JP2014089155A - Reprocessing system and reprocessing method of spent nuclear fuel - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a reprocessing system and reprocessing method of a spent nuclear fuel capable of increasing the plutonium enrichment of a nuclear fuel without reducing the nuclear diffusion resistance.SOLUTION: The reprocessing system of a spent nuclear fuel extracted from a nuclear reactor includes: separating means for separating uranium from a nitric acid solution in which the spent nuclear fuel is dissolved and obtaining plutonium nitrate solution; and mixing means for mixing an inactive base material to the plutonium nitrate solution that is fed from the separating means to manufacturing means of a nuclear fuel by a sol-gel method.

Description

  The present invention relates to a spent nuclear fuel reprocessing system and a reprocessing method.

  In recent years, an inert matrix fuel that becomes a stable compound even after combustion (for example, see Non-Patent Document 1) and a vibration filling method suitable for remote fuel production in a cell capable of handling a high-dose substance (for example, Research and development of non-patent document 2) are being conducted.

T. Shiratori, T. Yamashita, T. Ohmichi, A. Yasuda, K. Watarumi, "Preparation of rock-like oxide fuels for the irradiation test in the Japan Research Reactor No. 3", Journal of Nuclear Materials, 1999, Vol .274, p. 40-46. Masaya Kamitani, Hisao Kojima, Yoshihiko Shinoda, "Advanced wet plant design research (III)", Tokai Works, Power Reactor and Nuclear Fuel Development Corporation, 1998/03. WILLIAM S. CHARLTON and 4 others, "PROLIFERATION RESISTANCEASSESSMENT METHODOLOGY FOR NUCLEAR FUEL CYCLES", NUCLEAR TECHNOLOGY, 2007 FEB, VOL.157, p.143-156.

  Since Japan has concluded an NPT (Convention on Non-Proliferation of Nuclear Weapons), plutonium cannot exist alone for safeguards. For this reason, in the current reprocessing method, the uranium nitrate solution of the plutonium nitrate solution before denitration is mixed so that the final product becomes a mixed oxide of uranium and plutonium. Therefore, the product from the current reprocessing method cannot produce a plutonium-enriched fuel exceeding the mixing ratio, thereby narrowing the width of the core design.

  Then, this application makes it a subject to provide the reprocessing system and the reprocessing method of the spent nuclear fuel which can raise the plutonium enrichment degree of a nuclear fuel, without reducing proliferation resistance.

  In order to solve the above problems, the present invention focuses on the fact that the sol-gel method can directly use a nitric acid solution. That is, in the present invention, the plutonium nitrate solution generated in the reprocessing is guided to the means for producing nuclear fuel by the sol-gel method without denitration, and the inert base material is mixed before being led to the means for producing nuclear fuel. .

  Specifically, the present invention is a reprocessing system for spent nuclear fuel taken out from a nuclear reactor, separating uranium from a nitric acid solution in which the spent nuclear fuel is dissolved to obtain a plutonium nitrate solution, and the separation Mixing means for mixing an inert base material with the plutonium nitrate solution sent from the means to the means for producing nuclear fuel by the sol-gel method.

  Here, the inert base material is a material mixed with the plutonium nitrate solution for the purpose of maintaining nuclear diffusion resistance, and for example, a chemically and nuclearly inert material is suitable.

In the case of the above reprocessing system, the inert matrix is mixed with the plutonium nitrate solution sent to the means for producing nuclear fuel, so even if the nitric acid in the solution is lost, the plutonium oxide does not exist alone, and the nuclear diffusion Resistance is maintained. For this reason, by using the above-mentioned reprocessing system, it is possible to produce a fuel having a plutonium enrichment of 50% or more, for example, without reducing the proliferation resistance.

  In addition, since the concentration of the plutonium nitrate solution before denitration generated in the existing (current) reprocessing facility is already in the reduced speed region, the concentration treatment is performed to obtain the solution concentration required by the sol-gel method, which is a kind of nuclear fuel production method. If applied, the multiplication factor is further lowered, and safety can be improved from the viewpoint of criticality.

  The mixing unit may mix the inert base material before evaporating and concentrating the plutonium nitrate solution sent from the separating unit to the manufacturing unit. If inert matrix is mixed before evaporative concentration, plutonium oxide does not exist alone even if nitric acid in the solution is lost due to evaporative concentration, so nuclear diffusion due to increased plutonium concentration It is possible to prevent a decrease in resistance.

  The mixing unit may mix the inert base material with the plutonium nitrate solution at a mixing ratio at which the plutonium nitrate solution sent to the manufacturing unit has a prescribed nuclear diffusion resistance. .

  Here, the prescribed proliferation resistance is a value that makes it difficult to use plutonium for nuclear weapons, and for example, a value prescribed in safeguards can be applied. By mixing the inert base material with the plutonium nitrate solution at such a mixing ratio, the mixing means can prevent diversion of plutonium to a nuclear weapon.

  In addition, the mixing means further converts the uranium nitrate solution in which the uranium separated in the separation means is dissolved into the plutonium nitrate solution at a mixing ratio at which the nuclear fuel produced by the production means has a desired plutonium enrichment. You may mix.

  In the above reprocessing system, the proliferation resistance is ensured by mixing the inert base material, so that it is possible to produce nuclear fuel with a desired plutonium enrichment, thereby widening the range of core design. It is possible.

  The present invention can also be understood from the aspect of the method. That is, the present invention is a method for reprocessing spent nuclear fuel taken out of a nuclear reactor, for example, a separation step of separating uranium from a nitric acid solution in which the spent nuclear fuel is dissolved to obtain a plutonium nitrate solution, and the separation And a mixing step of mixing an inert base material with the plutonium nitrate solution sent from the step to the nuclear fuel production step by the sol-gel method.

  With the spent nuclear fuel reprocessing system and the reprocessing method according to the present invention, it is possible to increase the plutonium enrichment of the nuclear fuel without reducing the proliferation resistance.

  That is, with the spent nuclear fuel reprocessing system and the reprocessing method according to the present invention, the plutonium nitrate solution generated in the reprocessing is guided to a means for producing nuclear fuel by the sol-gel method without denitration. Since the inert base material is mixed before being introduced to the means, it is possible to increase the plutonium enrichment of the nuclear fuel without reducing the proliferation resistance. Thus, for example, it is possible to produce a fuel with a plutonium enrichment of 50% or more without reducing the proliferation resistance.

It is an example of the block diagram of the reprocessing system which concerns on embodiment. It is an example of the flowchart of the reprocessing method implement | achieved by the reprocessing system. It is an example of the flowchart at the time of manufacturing a fuel by the sol gel method (external gelation method). It is an example of the graph which showed the relationship between a plutonium density | concentration and an effective multiplication factor. It is an example of the block diagram of the present reprocessing system. It is an example of the flowchart of the reprocessing method implement | achieved by the present reprocessing system. It is an example of the flowchart in the case of manufacturing an isolated plutonium fuel following the current reprocessing method. It is an example of the graph which compared about proliferation resistance. It is an example of the graph which compared the criticality. It is an example of the graph which compared the extinction rate of Pu-239. It is an example of the flowchart at the time of manufacturing a fuel by the sol-gel method (internal gelation method).

  Hereinafter, embodiments of the present invention will be described. Embodiment shown below is one aspect | mode of this invention, and does not limit the technical scope of this invention to the following aspects.

FIG. 1 is an example of a configuration diagram of a reprocessing system according to the present embodiment. The reprocessing system 1 is a system for recovering uranium and plutonium from spent nuclear fuel. The reprocessing system 1 is a system that manages an acceptance / storage process, a shearing / dissolution process, a separation process, a purification process, a denitration process, a product storage I process, an adjustment / manufacturing process, and a product storage II process. The acceptance / storage process is a process in which the spent fuel assembly 2 is received and stored in the storage facility 3. The shearing / dissolution process is a process in which the fuel assembly 2 is sheared by the shearing equipment 4 and then dissolved by the nitric acid by the melting equipment 5. In the separation step, the FP (fissionable product) separation facility 6 separates FP from the nitric acid solution, and the nitric acid solution is further separated from the uranium / plutonium separation facility 7 (an example of “separation means” in the present application). And a plutonium nitrate solution. The purification step is a step of purifying the uranium nitrate solution with the uranium purification facility 8 and purifying the plutonium nitrate solution with the plutonium purification facility 9. The denitration process is a process of denitrating the uranium nitrate solution with the uranium denitration facility 10. The product storage I process is a process of storing the commercialized uranium oxide in the uranium storage facility 11. The adjustment / manufacturing process is a process for preparing a plutonium fuel by adjusting a plutonium nitrate solution in the plutonium adjusting / manufacturing facility 12 (which is an example of the “mixing means” in the present application). The product storage II step is a step of storing the commercialized plutonium fuel in the fuel storage facility 13.

  FIG. 2 is an example of a flowchart of a reprocessing method realized by the reprocessing system 1. The reprocessing system 1 receives and stores the spent nuclear fuel conveyed from the light water reactor or fast reactor as the fuel assembly (S101). The reprocessing system 1 finely shears the stored spent fuel assembly (S102), and dissolves the fissure fragments with nitric acid (S103). Strips such as cladding tubes that do not dissolve in nitric acid are separated in a nitric acid solution in which the fuel is dissolved. The reprocessing system 1 appropriately adjusts the temperature of the nitric acid solution in which uranium and plutonium are dissolved, and separates uranium and plutonium (S104).

  The reprocessing system 1 purifies a uranium nitrate solution obtained by separating plutonium from a nitric acid solution in which uranium or plutonium is dissolved (S105), denitrates it into uranium oxide (S106), and stores it in the uranium storage facility 11 ( S107).

Further, the reprocessing system 1 purifies the plutonium nitrate solution obtained by separating uranium from the nitric acid solution in which uranium and plutonium are dissolved (S108). In addition, the reprocessing system 1 prepares an inert base material in which the concentration and the like are appropriately adjusted by dissolving in a nitric acid solution (S109). The reprocessing system 1 appropriately mixes a purified plutonium nitrate solution with a uranium nitrate solution as necessary to obtain a desired plutonium enrichment, and further, an inert matrix so as to obtain an appropriate plutonium concentration. Are mixed (S110). An inert base material is a material that does not disappear even after processing such as sintering, and is a chemically and nuclearly inactive material. Specific examples of the material of the inert base material include yttrium stabilized zirconia. When yttrium stabilized zirconia is used as the inert base material, a zirconium nitrate solution or a yttrium nitrate solution is prepared and mixed so as to be easily mixed with the nitric acid solution. Note that the mixing ratio of the inert base material is such that the nitric acid solution sent to the fuel production facility has, for example, the subcriticality defined in the production facility or the proliferation resistance defined in the safeguards. It can be determined as appropriate. Then, the reprocessing system 1 performs a process such as evaporation concentration on the nitric acid solution in which uranium, plutonium, and an inert base material are mixed to appropriately adjust the concentration and the like (S111), and performs fuel production by the sol-gel method ( S112).

Specifically, for example, the following process can be applied to the above-described fuel production process (S112). FIG. 3 is an example of a flowchart for producing fuel by the sol-gel method (external gelation method). The reprocessing system 1 mixes a thickener with a mixed nitric acid solution of uranium and plutonium mixed with an inert base material (hereinafter referred to as “U / Pu solution”) to adjust the viscosity (S201). The thickener has a viscosity that allows the U · Pu solution to be appropriately dropped from a nozzle for dropping the aqueous ammonia solution. For example, PVA (polyvinyl alcohol) can be applied. When PVA is used as a thickener, a polymer film is prevented from being formed on the surface of the aqueous solution by previously mixing a water-soluble cyclic ether such as THFA (Tetrahydrofurfuryl alcohol) with the polymer aqueous solution. It is desirable to maintain a uniform dissolved state of the polymer in the aqueous solution.

  The reprocessing system 1 drops the U / Pu solution mixed with the thickener toward the ammonia water from the vibrating nozzle (S202). The U / Pu solution dripped from the nozzle is spheroidized by surface tension during dropping.

  The U / Pu solution dropped from the nozzle falls into the ammonia water, and gelation proceeds to the inside of the droplets in the process of settling in the ammonia water (S203). Since the U / Pu solution dropped from the nozzle is spheroidized by the surface tension during the dropping, it is gelled into a spherical shape in the process of settling in the ammonia water.

  The reprocessing system 1 wash | cleans the gel ball which gelatinized in the process of settling in ammonia water with a washing | cleaning liquid (S204). For example, when water or alcohol is used as the cleaning liquid, ammonium nitrate or ammonia water is generated.

  The reprocessing system 1 dries the gel balls in a wet state by washing (S205). The gel balls may be dried by any method. For example, the gel balls may be naturally dried or may be forcibly dried by applying warm air (for example, air having a temperature of about 350 to 770 K).

  The reprocessing system 1 sinters the dried gel balls (S206). The gel balls may be sintered by any method. For example, the atmosphere in which the gel balls are arranged is set to a high temperature (for example, about 1670K), and the properties of the sintered body (porosity, etc.) are desired. The pressure of the atmosphere in which the gel balls are arranged may be reduced so as to be. Sintered gel balls constitute fuel nuclei.

The reprocessing system 1 completes the fuel nucleus containing plutonium by performing the above-described series of processes. The inert base material mixed in the above-described step (S110) remains in the fuel core without disappearing even after subsequent processing such as sintering. Further, the process of denitrating the plutonium nitrate solution is not performed. Therefore, since plutonium oxide does not exist alone, it is possible to maintain nuclear diffusion resistance.

  In addition, by connecting the reprocessing process and the fuel manufacturing process directly by piping, plutonium is processed to the final oxide fuel sintered body by the same line in a series of processes from the reprocessing process to the fuel manufacturing process. Is possible. Such a process can be realized by directly using a plutonium nitrate solution generated by reprocessing because the fuel substance is dissolved in a nitric acid solution and used in fuel production by the sol-gel method. However, since the components of the plutonium nitrate solution generated by reprocessing and the plutonium nitrate solution required for fuel production by the sol-gel method are different, an adjustment step is required. Further, in the reprocessing system 1, evaporation and concentration are performed after mixing the inert base material. For example, in the case of the sol-gel method, a plutonium nitrate solution, a uranium nitrate solution, or a mixture thereof used for fuel production is 2 mol / lit. The inert base material added thereto needs to have a zirconium nitrate concentration of 7 mol / lit. , Yttrium nitrate was 2 mol / lit. Degree. Their composition depends on the design requirements. In the current (prior art) reprocessing method, distillation and concentration are performed after the purification step and before proceeding to the denitration step. In a typical facility, the enriched uranium concentration is about 350 gU / lit. Plutonium concentration of about 250 g Pu / lit. However, the reprocessing system 1 has a plutonium concentration of 480 g / lit. With a inert base material mixed in a typical design example. It is necessary to concentrate to the extent.

  Here, the criticality was evaluated. FIG. 4 is an example of a graph showing the relationship between the plutonium concentration and the effective multiplication factor. This evaluation assumed a system in which the solution was stored in a 10 cm thick container having an annular container with an infinite height (infinite ring) and an inner radius of 1 m. As is clear from the graph of FIG. 7, the criticality is 100 g / lit. The multiplication factor increases in proportion to the plutonium concentration, but is about 210 g / lit. The concentration reverses the trend and falls. This is due to the difference in the degree of neutron deceleration by the aqueous solution, which is a general reactor physics phenomenon. This 210 g / lit. This state is called optimum deceleration, and after that is a reduction speed region, and in the reduction speed region, the multiplication factor decreases in inverse proportion to the plutonium concentration. As described above, in a typical facility, the plutonium concentration of the plutonium nitrate solution generated by reprocessing is about 250 gPu / lit. The concentration is already in the reduced speed region. Therefore, when the concentration treatment for obtaining the solution concentration required by the sol-gel method (for example, 480 g Pu / lit.) Is performed on the U / Pu solution generated in the reprocessing system 1, the multiplication factor is further reduced. It is possible to improve safety from the viewpoint of criticality.

  Further, in the case of the reprocessing system 1 according to the above embodiment, since the evaporative concentration is performed after the inert base material is mixed with the U / Pu solution, the evaporative concentration is performed without mixing the inert base material with the U / Pu solution. Compared with the case of performing fuel, fuel can be produced while maintaining a state with high proliferation resistance.

  For example, the composition condition of the nitric acid solution used in the sol-gel method is plutonium nitrate 2 mol / lit. , Zirconium nitrate 7 mol / lit. Yttrium nitrate 2 mol / lit. Suppose that This composition ratio can be changed according to the final demand for fuel production, and is not a fixed value.

Here, when the inert base material is to be mixed after evaporation concentration, the concentration of the plutonium nitrate solution is set to 2 mol / lit. Higher concentration is required. If the mixed solution of plutonium nitrate and zirconium nitrate / yttrium mixed solution is 1: 1, each solution requires twice as many regulations, and plutonium nitrate 4 mol / l.
it. , Zirconium nitrate 14 mol / lit. Yttrium nitrate 4 mol / lit. It becomes. A plurality of combinations of such solution concentrations are conceivable, and the concentration of the plutonium nitrate solution is 2 mol / lit. To the saturation concentration which is the practical upper limit. The presence of a single plutonium nitrate solution with no inert base material in a high concentration is not preferable from the viewpoint of resistance to proliferation.

  The plutonium concentration was 2 mol / lit. The plutonium nitrate solution was mixed with an inert base material, and the U · Pu solution mixed with the inert base material was 2 mol / lit. In order to achieve this, crystalline zirconium nitrate and yttrium nitrate are required, but it is difficult to handle solids in a process of continuous treatment, and it is also difficult to control the concentration after dissolution.

  However, in the case of the reprocessing system 1, since the evaporative concentration is performed after the inert base material is mixed in the U / Pu solution, the evaporative concentration is performed without mixing the inert base material in the U / Pu solution. Compared to the above, it is possible to produce fuel while maintaining a state of high proliferation resistance, and it is possible to use an inert base material in a nitric acid solution that is easy to handle.

  FIG. 5 is an example of a configuration diagram of the current reprocessing system. The current reprocessing system 101 is a system that controls the receiving / storage process, the shearing / dissolving process, the separation process, the refining process, the denitration process, and the product storage process, like the reprocessing system 1 according to the above embodiment. However, in the denitration process of the current reprocessing system 101, denitration is performed not only for denitration of a uranium nitrate solution but also for a mixed solution of a uranium nitrate solution and a plutonium nitrate solution. In addition, in the product storage process of the current reprocessing system 101, not only commercialized uranium oxide but also a mixture of uranium oxide and plutonium oxide is stored as a product.

  FIG. 6 is an example of a flowchart of the reprocessing method realized by the current reprocessing system. The current reprocessing system 101 receives and stores the spent nuclear fuel conveyed from the light water reactor or fast reactor as the fuel assembly (S1101). The reprocessing system 101 finely shears the stored spent fuel assembly (S1102), and dissolves the fissure fragments with nitric acid (S1103). The reprocessing system 101 appropriately adjusts the temperature of the nitric acid solution in which uranium or plutonium is dissolved, and separates uranium and plutonium (S1104).

  The reprocessing system 101 purifies the uranium nitrate solution obtained by separating plutonium from the nitric acid solution (S1105), denitrates it into uranium oxide (S1106), and stores it (S1107).

  Further, the reprocessing system 101 purifies the plutonium nitrate solution from which uranium is separated from the nitric acid solution (S1108), and mixes the uranium nitrate solution with the purified plutonium nitrate solution so that the plutonium enrichment is 50% or less. The nitric acid solution is denitrated to form a uranium / plutonium mixed oxide (S1109) and then stored (S1110).

  When increasing the plutonium enrichment of nuclear fuel produced by reprocessing spent nuclear fuel, the following flow can be considered, for example, when the current reprocessing method of the reprocessing system 101 is followed.

FIG. 7 is an example of a flowchart for producing isolated plutonium fuel following the current reprocessing method. The processing from S1201 to S1207 in the flowchart shown in FIG. 7 is the same as the processing from S1101 to S1107 in the flowchart shown in FIG. In the case of producing an isolated plutonium fuel by following the reprocessing method performed by the current reprocessing system 101, the reprocessing system 101 purifies a plutonium nitrate solution obtained by separating uranium from a nitric acid solution in which uranium or plutonium is dissolved. (S1208) The plutonium nitrate solution is denitrated without mixing the uranium nitrate solution with the purified plutonium nitrate solution. Then, the reprocessing system 101 performs fuel production after storing and transporting the plutonium oxide obtained by denitrating the plutonium nitrate solution as necessary (S1210).

  Regarding nuclear proliferation resistance, as shown in Non-Patent Document 3, a quantitative evaluation method is proposed by Charleston et al. In this method, the proliferation resistance in each process is quantified from the viewpoint of convenience from the perspective of diverting nuclear fuel materials to weapons, difficulty in handling heat and radiation, and convenience in contact and transportation. ing. The proliferation resistance of the entire system can be evaluated by performing the synthesis by multiplying the proliferation resistance quantified in each process by the amount of material processed in each process and the weight of the time required for the process.

  FIG. 8 shows a case where plutonium is isolated by following the current reprocessing method (hereinafter referred to as Case 1) and a case where uranium and plutonium are mixed according to the current reprocessing method ( Hereinafter, it is an example of the graph which compared about each proliferation resistance of the case (henceforth case 3) which handles plutonium by the said reprocessing system 1). In addition, since it is after the separation process that the proliferation resistance differs for each case, FIG. 8 shows the proliferation resistance from the separation process to fuel production. The separation process is common in each case, and its proliferation resistance is about 0.4 (1 at the maximum). In the purification step, the proliferation resistance of this embodiment (Case 3) is slightly decreased (about 0.03) compared to the proliferation resistance of the conventional method. In the purification step, distillation concentration is performed as a preparation for the subsequent denitration step, but in this embodiment, 480 gU / lit. This is to perform concentration to a high concentration. Further, in this embodiment, the time of the purification process is about 800 hours and about twice as long as that of the conventional method, because the concentration of the plutonium nitrate solution is about twice, while the volume of the storage tank is constant, This is because the stay period in storage after concentration is long, and can be shortened by design change. In this embodiment (Case 3), it can be directly connected to fuel production after this purification step.

  On the other hand, in the conventional method, oxide powders of plutonium and uranium are produced by a denitration process, packed in a transportable container and stored as a product. In addition, the storage period is dominant as a period, and a general reprocessing facility is designed to be able to store for about two years. Here, the storage period is one year, but even if it is set to about 400 hours, the result does not change. In the denitration / storage process, the decrease in proliferation resistance is remarkably observed, and in case 2 it is reduced to about 0.24 and in case 1 it is reduced to about 0.19. In Case 1 and Case 2, plutonium stored in sealed pipes and containers until the refining process can be taken out by being refilled into a transportable container after the denitration process. Is the cause. In particular, in the case 1 in which plutonium is separated, the proliferation resistance is significantly reduced as compared with the case 2. Thus, in the conventional method, the diffusion resistance is lowered because the plutonium oxide exists in a form that can be taken out, but in the method of the present embodiment (case 3), the isolated plutonium can be handled. Nevertheless, high proliferation resistance can be ensured as compared with conventional methods (Case 1 and Case 2).

Thus, if it is the reprocessing system 1 which concerns on this embodiment, since the proliferation resistance at the time of fuel manufacture is high, it is not necessary to mix uranium. Moreover, if there is a demand from the core design, the plutonium enrichment can be set very widely including the use of plutonium alone. However, whether or not the isolated plutonium can be used depends on the proliferation resistance of the finally produced fuel. Because coated fuel particles used in gas-cooled reactors are highly proliferation-resistant fuel due to the difficulty of reprocessing, even if they use isolated plutonium alone, they are subject to restrictions on safeguards. There is a high probability that manufacturing is acceptable. If there is a problem with the proliferation resistance due to the use of isolated plutonium, an inert base material such as yttrium stabilized zirconia or thorium is used in the uranium nitrate solution as in the reprocessing system 1 according to the above embodiment. Instead, it is possible to produce a 100% plutonium enriched fuel that has a much higher proliferation resistance than the current light water reactor fuel reprocessing and MOX fuel. Further, the reprocessing system 1 according to the above embodiment is a modification of a part of the current reprocessing method. However, the reprocessing system 1 according to the present embodiment is modified to a part of the improved reprocessing (Co-processing method). Even if it is a form, the same effect can be obtained.

For example, in the current reprocessing method of light water reactor fuel, a denitration process is performed after mixing a plutonium nitrate solution and a uranium nitrate solution so that plutonium and uranium are in a one-to-one relationship based on safeguards. The reason for this is to prevent plutonium oxide from being present in a single state with low proliferation resistance. Therefore, in the current reprocessing method of light water reactor fuel, the plutonium enrichment of the mixed oxide (MOX) fuel that can be produced is limited to 50% at the maximum, which narrows the width of the core design. This restriction also applies when using the improved Co-processing method. Therefore, in order to produce a fuel having a plutonium enrichment exceeding 50%, a system capable of performing reprocessing and fuel production with high proliferation resistance is required. With the reprocessing system 1 according to the embodiment, it is possible to send the plutonium nitrate solution or the U / Pu solution from the reprocessing process to the fuel manufacturing process without reducing the proliferation resistance. Thus, for example, it is possible to produce a fuel with a plutonium enrichment of 50% or more without violating safeguards.

  With the reprocessing system 1 according to the above embodiment, nuclear fuel having a plutonium enrichment of 50% or more can be produced while maintaining the proliferation resistance. For example, the performance of a plutonium-burning furnace using a high temperature gas furnace is Greatly improved. There exists a furnace type called Deep Burn as such a plutonium-only firing furnace concept, and the use of isolated plutonium is assumed in this furnace type. As an outline of the design specifications of this reactor type, the plutonium loading per batch is about 300 kg, and the removal burnup is about 550 GWd / t. About 97% of loaded Pu-239 can be eliminated.

FIG. 9 is an example of a graph comparing the criticality due to the difference in fuel used in the Deep Burn core concept. The analysis method uses the integral neutron transport equation solved by the collision probability method. Deep Burn · US shown in the graph of FIG.
Similar to Burn, the criticality is shown when isolated plutonium is used. Deep Burn · JP shown in the graph of FIG. 9 shows the criticality when MOX fuel with the maximum plutonium enrichment of 50% obtained by the current reprocessing in Japan is used. Deep Burn · IMF shown in the graph of FIG. 9 shows the criticality of the fuel manufactured by the reprocessing system 1 according to the above embodiment. Instead of mixing uranium, yttrium stabilized zirconia, which is an inert base material, is used. It is mixed. The plutonium inventory itself at this time is the same as Deep Burn · JP. A comparison between Deep Burn · US and Deep Burn · JP confirms that the infinite multiplication factor of Deep Burn · JP is significantly reduced. This is because the neutron capture reaction by U-238 is increased by mixing uranium, and the criticality is greatly impaired. On the other hand, Deep Burn · IMF shows a high infinite multiplication factor. This is because uranium is not mixed, so that the above-described decrease in criticality is avoided. In this evaluation, Deep Burn · IMF has about half the plutonium inventory as compared to Deep Burn · US, as does Deep Burn · JP. Still, Deep Burn • IMF has a higher infinite multiplication factor than Deep Burn • US because it is a system that lacks neutron deceleration, and it is possible to obtain an equivalent infinite multiplication factor by design support. it can. However, the decrease in criticality due to the neutron capture reaction of U-238 shown in Deep Burn · JP is inevitable. Actually, since it is a multi-batch core, direct comparison is not possible. However, when comparing values at which the multiplication factor is 1 in FIG. 9, the ultimate burnup of Deep Burn • JP is estimated to be less than half that of Deep Burn • US. Is done.

  In conclusion, from the viewpoint of criticality, it can be said that the burnup of the Deep Burn core concept is reduced to less than half by using the current MOX fuel with a plutonium enrichment of 50% by uranium mixing.

On the other hand, the extinction rate of Pu-239 is as follows. FIG. 10 is an example of a graph comparing the disappearance rates of Pu-239. About Deep Burn * US and Deep Burn * IMF, the burn-up burnup 550 GWd / t, which is the design specification of Deep Burn, can be achieved by the above-mentioned comparison regarding criticality. On the other hand, Deep Burn · JP has a plutonium enrichment of 50%, and the plutonium inventory is deep.
Since it is about half that of the Burn core concept, the achievable burnup is 275 GWd / t.
It has become. However, in view of the decrease in criticality due to the addition of uranium, it seems that it is actually difficult to achieve a burnup of 275 GWd / t. Under this condition, when the extinction rate of Pu-239 is compared, Deep Burn • JP has an extinction rate of Pu-239 of about 80%, which is about 15% lower than that of the Deep Burn core concept. Although the burn-up is about half, the plutonium inventory is also about half, and if plutonium is preferentially burning, the extinction rate of Pu-239 should not change for Deep Burn · JP. However, the Pu-239 annihilation rate is reduced due to Pu-239 generated by the neutron capture reaction of U-238. When plutonium is extinguished, mixing of uranium, the source of plutonium, should be avoided. Moreover, when an inert base material is mixed instead of mixing uranium, a plutonium annihilation rate equivalent to the Deep Burn core concept can be expected.

  Thus, it can be seen that the criticality and plutonium annihilation rate of Deep Burn, which is the concept of a plutonium burning furnace, are impaired by the mixing of uranium by the current reprocessing. On the other hand, if isolated plutonium can be used, the core performance of Deep Burn can be achieved. Even if it is difficult to use isolated plutonium due to safeguards, if the reprocessing system 1 uses a fuel mixed with an inert base material instead of mixing uranium, it is almost equivalent to Deep Burn. It can be seen that the core performance can be achieved.

  Deep Burn also has an extinction reactor concept that uses an accelerator for the complete extinction of actinide nuclides. In that case, in the current reprocessing, mixing uranium equivalent to plutonium is very inefficient, and the reprocessing system 1 can achieve more efficient extinction.

<Modification>
For example, the following process may be applied to the above-described fuel manufacturing process (S112). FIG. 11 is an example of a flowchart for producing fuel by the sol-gel method (internal gelation method). The reprocessing system 1 mixes an ammonia donor or the like with the U / Pu solution mixed with the inert base material to adjust the U / Pu solution (S201).

  The reprocessing system 1 drops the inside of a droplet by dropping a U / Pu solution to which an inert base material and an ammonia donor are added into oil such as paraffin oil or silicon oil at a high temperature (for example, about 362 to 363 K). Gelling by heating with oil (S302).

  The reprocessing system 1 wash | cleans the gel ball which gelatinized into the spherical shape with a washing | cleaning liquid (S303), and dries the gel ball in a wet state by washing | cleaning (S304). The gel balls may be dried by any method. For example, the gel balls can be forcibly dried by applying hot air (for example, air having a temperature of about 333 to 343 K).

  The reprocessing system 1 heats the dried gel balls with high-temperature hot air (for example, air having a temperature of about 573 to 673 K) (S305).

  The reprocessing system 1 sinters the heated gel balls (S306). The gel balls may be sintered by any method. For example, the atmosphere in which the gel balls are arranged is set to a high temperature (for example, about 1623 K), and the properties of the sintered body (porosity, etc.) are desired. The high-temperature air in the atmosphere may be depressurized, or the place where the gel balls are arranged may be an atmosphere of hydrogen, argon, or the like. Sintered gel balls constitute fuel nuclei.

  The reprocessing system 1 can also complete the fuel nucleus using the internal gelation method. The inert base material mixed in the above-described step (S110) remains in the fuel core without disappearing even after subsequent processing such as sintering. Therefore, the presence of the inert base material in the fuel core suppresses the decrease in the critical amount due to the increase in the density of plutonium, and the plutonium oxide exists alone because the process of denitrating the plutonium nitrate solution is not performed. As a result, the decrease in the resistance to proliferation is suppressed.

DESCRIPTION OF SYMBOLS 1,101 ... Reprocessing system 2 ... Fuel assembly 3 ... Storage equipment 4 ... Shear equipment 5 ... Melting equipment 6 ... FP separation equipment 7 ... Uranium / plutonium separation equipment 8 ... Uranium refining equipment 9 ... Plutonium refining equipment 10 ... Uranium denitration equipment 11 ... Uranium storage equipment 12 ... Plutonium adjustment / manufacturing equipment 13 ... Fuel storage equipment

Claims (10)

A reprocessing system for spent nuclear fuel removed from a nuclear reactor,
Separation means for separating uranium from the nitric acid solution in which the spent nuclear fuel is dissolved to obtain a plutonium nitrate solution;
Mixing means for mixing an inert base material with the plutonium nitrate solution sent from the separation means to the means for producing nuclear fuel by the sol-gel method,
Spent nuclear fuel reprocessing system. The mixing means mixes the inert matrix before evaporative concentration of the plutonium nitrate solution sent from the separating means to the manufacturing means;
The spent nuclear fuel reprocessing system according to claim 1. The mixing means mixes the inert base material with the plutonium nitrate solution at a mixing ratio at which the plutonium nitrate solution sent to the manufacturing means has a prescribed diffusion resistance.
The spent nuclear fuel reprocessing system according to claim 1 or 2. The mixing means further mixes the uranium nitrate solution in which the uranium separated in the separation means is dissolved into the plutonium nitrate solution at a mixing ratio at which the nuclear fuel produced by the production means has a desired plutonium enrichment. ,
The spent nuclear fuel reprocessing system according to any one of claims 1 to 3. The inert base material is a chemically and nuclearly inert material,
The spent nuclear fuel reprocessing system according to any one of claims 1 to 4. A method for reprocessing spent nuclear fuel removed from a nuclear reactor,
A separation step of separating uranium from the nitric acid solution in which the spent nuclear fuel is dissolved to obtain a plutonium nitrate solution;
A mixing step of mixing an inert base material with the plutonium nitrate solution sent from the separation step to a nuclear fuel production step by a sol-gel method,
Reprocessing method of spent nuclear fuel. The mixing step mixes the inert matrix before evaporating and concentrating the plutonium nitrate solution sent from the separation step to the manufacturing step.
The method for reprocessing spent nuclear fuel according to claim 6. In the mixing step, the inert base material is mixed with the plutonium nitrate solution at a mixing ratio at which the plutonium nitrate solution sent to the manufacturing step has a prescribed diffusion resistance.
The method for reprocessing spent nuclear fuel according to claim 6 or 7. In the mixing step, the uranium nitrate solution in which the uranium separated in the separation step is dissolved is further mixed into the plutonium nitrate solution at a mixing ratio at which the nuclear fuel produced in the production step has a desired plutonium enrichment. ,
The method for reprocessing spent nuclear fuel according to any one of claims 6 to 8. The inert base material is a chemically and nuclearly inert material,
The method for reprocessing spent nuclear fuel according to any one of claims 6 to 9.

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