JPH0712986A - Decontaminating method for technetium in co-decontamination step in reprocessing process of spent nuclear fuel - Google Patents

Abstract

PURPOSE:To simplify a processing step and reduce costs by uniting a Tc stripping solution from a Tc stripping step set after a main extracting/cleaning step in a co-decontamination step of a reprocessing process, at a cleaning step of the precedent process. CONSTITUTION:In a co-decontamination step of a reprocessing purex process of spent fuel, septivalent technetium Tc (VII) extracted in a U-Pu organic product through the co-extraction of Zr, Pu and U is removed from the organic phase and discharged in a high-level water-phase waste liquid. For this purpose, a Tc stripping step is arranged after a main extracting/cleaning step in the co-decontamination process. The aqueous solution containing the stripped Tc is injected between a feeding stage of a feed liquid and a feeding stage of an extraction solvent in the main extracting/cleaning step, so that the Tc is discharged into the high-level water-phase waste liquid quantitatively.

Description

Detailed Description of the Invention

[0001]

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

[0002]

2. Description of the Related Art In the Purex process for reprocessing spent nuclear fuel [extraction and recovery process of U and Pu from spent nuclear fuel by tributyl phosphate (TBP) -nitric acid system], Tc
(VIII) was considered to be non-extractable, and no special consideration was given to Tc. However, in recent years, Tc
(VII) is extracted together with U and Pu in the co-decontamination step of the above step together with U and Pu.
It became known that the separation from u) was reached.

It has also become known that in the partitioning process, Tc catalytically promotes various redox reactions and interferes with the separation of U and Pu. Therefore, recently, at the UP-3 or Rokkasho reprocessing facility in France, in the co-decontamination process of the spent nuclear fuel reprocessing process, Tc is back-extracted from the extracted organic phase containing U, Pu and Tc. .

Since the back-extracted aqueous solution contains non-negligible amounts of U and Pu other than Tc, it was 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 not extracted in the auxiliary extraction process
The aqueous phase containing c is combined with the high-level aqueous waste liquid.

[0005]

[Problems to be Solved by the Invention] Tc in the prior art
In the reverse extraction removal method of T.sub.c, 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.
The method of treating c has a problem that the cost is heavy.

[0006]

In the present invention, in the co-decontamination process of the spent nuclear fuel reprocessing process, the Tc from the Tc back extraction process placed after the U-Pu extraction and washing process (main extraction process). The back extraction aqueous solution is allowed to join the above-mentioned extraction (washing) step installed before. By doing so, 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, the auxiliary extraction step for separating U and Pu from Tc is not required, so the processing steps are extremely simplified in terms of process.

[0007]

In the present invention, in the tributyl phosphate (TBP) -nitric acid extraction system which is the basis of the Purex method,
The extraction distribution characteristic of Tc (VII) and the co-extraction effect by U and Zr were experimentally obtained, and the data and other public data were processed by a computer, and the distribution ratio calculation formula of Tc (VII) was derived. 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 for various first cycle extraction process flow sheets. It was confirmed that the U-Pu product obtained a high decontamination coefficient of Tc (removal of Tc from U and Pu) by merging.

[0009]

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

The last two stages (17, 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 the extraction solvent for extracting U and Pu is extracted from the extraction stage 1.
Will be introduced to. Both of them come into countercurrent contact between the extraction stage 1 and the extraction stage 6, the high-level aqueous phase waste liquid (AW) is discharged from the extraction stage 1, and the extraction solvent that also extracts Tc and Zr together with U, Pu, etc. Further cleaning processing is performed while shifting from 6 to the cleaning stage 10.

A low-concentration cleaning liquid is added in the cleaning stage 10. The washed organic solvent is introduced into the Tc back-extraction step attached to the extraction step, where it countercurrently contacts 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 the excess nitric acid is removed by the washing stages 17-18, and the Tc-containing back extraction aqueous solution is extracted in the washing stage 11 to extract the extraction stage 4 of the previous step. Returned to.

FIG. 2 shows the results of the concentration distribution of the main elements treated in the extraction and washing steps in the present invention. From this figure, it can be seen 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. It is also understood that the concentrations of U and Pu in this aqueous phase waste liquid are extremely low at 1 mg / l or less.

[0013]

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

[Brief description of drawings]

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

FIG. 2 shows Tc (VII) in the flow sheet of FIG.
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, washing step) 11-16 Tc Back extraction step (or washing step) 17-18 Nitric acid washing step (or washing step)

[Procedure amendment]

[Submission date] July 13, 1993

[Procedure Amendment 1]

[Document name to be corrected] Drawing

[Name of item to be corrected] Figure 2

[Correction method] Change

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[Procedure amendment]

[Submission date] January 12, 1994

[Procedure Amendment 1]

[Document name to be amended] Statement

[Name of item to be corrected] Claim 1

[Correction method] Change

[Correction content]

[Procedure Amendment 2]

[Document name to be amended] Statement

[Name of item to be corrected] 0002

[Correction method] Change

[Correction content]

[0002]

2. Description of the Related Art In the Purex process for reprocessing spent nuclear fuel [extraction and recovery process of U and Pu from spent nuclear fuel by tributyl phosphate (TBP) -nitric acid system], Tc
(VII) is considered to be non-extractable and no special consideration was given to Tc. However, in recent years, Tc
(VII) is extracted together with U and Pu in the co-decontamination step of the above step together with U and Pu.
It became known that the separation from u) was reached.

Claims (1)

[Claims] 1. In the co-decontamination process of the spent nuclear fuel reprocessing Purex process, by co-extraction with zirconium (Zr), plutonium (Pu) and uranium (U),
In order to remove the 7-valent technetium [Tc (VII)] extracted in the U-Pu organic product from the organic phase and discharge it into the high-level aqueous phase waste liquid, the main extraction and washing steps in the co-decontamination step After that, a Tc back-extraction step is installed, and the back-extracted Tc-containing aqueous solution is injected between the feed stage of the feed solution and the feed stage of the extraction solvent in the main extraction and washing steps to quantitatively increase Tc. Level Method of discharging into aqueous waste liquid.