RU2249267C2 - Method for recovering irradiated nuclear fuel (alternatives) - Google Patents

Abstract

FIELD: handling spent nuclear fuel.

SUBSTANCE: proposed method meant for recovering irradiated nuclear fuel to obtain joint solution of U and Pu includes extraction of U, Pu, Np, and Te from nitric acid solution by means of 30% solution of tributyl phosphate and separation of Np and Te. Next re-extraction of Pu is effected by means of depleted or regenerated four-valence uranium solution. The latter is supplied to central part of Pu re-extraction unit at flowrate ensuring weight proportion of Pu : U = 1 : 4 in re-extract. At the same time solution of 0.1 - 0.7 mole/l of nitric acid is fed to end of unit at flowrate excluding breakthrough of U(IV) with extract of U(VI). Joint re-extraction of plutonium and part of uranium is also conducted by completing agent solution supplied in constant composition and at flowrate affording complete re-extraction of plutonium. At the same time solution of 0.1 - 0.7 mole/l of nitric acid is fed to end of unit at flowrate ensuring desired amount of uranium supply to mixed re-extract.

EFFECT: provision for preventing spread of fissionable materials.

2 cl, 1 dwg, 2 tbl, 2 ex

Description

The invention relates to methods for processing irradiated nuclear fuel (SNF) with the provision of the reuse of regenerated nuclear materials in mixed uranium-plutonium oxide MOX fuel.

Known methods for producing MOX fuel include the production of individual fractions of plutonium oxide and U ²³⁵ depleted uranium oxide with their further mixing [1]. However, processes in which the separation of a separate fraction of plutonium oxide is present cannot be considered as ensuring the principle of non-proliferation of fissile materials. In the best way, the principle of non-proliferation is ensured when a joint composition of Pu and U is obtained in an aqueous solution and a mixture of oxides is obtained from it.

There is a known method of SNF processing [2], in which it is proposed to separate Pu and U in the desired ratio using AGK after separation of Np and Tc from an organic solution obtained by extraction and containing U, Pu, Np and Tc. However, this method does not indicate ways to implement this solution and maintain a given ratio of Pu and U when changing their composition within narrow limits. Another disadvantage of this method is the impossibility of including depleted uranium in the circulation. This method is the closest to the claimed method.

The present invention solves the problem of obtaining in the process of SNF reprocessing joint reextracts of uranium and plutonium of a given composition using both depleted and regenerated uranium, subject to the principle of non-proliferation of fissile materials.

To obtain an aqueous composition having a weight ratio of Pu: U = 1: 4, it is proposed from an organic solution obtained from reprocessing irradiated nuclear fuel of nuclear power plants (SNP of nuclear power plants) according to the Purex process scheme and containing Pu, Np, Tc and U, of characteristic isotopic composition , separate Np and Тc and, according to the first option, carry out Pu reextraction with a solution of depleted or regenerated U (IV) supplied to the middle part of the Pu reextraction unit with a flow rate that ensures 1: 4 ratios of Pu and U in the reextract, while feeding to the end of the block 0.1-0.7 mo l / l of nitric acid with a flow rate excluding the slip of U (IV) with the extract of U (VI). The resulting aqueous composition contains Pu and U (depleted composition). During this operation, the isotopic composition of the regenerated uranium changes insignificantly in the amount of U ²³⁵ .

According to a second embodiment of the invention, to solve the problem, it is proposed to reextract Pu and a part of the uranium from the organic solution after separation of Np and Tc with a complexing agent solution supplied with a constant composition and flow rate, which ensures complete re-extraction of plutonium with a 0.1-0 solution being simultaneously fed to the end of the block, 7 mol / l of nitric acid, with a flow rate ensuring a predetermined supply of uranium to the mixed re-extract, in the absence of washing the latter from U (VI) with a reverse extractant. Acetohydroxamic acid may be used as a complexing agent.

The combination of two technical solutions in one application is due to the fact that two of these methods solve the same problem in basically the same way, while ensuring the same technical result — obtaining a joint reextract of Pu and U with a weight ratio of 1: 4. On the other hand, the division into variants is due to the fact that in the first case, uranium, depleted or regenerated is used to obtain the Pu-U composition, and in the second case, uranium is taken from the organic extract. Thus, both options are equivalent for solving the problem.

The drawing shows a block diagram for obtaining a joint reextract of Pu and U. The block consists of 12 extraction cells, the numbers of which are indicated by numbers from 1 to 12. The arrows represent the incoming and outgoing flows, having the following notation: 1 - extract of Pu and U, 2 - extractant 3 - reextracting agent Pu, 4 - washing solution, 5 - reextracting agent Pu and U, 6 - U extract.

The method is as follows. The initial solution of the composition: НNО ₃ - 3 mol / L, U - 250 g / L, Pu - 2.4 g / L, Np - 180 mg / L, Tc - 200 mg / L, is fed to extraction with 30% tributyl phosphate in a diluent; a solution of Cr (VI) is used to oxidize Np (IV) to Np (VI). After Tc re-extraction with a solution of concentrated HNO ₃ and Np reductant solution, Pu extract enters the 3rd stage of the stripping unit for re-extraction of Pu with U (IV) solution obtained from depleted or regenerated U with a low U (VI) content, which is fed into the 9th stage of the stripping unit . In the last stage of the block is fed a solution of HNO _{3 of a} given concentration. Reextract U (IV), Pu (III) is processed in two stages with a reverse extractant. After separation of Pu, U is reextracted with a dilute solution of nitric acid.

In the second variant, after separation of Np and Tc, the extract of Pu and U enters the stripping of Pu with a nitric acid solution of acetohydroxamic acid at a constant flow rate. The adjustment of the Pu: U ratio of 1: 4 is carried out by a stream of nitric acid.

Example 1 (1st option).

Pu and U extract obtained during SNF reprocessing, after separation of Np and Тc, enters the 3rd stage of the extraction block for re-extraction of Pu with U (IV) solution, which is fed to the 9th stage of the re-extraction block. Re-extract of U (IV) -Pu (III) is processed in 2 steps with a reverse extractant. Technological indicators are shown in table 1. As a result of the operation, a re-extract was obtained containing Pu and U in the required ratio of 1: 4.

Table 1
Technological indicators of the process of obtaining a joint reextract of Pu and depleted U at a ratio of 1: 4 No. Stream name Consumption, m ³ / t The concentration of components, g / l HNO ₃ U (VI) U (iv) Pu 1 Pu Extract, U 15.0 6 66 0.64 2 Extractant 2.0 3 Stripping agent Pu 1,0 63 8 42 4 Wash solution 0.5 thirteen 5 Reextract Pu, U 1,5 80 <0.5 25.6 6.4 6 Extract 17 <1 60 0.002

Example 2 (2nd option).

Pu and U extract obtained during SNF reprocessing, after separation of Np and Tc, is fed to Pu reextraction with nitric acid solution of acetohydroxamic acid at a constant flow rate. The re-extractant solution Pu is fed into the middle of the block (along the extractant). To adjust the U content in the Pu reextract, the НNО ₃ solution _{of the} given concentration and the calculated ratio of the aqueous and organic phases flows is fed into the last stage of the block. Technological indicators are given in table 2. A change in the concentration of components while maintaining a given ratio affects in practice a linear function of the flow rate of the acid solution. The obtained reextract containing Pu and U in a ratio of 1: 4, meets the task.

table 2
Technological indicators of the process of obtaining a joint reextract of Pu and regenerated U and with a ratio of 1: 4. No. Name Consumption, Component concentration flow m ³ / t HNO ₃ U (VI) U (iv) Pu 1 Pu Extract, U 15.0 6 66 0.64 2 Extractant not served 3 Stripping agent Pu 0.5 thirteen 45 4 Wash solution 1.15 thirteen 5 Reextract Pu, U 1.65 70 23,2 5.8 fifteen 6 U extract fifteen <1 64.5 0.002

List of sources

1. Mixed Oxide Fuel (MOX) Explatation and Destruction in Power Readctors. Edited by Erich R. Merz, Carl E. Walter and Gennady M. Pshakin. NATO ASI Series., P. 165-191.

2. B.Ya. Zilberman, Yu.S. Fedorov, E.N. Mishin, L.I. Sytnik, O.V. Schmidt "Superpurex as a TBP-compatible process for the extraction and fractionation of long-lived radionuclides from spent nuclear fuel." - Chemical technology, 2002, t.3, No. 7, articles 33-37.

Claims (2)

1. A method of processing irradiated nuclear fuel to produce a joint solution of U and Pu, including extraction of U, Pu, Np, Tc from a nitric acid solution with a 30% tributyl phosphate solution, separating Np and Tc, characterized in that the subsequent re-extraction of Pu is carried out with a depleted or of regenerated tetravalent uranium supplied to the middle part of the Pu stripping unit with a flow rate providing the weight ratio Pu: U = 1: 4 in the stripping after washing the stripping with reverse extractant from excess U (VI), while feeding in horse solution unit 0.1-0.7 mol / l of nitric acid at a rate of precluding slippage U (IV) with an extract of U (VI). 2. A method of processing irradiated nuclear fuel to produce a joint solution of U and Pu, including extraction of U, Pu, Np, Tc from a nitric acid solution with a 30% tributyl phosphate solution, separating Np and Tc, characterized in that the joint re-extraction of plutonium and parts of uranium is carried out a complexing agent solution, supplied with a constant composition and flow rate, ensuring complete re-extraction of plutonium, while at the same time supplying 0.1-0.7 mol / L nitric acid to the end of the block with a flow rate providing a predetermined supply of uranium to mixed eekstrakt.