JP2720602B2 - Reprocessing of spent nuclear fuel - Google Patents

Description

Description: FIELD OF THE INVENTION The present invention relates to a method for reprocessing spent nuclear fuel using the Purex method.

"Prior art" An example of a conventional method for reprocessing spent nuclear fuel using the Purex method is shown in FIG.

In this conventional method, first, a dissolved solution of spent nuclear fuel is adjusted to a nitric acid concentration of about 3N, an undissolved residue is separated in a clarification step, and a filtrate is sent to a first cycle of separation. In the first cycle of separation, U and Pu are simultaneously extracted using an organic solvent such as TBP, and then Pu and U are sequentially back-extracted into an aqueous phase to partition U and Pu.

Then, for each of U and Pu, purification by solvent extraction and back extraction was repeated twice, respectively, to obtain U product and
Pu products are getting.

[Problem to be Solved by the Invention] As described above, in the conventional method for reprocessing spent nuclear fuel, solvent extraction and purification are performed in all steps, but the following solvent is used. There are fundamental problems and improvements.

In the extraction (co-decontamination) step, since the impurity concentration in the filtered and clarified solution is high, the third
Phases are likely to occur, making operation management difficult. In addition, when the third phase is generated, the degree of separation between U and Pu and the fission products is deteriorated. When the third phase is further accumulated, the extraction operation cannot be performed, and the third phase is removed. And other maintenance operations are required, leading to a reduction in the plant operation rate.

In the co-decontamination / distribution process, since the solution having a high radiation dose comes into contact with the organic solvent, the organic solvent is significantly deteriorated, the generation of the third phase is increased, and a waste solvent which cannot be used as the extraction solvent is generated. Further, equipment for waste solvent treatment is required.

In any of the steps, it is necessary to perform a washing operation on the extracted and back-extracted solvent in order to increase the decontamination coefficient, and a washing waste liquid is generated. For this reason, a large amount of waste is generated in the reprocessing process, and a cleaning liquid treatment facility is required.

The extraction operation involves repeated extraction, washing, and back-extraction, which complicates the process and requires equipment for treating washing waste liquid and waste solvent. Therefore, a compartment (cell) for disposing these devices is required. Are large in size and number, and the cost required for operating the exhaust equipment in the cell is high. Therefore,
In order to reduce costs, a purification method that requires only compact equipment is desired.

More recently, the use of nuclear fuel for longer periods,
So-called high burn-up plans are being considered, and the use of high burn-up fuels in light water reactors, FBR fuels, MOX fuels, etc. is considered, but the spent nuclear fuel resulting from these will contain fission products (FP). The amount increases several times compared to the conventional case. Therefore,
In particular, the deterioration of the organic solvent used for extraction in the first cycle of separation is accelerated, the amount of radioactive waste solvent is further increased, and the degree of separation of fission products (FP) is deteriorated. doing.

The present invention has been made in view of the above circumstances, and it is an object of the present invention to provide a method for reprocessing spent nuclear fuel that can reduce the amount of solvent used in the entire process and can reduce the size of equipment.

[Means for Solving the Problems] Hereinafter, a method for reprocessing spent nuclear fuel according to the present invention will be specifically described.

FIG. 1 is a schematic process diagram showing an example of a reprocessing method according to the present invention. In this example, first, the spent nuclear fuel is dissolved with nitric acid in the same manner as in the conventional method, and the solution is once clarified to remove insoluble residues.

Next, in a liquid preparation step (not shown), a reducing agent is added to the dissolved liquid from which insoluble residues have been removed, and Pu ⁶⁺ in the solution is converted to Pu ^4+.
To be reduced to The properties required for this reducing agent are that it selectively reacts with Pu ^{6+ over} U ^{6+ in the} lysate to convert it to Pu ⁴⁺ , and the reducing agent itself does not interfere with the subsequent purification process. In addition, it does not remain in the purified product.

U to selectively react with Pu ⁶⁺ than is ^6+, oxidation-reduction potential of the reducing agent Pu ⁶⁺ / Pu ⁴⁺ oxidation-reduction potential 1.04V lower it essential than the further the redox potential Is desirably higher than the redox potential of U ⁶⁺ / U ⁴⁺ 0.33V.

A particularly preferable substance as a reducing agent satisfying these conditions is U ⁴⁺ , but there is a possibility that hydrazine, hydroxylamine nitrate, iron sulfamate and the like can be used in addition.

When U ⁴⁺ is used as a reducing agent, U is a substance originally present in the solution, and therefore has no possibility of interfering with the subsequent purification treatment. In order to add U ⁴⁺ to the solution, a method of adding uranyl nitrate solution obtained by electrolytic reduction of the uranyl nitrate solution and hydrazine as a stabilizer to the solution is adopted.

The preferable addition amount of U ⁴⁺ is about 1 to 3 times, more preferably 1 to 3 times the number of moles calculated from the content of Pu ^{6+ in} the solution.
~ 2 times. If it is less than 1, Pu ⁶⁺ may not be completely reduced. On the other hand, if it is more than three times, the residual amount of U ⁴⁺ will increase, and this residual U ⁴⁺ will remain in the mother liquor without being precipitated, which is not preferable because it increases the load in the subsequent solvent extraction step.

In the solution preparation step, further, the nitric acid concentration in the solution is adjusted to 3 to 6 in order to increase the precipitation efficiency in the subsequent crystal purification step 1.
N, preferably 4-6N, and the uranium concentration is adjusted to 200-500g /, preferably 350-400g /.
This preparation may be performed before or after the addition of the reducing agent, or simultaneously. If the nitric acid concentration after preparation is less than 3N, the uranium concentration in the mother liquor at the freezing point of the solution is high, and a sufficient uranium recovery cannot be obtained. Conversely, if the nitric acid concentration is higher than 6N, it becomes difficult to handle the solution. On the other hand, if the uranium concentration is less than 200 g /, a sufficient yield cannot be obtained. If the uranium concentration is more than 500 g /, the amount of generated crystals is large, solid-liquid separation becomes difficult, and the decontamination coefficient is deteriorated.

Next, the prepared solution is transferred to a precipitation tank in the crystal purification step 1, and the solution is cooled to perform crystallization. As a cooling method, the entire precipitation tank may be cooled. However, for more efficient operation, a method of immersing a cooling medium in a solution is adopted. Further, a cooling method can also be adopted in which a refrigerant such as dry ice, liquid nitrogen, or chlorofluorocarbon is directly introduced into the liquid.

As the cooling medium, for example, a bundle of a number of U-shaped tubes made of a stainless steel pipe or the like is used, and a cooling medium such as ethylene glycol is circulated through a cooling liquid passage formed therein. The shape of the cooling medium is not limited to a tubular body such as a U-shaped tube, and may be appropriately changed in shape, such as a flat plate, a prism, or a cylinder, as needed.

In this way, the entire solution is cooled to −10 ° C. to −30 ° C.
Preferably, it is cooled to -10 to -20C. Then, U ⁶⁺ in the solution is precipitated as uranyl nitrate, but the nitrate of Pu ⁴⁺ is not contained in the precipitate because it is more difficult to precipitate than uranyl nitrate.

If the final temperature in the crystal refining step 1 is lower than −30 ° C., water or / and nitric acid trihydrate in the solution may precipitate together with the crystals, and impurities may be taken into the crystals to lower the purity. . The temperature of the solution at the start of cooling is 20
3535 ° C. is desirable, and it is cooled to the final temperature in 60-180 minutes, more preferably 120-180 minutes. The cooling rate in the cooling process is set to 1 ° C./min or less, more preferably 0.5 ° C./min or less, so that impurities are hardly taken into the precipitated crystals.

When the crystallization is completed, the precipitate is recovered. When the above-mentioned cooling medium is used, uranyl nitrate precipitates as dendritic crystals on the surface of the cooling medium, so that it is easy to recover and insoluble residues remaining in the solution are removed from the precipitation tank. Convenient because it precipitates at the bottom and is removed from the mother liquor.

Next, the mother liquor remaining in the precipitation tank is filtered as required, and then adjusted to the same nitric acid concentration as in the conventional solvent extraction, or sent to the extraction step 2 as it is. In this extraction step 2, as in the conventional method, U, P
After extracting u, the organic phase is sufficiently washed with an aqueous phase (aqueous nitric acid solution) using a washing tower. The extraction residue generated in the extraction step 2 is sent to a high-level waste liquid treatment step.

Next, the washed organic phase is sent to the partitioning step 3, in which plutonium nitrate is back-extracted into the aqueous phase and then back-extracted again under different conditions, and U remaining in the organic phase is back-extracted into the aqueous phase as uranyl nitrate. Then, U and Pu are separated.

On the other hand, uranyl nitrate recovered in the crystal purification step 1 is
It is sent to the U purification step 4 together with the uranyl nitrate solution obtained in the distribution step 3. In this U purification step 4, U is extracted from the uranyl nitrate solution into an organic phase, the organic phase is washed with an aqueous nitric acid solution, and uranyl nitrate as a product is back-extracted.

On the other hand, plutonium nitrate obtained in the distribution step 3 is Pu
Send to the first purification step 5 and the second Pu purification step 6 to extract
The back extraction is repeated twice to convert to plutonium nitrate product.

According to the spent nuclear fuel reprocessing method having the above-described configuration, most (about 95%) of U in the spent fuel solution is recovered as uranyl nitrate crystals in the crystal refining step 1. The total amount of U and Pu to be sent can be greatly reduced, and the amount of solvent used in the extraction step 2 and the distribution step 3 can be reduced to 1/10 to 1/20 of the conventional method. When compared at the same scale of throughput, the equipment required for crystal purification can be downsized compared to the equipment required for solvent extraction, so the above method can also downsize the equipment as a whole process. .

Further, since the uranyl nitrate crystal recovered in the crystal purification step 1 is of high purity, it can already meet the product specifications at the time of passing through only one U-purification step 4, which has been conventionally required. The second U purification step can be omitted. Therefore, the amount of solvent used after the distribution step 3 is also about 1 /
2, the amount of solvent used can be reduced from this point, and the amount of waste related to the solvent can also be reduced.

Next, FIG. 2 shows an example in which the extraction step 2 to the distribution step 3 of the above method are changed. In this example, extraction step 2
Back-extraction step to increase the degree of purification between
10 and an extraction step 11 are provided.

According to this method, the degree of purification in the extraction step 2 to the distribution step 3 can be increased.
The Pu purification step 5 which was required in one step can be completed in one step. At the same time, the final purity of U is also improved.

Next, FIG. 3 shows a third embodiment of the present invention.
U purification step 4 using solvent extraction and back extraction in the first example
, A crystal refining step 20 similar to the crystal refining step 1 is provided. The purification conditions in this crystal purification step 20 may be the same as in the first crystal purification step 1.

According to this method, since the U purification step 4 is eliminated, the first
The amount of solvent used can be further reduced as compared with the method of the example.

A model when the process of FIG. 1 is applied to ordinary fuel processing is shown below. This is because the U amount is 680kg /
It is assumed that a spent nuclear fuel solution with a Pu amount of 8.0 kg / day is treated.

Reducing agent type: U ⁴⁺ Concentration after preparation: 400 gU / 4.5 g Pu / Nitric acid concentration: 3 N Reductant addition amount: 4.0 gU / U recovery rate in crystal purification step 1: 95% Recovery amount in crystal purification step 1 U: 645 kg / day U remaining amount in mother liquor: 35 kg / day Pu remaining amount: 8.0 kg / day Solvent usage in extraction step 2: 20 / hr (300 /
hr) Solvent usage in distribution step 3: 22 / hr (340 / h in conventional method)
r) As described above, according to the process of FIG. 1, the amount of the solvent that comes into contact with the solution having a high radiation concentration can be reduced as compared with the conventional method, and the amount of waste related to the solvent can be reduced.

The main feature of the present invention is to provide a crystal refining step 1 for separating U from spent nuclear fuel,
Regarding the processing conditions and processes after the distribution step 3, the configuration may be appropriately changed as necessary.

"Example" Uranyl nitrate 400gU /, Neptunium nitrate 10μCi
/ (About 14ppm), other fission products (Cs, Sr, Y, Zr, P
d, Fe, Ce, Eu, and Ru) were prepared, each containing several g / nitrate and having a nitric acid concentration of 6N.

Neptunium nitrate is an alternative to plutonium nitrate. Neptunium and plutonium have very similar chemical properties, both have a valence of +3 to +6 in aqueous solution, and both tend to be lower valence and more stable than uranium. The valence of neptunium in the simulated liquid is 5
The valence of uranium is hexavalent, and uranium is hexavalent.

First, 0.04 gU / U4 ⁺ and 0 hydrazine were added to the above simulated liquid.
Each was added to be 1 mol /. This addition amount is twice the equivalent amount capable of reducing the entire amount of neptunium to Np ⁴⁺ . At the time of addition, the simulated liquid temperature was 33 ° C.

Next, the simulation liquid was transferred into the precipitation tank, and a cooling pipe through which a cooling medium was passed was immersed in the simulation liquid, and the entire solution was cooled to −20 ° C. over 180 minutes. Then, uranyl nitrate was deposited on the cooling tube, and the uranium concentration in the liquid became about 20 gU /. At this time, the uranium recovery in the crystallization step was 95%.

Further, the content of fission products in the recovered uranyl nitrate crystals was measured, and the separation coefficient was calculated. The result is
It is shown in the table.

[Effects of the Invention] As described above, according to the method for reprocessing spent nuclear fuel according to the present invention, the following excellent effects can be obtained.

The crystal purification process using the crystallization method recovers most (about 95%) of U in the spent fuel solution as uranyl nitrate crystals, so the total amount of U and Pu sent to the extraction process is greatly reduced. it can. Therefore, the amount of solvent used in the extraction step and the distribution step can be reduced to 1/10 to 1/20, and at the same time, the amount of waste solvent is reduced.

When compared at the same throughput, the equipment required for crystal refining can be downsized compared to the equipment required for refrigerant extraction, so that the present invention allows the equipment as a whole process to be downsized. .

Since the uranyl nitrate supplied from the crystal purification process is of high purity, the subsequent uranium purification process can be performed in a single step to meet product specifications, eliminating the need for the second uranium purification process that was conventionally required. And the amount of solvent used in the entire process can be reduced.

In the solvent extraction treatment of the mother liquor after crystallization purification, U,
From the viewpoints of Pu content, throughput, etc., it is possible to use the reprocessing equipment of the fast breeder reactor (FBR) as it is.

[Brief description of the drawings]

1 to 3 are process diagrams schematically showing different examples of a method for reprocessing spent nuclear fuel according to the present invention, and FIG. 4 is a process diagram of a conventional method for reprocessing spent nuclear fuel.

Continuation of the front page (72) Inventor Akira Tanaka 1002--14, Mukaiyama character, Nakamachi, Naka-gun, Naka-gun, Ibaraki Pref. A) Tokiko 39-9950 (JP, B1)

Claims (2)

(57) [Claims] 1. A reducing agent is added to a solution of spent nuclear fuel to reduce Pu ⁶⁺ in the solution to Pu ⁴⁺ , and the solution is cooled to precipitate uranyl nitrate while remaining The mother liquor is subjected to a solvent extraction treatment using the Purex method to separate and purify uranyl nitrate and plutonium nitrate, respectively, and further to purify uranium by combining the precipitate and uranyl nitrate obtained by the solvent extraction treatment. A method for reprocessing spent nuclear fuel, characterized in that: 2. The oxidation-reduction potential of Pu ⁶⁺ / Pu ^{4+ as} the reducing agent.
The method for reprocessing spent nuclear fuel according to claim 1, wherein the substance has a redox potential lower than 1.04V.