JP3308045B2 - Technetium decontamination method in co-decontamination process of spent nuclear fuel reprocessing - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an improvement in a technique for reprocessing spent nuclear fuel generated from a nuclear power reactor or the like.

[0002]

2. Description of the Related Art In the reprocessing Purex process of spent nuclear fuel [extraction and recovery of U and Pu from spent nuclear fuel by tributyl phosphate (TBP) -nitric acid system], Tc is used.
(VII) was considered non-extractable and no special consideration was given to Tc. However, in recent years, Tc
In the co-decontamination step (VII), part or all of (VII) is extracted together with U and Pu, and then distributed (U and P).
separation from u).

In the distribution step, it has been known that Tc catalytically promotes various oxidation-reduction reactions and hinders the separation of U and Pu. Therefore, recently, in the UP-3 or Rokkasho reprocessing facility in France, in the co-decontamination step of the spent nuclear fuel reprocessing step, Tc is back-extracted from the organic phase containing extracted U, Pu and Tc. .

Since the aqueous back-extraction solution contains not only negligible amounts of U and Pu besides Tc, it is introduced into another auxiliary extraction step to extract and recover only U and Pu with an extraction solvent. The organic phase is recycled to the main extraction process. And T which was not extracted in the auxiliary extraction process
The aqueous phase containing c is combined with the high-level aqueous phase effluent.

[0005]

SUMMARY OF THE INVENTION Tc in the prior art
In the reverse extraction and removal method, an auxiliary extraction step is provided after Tc back extraction, in which U and Pu are recovered by an extraction solvent from an aqueous solution in which Tc is back-extracted.
There is a problem that the cost of the processing method c is large.

[0006]

In the present invention, Tc from the Tc back-extraction step placed after the U-Pu extraction and washing step (main extraction step) in the co-decontamination step of the reprocessing step of spent nuclear fuel. The aqueous back-extraction solution is directly combined with the extraction (washing) step installed before. By doing so, the U and Pu contained in the Tc back extraction aqueous solution are recovered in the extraction step of the main extraction step, and Tc is discharged to the high-level aqueous phase waste liquid stream. As a result, in the present invention, an auxiliary extraction step for separating U and Pu from Tc is not required, so that the processing steps are extremely simplified in terms of process.

[0007]

According to the present invention, in the tributyl phosphate (TBP) -nitric acid extraction system, which is the basis of the Purex method,
The extraction and distribution characteristics of Tc (VII) and the co-extraction effect of U and Zr are experimentally determined, the data and other public data are processed by a computer, and the distribution ratio calculation formula of Tc (VII) is derived and implemented. did.

Next, this calculation formula was introduced into the simulation code of the Purex process, and the behavior of each component such as Tc was analyzed with respect to various first cycle extraction process flow sheets. It was confirmed that the merging provided a high Tc decontamination (removal of Tc from U and Pu) coefficient for the U-Pu product.

[0009]

[Example] Purex process (30% TBP-nitric acid system)
Process calculation code: EXTRA, average extraction,
The flow sheet in which the Tc reverse extraction step (6 steps) was added to the washing step (10 steps) was analyzed. The flow sheet is shown in FIG. Here, assuming that all Tc is extracted in the extraction step, the concentration of Zr in the supply liquid is changed to the U concentration (250%).
g / l) of 0.4% (1.0 g / l).
The aqueous back-extraction solution from the Tc back-extraction step is injected into the fourth stage of the extraction step.

The last two stages (17 and 18) are for removing excess nitric acid in the organic phase. U, Pu, T
The spent nuclear fuel solution containing c and Zr is introduced into the extraction stage 6, and an extraction solvent for extracting U and Pu is extracted from the extraction stage 1.
Will be introduced. Both are in countercurrent contact between the extraction stage 1 to the extraction stage 6, the high-level aqueous effluent (AW) is discharged from the extraction stage 1, and the extraction solvent that has extracted Tc and Zr together with U, Pu, etc. is the extraction stage. The cleaning process is further performed while moving from 6 to the cleaning stage 10.

A low concentration cleaning solution is added in the cleaning stage 10. The washed organic solvent is introduced into a Tc back-extraction step attached to the extraction step, where it is counter-currently contacted with the high-concentration washing solution supplied to the washing stage 16 to back-extract Tc from the co-extraction solvent. The extraction solvent (OP) containing U and Pu is removed from the washing stage 18 after removing excess nitric acid by the washing stages 17-18, and the Tc-containing aqueous back-extraction solution is extracted in the washing stage 11 and extracted in the extraction stage 4 in the preceding step. Is returned to.

FIG. 2 shows the results of the concentration distribution of the main elements processed in the extraction and washing steps in the present invention. From this figure, it is understood that 99% or more of Tc is contained in the aqueous phase waste liquid in the first stage of the extraction step and flows out. Also, it is understood that the concentrations of U and Pu in the aqueous phase waste liquid are extremely low at 1 mg / l or less.

[0013]

According to the present invention, the aqueous back-extraction solution of the Tc back-extraction step installed after the main extraction and washing step of the first cycle of the Purex process can be directly supplied to the main extraction and washing step without adding a new extraction step. By injecting between the supply stage of the (spent nuclear fuel solution) and the supply stage of the extraction solvent, Tc can be quantitatively discharged into the high-level aqueous phase waste liquid.

[Brief description of the drawings]

FIG. 1 is a view showing a flow sheet for removing Tc (VII) in a first recycle of a spent nuclear fuel reprocessing step of the present invention. The numbers in the circles indicate the relative ratios of the flow rates.

FIG. 2 shows Tc (VII) in the flow sheet of FIG. 1,
It is a figure which shows the concentration distribution of U (VI) and Pu (IV).

[Explanation of symbols]

 1-10 Extraction washing step (or extraction and washing step) 11-16 Tc back extraction step (or washing step) 17-18 Nitric acid washing step (or washing step)

Continuation of the front page (56) References JP-A-63-198897 (JP, A) JP-A-7-12987 (JP, A) JP-A-1-78136 (JP, U) Shoichi Iimori, Purex Process First Examination of Tc decontamination flow sheet in cycle, Proceedings of the Annual Meeting of the Atomic Energy Society of Japan (Venues A to M), Japan, Atomic Energy Society of Japan, March 10, 1993, 1993 (31st) , P. 475 (58) Fields investigated (Int. Cl. ⁷ , DB name) G21C 19/46 G21F 9/06 581

Claims (1)

(57) [Claims] In the co-decontamination step of the spent nuclear fuel reprocessing Purex step, the co-extraction with zirconium (Zr), plutonium (Pu), and uranium (U) results in
In order to remove heptavalent technetium [Tc (VII)] extracted in the Pu organic product from the organic phase and discharge it into the high-level aqueous phase waste liquid, the Tc after the main extraction and washing steps in the co-decontamination step A back extraction process is installed, and the back extracted T
By injecting the c-containing aqueous solution between the supply stage of the spent nuclear fuel solution and the extraction solvent in the main extraction and washing steps, Tc is quantitatively discharged into the high-level aqueous phase waste liquid. (1) Tributyl phosphate extraction solvent is supplied to the main stage of the main extraction and washing step, and spent nuclear fuel solution is supplied to the main stage of the main extraction and washing step. To extract Zr, Pu, U and Tc into an extraction solvent by countercurrent contact with
The low-concentration nitric acid washing solution is supplied to the latter stage of the main extraction and washing step to wash the extraction solvent, Zr is back-extracted, and then supplied to the first stage of the Tc back-extraction step, and the Zr and Tc-containing high-level aqueous phase waste solution is mainly supplied. (2) A high-concentration nitric acid washing solution is supplied after the Tc back-extraction step, and the washing solution and the extraction solvent are brought into countercurrent contact with each other in the back-extraction step to bring Tc into the washing solution. Back-extracting, removing the U- and Pu-containing extraction solvent from the latter stage of the back-extraction step;
A method of extracting the remaining U and Pu by injecting between the supply stage of the spent nuclear fuel solution and the supply stage of the extraction solvent in the washing step and contacting with the extraction solvent.