JPH11287890A - Reprocessing method for spent nuclear fuel - Google Patents

Abstract

PROBLEM TO BE SOLVED: To reprocess spent fast reactor fuel at low cost, by recovering plutonium and uranium from fast reactor fuel in a reprocessing facility for fast reactor and from light water reactor fuel and recovered residual liquid of facility for fast reactor. SOLUTION: Sheared pieces of fast reactor accepted in a reprocessing facility are heated and solved in nitric acid as solution, which is controlled at specific concentration of uranium, plutonium and nitric acid and sent to a multiple step extraction device. The solution is flown from the first step to the final step of the multiple step extractor. Also, mixture solvent of a specific composition is flown from the final step to the first step at a specific flow ratio. By this, the mixture in mixed oxide(MOX) fuel composition can be separated. The solution liquid from the final step of the multiple step extractor is sent to a reprocessing facility for light water reactor. On the other hand, mixed solvent which uranium and plutonium are extracted from is contacted in counter- current with diluted nitric acid water solution containing reducer by using the multiple step extractor. By this, the mixture of totally stripped plutonium and uranium is taken out as diluted nitric acid water solution, denitrated and converted into oxide to be MOX fuel.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for reprocessing spent nuclear fuel generated from a nuclear power plant.

[0002]

2. Description of the Related Art Uranium and plutonium in spent nuclear fuel generated from a nuclear power plant are separated and recovered by a chemical separation process called the Purex method. The Purex method is based on Nuclear Chemical by M. Benedict and two others.
Engineering, 2nd edition, pages 466 to 514, published by McGraw-Hill.

In the Purex method for light water reactor fuel, spent nuclear fuel is dissolved in nitric acid, and a solvent in which 30% of tributyl phosphoric acid (TBP) is dissolved in normal dodecane in a co-decontamination process is used. The total amount of uranium and plutonium is extracted and separated from fission product nuclides. In the co-decontamination step, countercurrent extraction is performed by combining extractors such as pulse columns in multiple stages. The countercurrent extractor includes an extraction unit and a scrub unit.

[0004] In the extraction unit, uranium and plutonium are extracted from the solution. However, the current extraction condition of the Purex method is a condition for extracting uranium preferentially over plutonium. In addition, the flow rate ratio of TBP to the dissolving liquid has a margin. At this time, since fission product nuclides are extracted by surplus TBP, the TBP is washed with a nitric acid aqueous solution in the scrub portion. The raffinate and the scrub liquid are mixed and sent to a waste liquid treatment system.

On the other hand, the TBP from which uranium and plutonium have been extracted is sent to a distribution step, where the plutonium is back-extracted while reducing plutonium with dilute nitric acid containing a reducing agent. I do. Uranium and plutonium are each sent to a purification process and purified.

[0006] Fast reactor fuels do not require as high a product purity as light water reactors. In the Purex method for fast reactor fuel, the core fuel and blanket fuel are dissolved with nitric acid,
Up to the step of extracting and distributing both uranium and plutonium together in the co-decontamination step, the flow is the same as above,
The purification step can be omitted.

[0007] Plutonium recovered by the Purex process is used together with uranium as a mixed oxide fuel (MOX fuel). For this reason, plutonium recovered from spent light water reactor fuel and fast reactor fuel is mixed with uranium and plutonium again so as to have the composition of MOX fuel for light water reactor and fast reactor, respectively.

[0008]

SUMMARY OF THE INVENTION It is an object of the present invention to provide a fast reactor fuel reprocessing facility having a small facility and economical load. Therefore, it is an issue to reduce the heavy metal processing amount in the fast reactor reprocessing facility. When reprocessing the fast reactor fuel, the plutonium / uranium ratio in the fuel is about 0.1. When this is increased to about 0.3 as MOX fuel, about 70% of uranium becomes surplus. By appropriately treating the surplus uranium, the amount of heavy metal to be treated can be reduced. The present invention has solved this problem by the following means.

[0009]

A first feature of the reprocessing method of the present invention resides in that a reprocessing facility for a fast reactor and a reprocessing facility for a light water reactor are used together. The fast reactor reprocessing facility receives the core fuel and blanket fuel used in the fast reactor, and collects only the mixed product of plutonium and uranium of the MOX fuel composition of the fast reactor. The light water reactor reprocessing facility receives the fuel used in the light water reactor and the residual liquid collected from the fast reactor reprocessing facility, and recovers a mixed product of the MOX fuel composition for the light water reactor and uranium.

[0010] A second feature of the reprocessing method of the present invention is that an existing reprocessing facility is used as a reprocessing facility for a light water reactor. That is, part of the uranium in the spent fast reactor fuel is transferred to a reprocessing facility that is already operating for light water reactors and processed.

A third feature of the present invention is that, in a reprocessing facility for a fast reactor, a mixed solvent of tributylphosphoric acid (TBP) and trinormal octylamine (TOA) is used.
The purpose of the present invention is to extract and separate the entire amount of plutonium from a fuel solution together with uranium in such a ratio that the composition becomes equivalent to that of the mixed oxide fuel. At this time, the composition of the mixed solvent is the composition described in claim 4.

That is, the principle by which the present invention can provide a reprocessing facility with small equipment and economical load is as follows.

[0013] The fuel used in the fast reactor cannot be accepted in a light water reactor reprocessing facility from the viewpoint of criticality control because of the large amount of plutonium. From the viewpoint of safety, two reprocessing facilities for the fast reactor and the light water reactor are required. However, if the entire amount of plutonium is recovered from spent fast reactor fuel, the load of criticality management is reduced, so that it can be accepted into a light water reactor reprocessing facility. Therefore, in the reprocessing facility for the fast reactor, only the mixed product of the MOX fuel composition used in the fast reactor is recovered, and the surplus uranium is treated together with the spent light water reactor fuel in the reprocessing facility for the light water reactor.

At this time, by using the existing reprocessing facility, the newly required reprocessing facility for the fast reactor is:
Only the amount of extraction solvent needed to extract all of the plutonium and part of the uranium is needed. Since uranium in the fuel solution contains about 10 times as much as plutonium, the amount of extraction solvent is small when recovering only part of uranium as compared with the case where both uranium and plutonium are extracted in total. I'm done.

As a result, the load on the solvent regeneration system and the waste solvent treatment system is reduced, and the amount of waste generated due to the waste solvent can be reduced. When plutonium and uranium are extracted to have a MOX fuel composition, there is no need to separate them by back extraction. For this reason, since the distribution step is only a simple back-extraction operation, the amount of back-extraction liquid can be reduced. As a result, the burden on the product concentration system is small. As described above, the equipment and economic load of the reprocessing facility for the fast reactor is reduced.
On the other hand, as for the reprocessing facility for light water reactors, uranium is received from the above facility, so that the throughput increases, but the ratio of plutonium decreases, so that criticality management becomes easier.

In the reprocessing facility for a fast reactor, the total amount of plutonium and a part of uranium can be recovered by a solvent extraction method using a mixed solvent of TBP and TOA. This is based on the results of independent research conducted by the inventors. This can be achieved by mixing 30% of TBP with 3% or more of TOA. However, when the TOA concentration is less than 7%, TBP and TOA show a synergistic effect, and the extraction behavior becomes complicated. Therefore, it is desirable that the TOA concentration be 7% or more. In that case, a third phase is formed by extracting a high concentration of uranium.

This can be solved by adding isooctyl alcohol. For example, when the TOA concentration is 10%, 8% isooctyl alcohol is required. TO
When the A concentration increases, the isooctyl alcohol concentration required for the suppression of the third phase also increases, but the distribution coefficient of plutonium decreases accordingly, and when it exceeds 17%, the distribution coefficient of plutonium becomes smaller than that of uranium. .

From these, it is desirable that the isooctyl alcohol concentration is about 10%. At this time, the maximum concentration of TOA in which the third phase is not formed is 15%.

From the above, the mixed solvent according to the present invention is 3
0% TBP, 3% or more, preferably 7 to 15% TOA, and 8 to 17%, preferably about 10% isooctyl alcohol diluted with normal dodecane. Using this, plutonium and uranium can be MO
Separate to the composition of X fuel.

[0020]

(Embodiment 1) A first embodiment according to the present invention will be described with reference to FIG. In this embodiment, 1
Using a mixed solvent of 5% TOA, 30% TBP, and 10% isooctyl alcohol diluted with normal dodecane, the total amount of plutonium is extracted from the fuel solution together with uranium in such a ratio as to make the composition of MOX fuel. I do. Thereafter, excess uranium is extracted using a mixed solvent having the same composition as above. Hereinafter, the process will be described in detail by taking as an example the reprocessing of spent fuel in which uranium is present in an amount about 10 times that of plutonium.

[0021] The fast reactor fuel is received in a reprocessing facility, stored and cooled for a period of time. Thereafter, in the shearing step 1,
It is mechanically sheared using a shearing machine so as to be easily dissolved in the subsequent dissolving step 3. The sheared fragments are transferred to a dissolving tank and dissolved by heating with a specified amount of nitric acid. In the fining step 3, the concentration of uranium and plutonium is analyzed in a measuring / adjusting tank after removing insoluble residues from the lysate using, for example, a centrifugal fining device. The concentration of L and nitric acid is adjusted to 3N and sent to the intermediate storage tank. In the above steps, a known technique used in the conventional Purex method can be applied.

The dissolving solution is supplied from the intermediate storage tank to plutonium-
It is sent to the uranium co-extraction step. Here, an extraction device in which extractors such as a mixer setra or a centrifugal extractor are assembled in multiple stages is used. The lysis solution flows from the first stage to the last stage of the multistage extractor, and the mixed solvent flows from the last stage to the first stage. At this time, the total amount of plutonium and uranium in an amount three times the amount of plutonium are extracted, and the flow ratio of the mixed solvent and the solution is adjusted so that about 90% or more of both TBP and TOA in the mixed solvent are combined with uranium or plutonium. decide.

As a result, the mixture of the MOX fuel composition can be separated from the fission product nuclides while being excellently decontaminated. The lysate from the last stage of the multistage extractor is sent to the existing light water reactor reprocessing facility. On the other hand, the mixed solvent obtained by extracting uranium and plutonium is sent to a plutonium-uranium back extraction step without scrubbing.

In the plutonium-uranium back extraction step, the mixed solvent is brought into countercurrent contact with a dilute aqueous nitric acid solution containing a reducing agent such as divalent iron ion using the same multistage extractor as described above. Thereby, the whole amount is back-extracted while reducing plutonium from tetravalent to trivalent, and at the same time, the whole amount of uranium is also back-extracted. In this way, the mixture of plutonium and uranium is taken out as a dilute aqueous nitric acid solution, denitrated by irradiating microwaves or the like, converted into oxides, and used as MOX fuel. The mixed solvent after back-extraction is sent to a solvent washing system, where it is brought into contact with an aqueous solution of alkali carbonate or the like to wash and remove degraded products and extractable fission product nuclides such as technetium and platinum group elements, and to adjust the mixed solvent composition Analyze and adjust and reuse.

[0025]

As described above, according to this embodiment, inexpensive spent fast reactor fuel reprocessing can be performed.

[Brief description of the drawings]

FIG. 1 is a flowchart showing a separation flow of a fast reactor fuel reprocessing method according to the present invention.

[Explanation of symbols]

 1 ... shearing step, 2 ... dissolving step, 3 ... fining step.

Claims (4)

[Claims] In a reprocessing method for spent nuclear fuel, a first reprocessing facility for recovering a total amount of plutonium and a part of uranium from a fuel solution of a fast reactor so as to have a core fuel composition of the fast reactor. And a second reprocessing facility for recovering uranium and plutonium from a residual solution collected in the reprocessing facility and a light water reactor fuel solution, and a method for reprocessing spent nuclear fuel. 2. The reprocessing method according to claim 1, wherein
A method for reprocessing spent nuclear fuel, wherein the second reprocessing facility is an existing light water reactor fuel reprocessing facility. 3. The reprocessing method according to claim 1, wherein
Is characterized in that the entire amount of plutonium and a part of uranium are recovered with the composition by a solvent extraction method using a mixed solvent obtained by mixing a tertiary amine-based extraction reagent and tributyl phosphoric acid. Reprocessing of spent nuclear fuel. 4. The reprocessing method according to claim 1, wherein
Dilute a mixture of 30% tributyl phosphoric acid, 3% or more, preferably 7 to 15% trinormal octylamine, and 8 to 17%, preferably about 10% isooctyl alcohol with normal dodecane. Reprocessing of spent nuclear fuel, characterized in that it is a mixed solvent.