Separator (Example 2)
41 Supply tank 42 Co-decontamination apparatus 43 Waste liquid receiving tank 44 Pu extraction apparatus 45 U / Pu solution receiving tank 46 U extraction apparatus 47 U solution receiving tank 48 Buffer tank (Example 2)
49 Adjustment tank (Example 2)
51 Reprocessing facility (Example 3)
55 Enrichment adjusting device (Example 3)
61 U Concentrator 62 U Solution Receiving Tank 63 U Dispensing Tank 64 U / Pu Concentrator 65 U / Pu Solution Receiving Tank 66 U / Pu Dispensing Tank 67 Adjustment Tank (Example 3)
68 Buffer tank (Example 3)
71 Reprocessing facility (Example 4)
75 Fuel production device (Example 4)
81 U storage container 82 U / Pu storage container 83 Mixing section 84 Nuclear fuel production section

Claims (6)

A method for reprocessing spent nuclear fuel in which different types of spent nuclear fuel are reprocessed in a single reprocessing facility to produce new nuclear fuel,
The fissile rate of the fissionable radioisotope of the nuclear fuel material contained in the nuclear fuel becomes the target fissile rate, and the enrichment degree of the nuclear fuel material contained in the loading fuel manufactured using the new nuclear fuel is: A mixing step of mixing the high fissile nuclear fuel with the high fissile rate and the low fissile nuclear fuel with the low fissile rate so as to be equal to or lower than the upper limit enrichment,
The mixing device used in the mixing step is
One or more buffer portions each storing a plurality of types of the spent nuclear fuel having different fissile rates;
Have a, a mixing unit for mixing the spent nuclear fuel to be supplied from said each buffer section,
In the mixing step, a plurality of types of the spent nuclear fuel stored in the one or more buffer units are supplied to the mixing unit one by one in order from the spent nuclear fuel having a high fissile rate. A feature of a method for reprocessing spent nuclear fuel. 2. The method for reprocessing spent nuclear fuel according to claim 1, wherein the fissile radioisotope contains at least one of plutonium 239 and plutonium 241. The nuclear fuel material is plutonium;
The method for reprocessing spent nuclear fuel according to claim 1 or 2 , wherein the upper limit enrichment is 30%. A mechanical treatment step of exposing the spent nuclear fuel contained in the spent fuel assembly by mechanical treatment;
Using nitric acid, dissolving the spent nuclear fuel after mechanical treatment, and forming a fuel solution,
A clarification step of removing insoluble residues in the fuel solution;
A weighing step for weighing the fuel solution from which the insoluble residue has been removed;
A separation step of separating the solution containing the nuclear fuel material from the fuel solution after measurement;
A purification step for purifying the solution after separation;
An enrichment adjusting step of concentrating the solution after purification and adjusting the enrichment of the nuclear fuel material in the solution;
A denitration step of denitrating the fuel solution after the enrichment adjustment to generate the new nuclear fuel containing the nuclear fuel material;
A fuel production process for producing fuel for loading from the nuclear fuel after denitration,
The mixing step is between the mechanical treatment step and the dissolution step, between the dissolution step and the clarification step, between the clarification step and the measurement step, and between the measurement step and the separation step. , Between the separation step and the purification step, between the purification step and the enrichment adjustment step, between the enrichment adjustment step and the denitration step, and between the denitration step and the fuel production step. The spent nuclear fuel reprocessing method according to any one of claims 1 to 3 , wherein the method is performed between any of the above. A spent nuclear fuel reprocessing facility that reprocesses multiple different types of spent nuclear fuel to produce new nuclear fuel,
The fissile rate of the fissionable radioisotope of the nuclear fuel material contained in the nuclear fuel becomes the target fissile rate, and the enrichment degree of the nuclear fuel material contained in the loading fuel manufactured using the new nuclear fuel is: A mixing device for mixing the high fissile nuclear fuel with the high fissile rate and the low fissile nuclear fuel with the low fissile rate so as to be equal to or lower than the upper limit enrichment;
The mixing device includes:
One or more buffer portions each storing a plurality of types of the spent nuclear fuel having different fissile rates;
Have a, a mixing unit for mixing the spent nuclear fuel to be supplied from said each buffer section,
A plurality of types of spent nuclear fuel stored in the one or more buffer units are supplied to the mixing unit one by one in order from the spent nuclear fuel having the highest fissile rate. Nuclear fuel reprocessing facility. A mechanical processing device for exposing the spent nuclear fuel contained in the spent fuel assembly by mechanical processing;
Using nitric acid to dissolve the spent nuclear fuel after mechanical treatment, and a fuel dissolving solution;
A refining device for removing insoluble residues in the fuel solution;
A weighing device for weighing the fuel solution from which the insoluble residue has been removed;
A separation device for separating the solution containing the nuclear fuel material from the fuel solution after measurement;
A purification device for purifying the solution after separation;
An enrichment adjusting device for concentrating the solution after purification and adjusting the enrichment of the nuclear fuel material in the solution;
A denitration device for denitrating the solution after the enrichment adjustment and generating new nuclear fuel containing the nuclear fuel material;
A fuel production device that produces fuel for loading from the nuclear fuel after denitration,
The mixing device is any one of the mechanical processing device, the dissolving device, the clarification device, the metering device, the separation device, the purification device, the enrichment adjusting device, the denitration device, and the fuel production device. The spent nuclear fuel reprocessing facility according to claim 5 , wherein the facility is provided in the apparatus.

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