RU2132578C1 - Method of processing irradiated nuclear fuel on nuclear power plants - Google Patents

Abstract

FIELD: nuclear power engineering. SUBSTANCE: method consists in extraction processing of irradiated nuclear fuel using diluted tributyl phosphate and includes following steps: extracting Zr together with U, Pu, Np, and Tc; washing extract under conditions favoring retention of Zr in extract; completely reextracting Np, Tc, and Zr into nitric acid containing hydrazine nitrate and hydrogen peroxide; separating Pu with mediation of U(IV); and barrier washing of extract to remove simultaneously Pu and Np by the aid of lower hydroxamic acids as nonionic complexing agent. Solution from the latter operation is returned into combined Np, Tc, and Zr reextraction stage. EFFECT: enabled purification of uranium product in one cycle. 5 cl, 2 dwg, 1 tbl

Description

 The invention relates to the field of spent nuclear fuel reprocessing after dissolution in nitric acid and is a complex extraction process using tri-n-butyl phosphate (TBP) as an extractant, with the aim of extracting, purifying and separating U and Pu for reuse. However, some other actinides and fission products, such as Np, Zr and Tc, are extracted to some extent in TBP or coextracted with U and Pu. Therefore, the first extraction cycle includes a number of stages (operations) carried out in separate extraction sections and preventing pollution of U and Pu by these elements or their undesirable effect on the separation of U and Pu.

 The known method [1], which provides, after the operation of extraction of uranium and plutonium proceeding with the capture of zirconium and technetium, first separation of zirconium in a special washing operation at a nitric acid concentration of 1-4 mol / l with its removal in highly active raffinate, and then separation of technetium during a two-zone operation, where a solution of 0.5-3 mol / L nitric acid is supplied to the extreme stage, and a solution of stronger nitric acid is supplied between the introduction of this solution and the input of the uranium, plutonium and technetium extract, with U and P u, transferred to an aqueous solution, additionally washed with a stream of extractant and returned to the extraction operation. Later, the UP-3 plant (France) specified [2] the concentration of nitric acid during Zr reextraction (2 mol / L) and in Tc reextract (4-5 mol / L); the flow of Te re-extract relative to the extract is 0.2-0.25 (see the example in [1]).

 The processed extract of U and Pu enters into their separation with the reduction of plutonium by uranium (IV). To increase efficiency, the operation can be carried out in two stages, as was suggested in the scheme of the THORP plant [3], the second stage being carried out with a longer phase contact and is called the plutonium barrier.

 Subsequently, both of these techniques were included in a single process [4], adopted for implementation during the modernization of the UP-2 plant (France) and Rokkasho-mura (Japan). Such a technological scheme (see Fig. 1) is taken by us as a prototype. It consistently includes the operations of extraction of U, Pu, Np and Tc with diluted TBP, washing the extract from fission products, including Zr, two-zone Tc reextraction with 4-5 mol / L nitric acid at a ratio of aqueous / organic phase flows O / B = 5, separating U and Pu with Np transfer together with U when using U (IV) as a reducing agent, barrier washing from Pu, U-stripping and extractant regeneration; both operations on intermediate re-extraction are accompanied by additional extraction of U and other extraneous extractive elements. However, in this process, Zr is washed off in the raffinate of the 1st cycle in a very narrow range of the acidity of the washing solution, Tc is back-extracted only by 80%, while Np is "smeared" between all flows of the 1st cycle, and separation of U and the main part of Np is achieved only in uranium purification cycle. The disadvantages of the process are the return of the extract from the additional extraction of U and Pu from the technetium reextract to the uranium-free zone of the head extractor, which makes it difficult to control the process of saturation of the extract with uranium. An additional problem is the purification of Pu reextract from Np, which requires regulation of the acidity of this reextract.

 There is also a method where a joint reextraction of Np and Tc is proposed by using an organo-soluble reducing agent [5], but this treatment has proved to be ineffective. Another process [6] provides for the complete extraction of Zr together with U, Pu, Np, Tc and subsequent removal of Zr into a separate weakly acidic stream, and it is supplemented by the method [7], in which during the separation of U and Pu with Np using a salt-free complexing agent technetium is retained in the extract, followed by Tc reextraction using the same complexing agent in a mixture with hydrazine. Acetohydroxamic acid is proposed as such a complexing agent. The disadvantages of this process as a whole are the lack of washing of the extract, which is replaced by direct Zr stripping, which leads to the capture of a highly active emulsion and contamination of medium-level technological waste, as well as difficulties in cleaning Pu reextract from Np, as well as the Tc reextract with Np from traces of Pu in the presence of a complexing agent, which is slightly susceptible to influences during the process [7].

The objective of the invention is the creation of a technological process for reprocessing spent nuclear fuel of nuclear power plants, free of these disadvantages in comparison with the prototype [1]. The problem is solved by sequentially carrying out the following operations (see figure 2):
- extraction of Zr together with U, Pu, Np and Tc from the combined stream of the initial and washing solutions in 30% TBP in a diluent;
- washing the extract with 2.5-6 mol / L HNO ₃ when the ratio of the flows of the organic and aqueous phases (O / B) is more than 10;
- two-zone complete reextraction of Zr, Tc and Np in 3.8 - 4.7 mol / L HNO ₃ containing hydrogen peroxide and hydrazine nitrate, with O / V = 0.8 - 2.0 in the first zone and at 0, 7-1.5 mol / L HNO ₃ at O / B = 4-10 in the second zone, the combined stripping stream being created by introducing more concentrated nitric acid with hydrogen peroxide and hydrazine into a weakly acid stripping solution;
- additional extraction of U and Pu from the combined joint reextract of Zr, Tc and Np at O / B = 0.2-0.3, which is unfavorable for the reextraction of Np (IV), moreover, this extract is combined with the main one;
- separation of Pu and U in the absence of Tc using U (IV), the latter being completely extracted together with U (VI) with the complete extraction of residual Np (IV) due to the acidification of the additional extraction zone while maintaining the reducing medium;
- barrier washing of the uranium extract from residual amounts of Np and Pu using a salt-free complexing agent, such as one of the lower hydroxamic acids (formo- or acetohydroxamic), moreover, the washing solution is recycled to the strongly acidic stripping zone of Zr, Tc and Np to destroy the complexing agent and its compounds with Pu and Np;
- U stripping followed by regeneration of the extractant.

 The process is intended for the reprocessing of spent nuclear fuel at a nuclear power plant with a standard burn-up of 40 GW • day / t and an aging of 3 years; however, this does not mean the impossibility of using this method in other similar cases.

 The extraction of U, Pu, Np, Zr, and Tc with diluted TBP is limited by the Zr distribution, for which an O / B ratio of ≥ 3.0 is necessary [3], where the water flow is formed together with the initial and wash solutions; while the acidity of the raffinate exceeds 3 mol / L. Therefore, the washing of the extract is carried out with the minimum possible accumulation of Zr in the extraction zone of the combined extraction section, which is practically achieved by limiting the concentration of nitric acid in the washing solution in the range of 2.5-6 mol / L at an O / B ratio> 10 (the upper O / B limit is limited extraction equipment capabilities). At a higher acid concentration, the distribution coefficients of most elements, in particular Tc, decrease, while at a lower acidity, partial hydrolysis of Zr takes place.

 Joint reextraction of Zr, Tc and Np is carried out, as mentioned above, in a two-zone process, with all technetium, the main part of zirconium and neptunium in the pentavalent state, being reextracted in the first zone. Adding hydrogen peroxide with 5-8 mol / L nitric acid avoids the precipitation of uranium peroxide. The process regime in the second zone is dictated by the re-extraction of all Zr and Np (IV), provided that Pu (IV) is extracted. Ultimately, the combined effectiveness of all three operations, including additional extraction, is determined by the separation of Pu (IV) and Np (IV) with minimal acidity, which provides Tc reextraction. This leads to some retention of Np by an extract containing U and Pu, which necessitates the separation of Pu (III) and U with Np (IV) when using hydrazine stabilized U (IV). However, the latter is significantly simplified in comparison with the prototype [1], since only traces of neptunium, which managed to recover to the tetravalent state, are transferred to the separation operation of U and Pu. At the same time, the U (IV) consumption in the reduction department of Pu is also significantly reduced.

Of particular note is the order in which reagents are introduced during barrier flushing from actinides. At the process temperature of 50-65 ^o C, a hot uranium concentrate containing hydrazine is introduced into the end part of the extraction section (at the outlet point of the washed extract), which prevents the extraction of the corresponding carboxylic acids (formic or acetic) and helps to clean U from Ru-106 (including including the next then back extraction U), while the hydroxamic acid itself is introduced into the middle (cold) part of this section. Recycling this solution from washing to a joint re-extraction operation supplies a certain amount of hydrazine nitrate to restore Np (VI) and preserve Pu (IV) at a given acidity of the process, as well as to destroy residual complexing agents.

 Examples.

 Example 1

A clarified solution of spent fuel based on uranium oxides extracted from pressurized water reactors and having a burnup of 38 GW • day / t and aging for 3.5 years is fed into the extraction process. The initial solution contains 330 g / L U, 3.2 mol / L HNO ₃ , 3.1 g / L Pu, 155 mg / L Np, 275 mg / L Tc, 1.2 g / L Zr and all other side actinides and fission products are proportional to burnout, with β, γ-emitting nuclides contaminating at 280 Ci / L (including 24 Ci / L Ru-106).

 The flow chart in the embodiment of using multi-stage sections of mixer settlers or centrifugal extractors is shown in Fig. 2, with the total number of stages being 65. The figure visually shows approximate relationships between the zones of the technological process and their belonging to one or another extraction section. The sequence of operations and the conditions for their implementation correspond to the description; in more detail, the optimal conditions are described in the table, which shows the composition of the solutions and their costs in relation to the initial solution.

It should be noted that the concentration of HNO ₃ in the joint reextract of Zr, Tc and Np is 4.2 mol / l, and the arrival of the reextract does not exceed the consumption of extractant in the head section. In addition, plutonium re-extract contains about 10 g / l Pu at a content of about 1% U (IV) and <0.05% Np (IV) with respect to Pu and at a molar concentration of HNO _{3 of} 2.2 mol / l.

 From the same table it follows that all the declared final indicators for the quality of the uranium product are achieved in one extraction cycle with a typical structure.

 Examples 2-4.

 The process is carried out as in example 1, however, changes are made to the conditions for the extraction of Zr, Tc and Np, consisting in the fact that the operation is carried out at a flow ratio of O / B = 0.7 while maintaining the concentration of nitric acid 4.2 mol / l in the joint reextract. In this case, the flow of extractant for additional extraction is increased in proportion to the flow rate of the reextract. As a result, the reextract contains about 40 mg / L Pu in addition to Zr, Tc, and Np.

 With an increase in the concentration of nitric acid to 4.7 mol / L, the concentration of Pu decreases to 1 mg / L without deterioration of other indicators. While maintaining the molar acid concentration of 4.7 mol / L and increasing the O / B flux ratio to 2.0, the concentrations of Np and Zr in the effluent extract increase to ~ 20 mg / L each, and the technetium concentration reaches 5 mg / L.

 Example 5

 The process is carried out as in example 1, but the reextraction of Zr, Tc and Np is carried out at a concentration of nitric acid in a joint reextract of 3.8 mol / L. The concentration of Pu in this reextract is 2-3 mg / L, and the concentration of Tc B of the effluent is increased to 7 mg / L.

The inventive method allows to extract from a highly active starting solution more than 99.95% U and Pu, 99.8% Np and Tc, 98.5% Zr from a highly active starting solution, while Pu re-extract contains (compared to the initial solution) less than 0, 5% Np and Tc, while the coefficient of purification U from Pu and Np is about 2-5 • 10 ⁶ and 5 • 10 ⁴ , respectively. The cleaning coefficient of fragmentation β, γ-emitters is 3 • 10 ⁶ (5 • 10 ⁴ from Ru-106). Thus, the process allows you to send the aqueous uranium product directly to thermal denitration without intermediate processing.

List of sources
1. Boullis B., Gue JP., Bernard S. "Process for separating technetium present in an organic solvent with zirconium and at least one other metal such as uranium or plutonium, which can be used especially for the reprocessing of spent nuclear fuels" . European patent N 0270453 A1, priority from 11.26.87.

 2. Fournier W., Hugelmann D., Bernard C., Leudet A. et al. "Purex process improvements for UP-3 spent fuel reprocessing plant ...", Proceed, of Int. Solv. Extr. Conf. "ISEC'90", Kyoto (Jap.), July 1990, part A, p. 747-752. Amsterdam; Elsevier, 1992.

 3. Philips C. "Development of THORP Chemical Separation Plant". ATOM, 1989 (August), N 394, p.13-20.

 4. Bernard S., Miquel P., Viala M. "Advanced Purex process for the new reprocessing plants in France and Japan". Proceed, of Third Int. Conf. "RECOD'91", 1991, v. 1, p. 83, AESJ & JAIF, Sendai (Jap). - prototype.

 5. Maeda M., Fujine S., Uchiyama G. et al. "A process concept of an advanced reprocessing..." Proceedings of Int. Conf. "Global'95", Versailles (France), Sept. 1995, V.1, p. 566-563.

 6. Dzekun E. G., Mashkin A. N., Zilberman B.Ya., Fedorov Yu.S. and others. "The processing of irradiated fuel elements." USSR patent 1804652 (priority from 03.10.89).

 7. Zilberman B. Ya., Mashkin A.N., Nardova A.K., Sytnik L.V. and others. "The process of processing irradiated fuel of nuclear power plants." RF patent 2012075 (priority from 05/14/92).

Claims (5)

1. A method of processing irradiated nuclear fuel (SNF) of a nuclear power plant, including extraction of U, Pu, Np and Tc with diluted TBP, washing the extract, reextracting Tc with medium concentration nitric acid, separating U and Pu using U (IV) when transferring Np with organic uranium flow, barrier flushing from plutonium, additional extraction of technetium and plutonium from reextracts, as well as U stripping and extractant regeneration, each operation being carried out in the corresponding section of extraction apparatus, characterized in that Zr is extracted with together with U, Pu, Np and Tc, completely Np, Zr and Tc are completely re-extracted from the washed extract in 3.8 - 4.7 mol / L HNO ₃ containing hydrazine nitrate in the presence of a weak oxidizing agent with respect to organic and aqueous flows O / B = 0.8 - 2.0, Pu is separated by acidifying the Pu reextract in the zone of additional extraction U, and the barrier washing is carried out simultaneously from Pu and Np using a non-salt-forming complexing agent, and the solution from the washing is returned to the joint stripping of Np, Zr and Tc. 2. The method according to claim 1, characterized in that Zr is extracted from the combined stream of the source and wash solutions, moreover, the extract is washed with 2.5 - 6.0 mol / L HNO ₃ with a ratio of the extractant to the wash solution of more than 10. 3. The method according to claim 1 or 2, characterized in that the water stream for the complete reextraction of Zr, Tc and Np is formed by mixing in the extraction section 5-8 mol / L HNO ₃ containing hydrazine nitrate and hydrogen peroxide as a weak oxidizing agent , the return flow from the barrier washing and stripping solution containing 0.7-1.5 mol / L HNO ₃ at the inlet and used to reextract residual Zr in the same extraction section at O / B = 4-10, and the auxiliary extraction extract U and Pu is combined with the main extract stream.  4. The method according to any one of claims 1 to 3, characterized in that the additional extraction of U (IV) and Np (IV) from the Pu (III) reextractor is carried out by acidification of the stripping stream with nitric acid containing hydrazine supplied to the middle part of the section extraction.  5. The method according to any one of claims 1 to 4, characterized in that the residual amounts of Np and Pu are washed out using one of the lower hydroxamic acids, for example, formo- or acetohydroxamic, when it is fed into the middle of the barrier section, into which the tail is fed hot uranium concentrate containing hydrazine nitrate, wherein the solution from the barrier section is returned to the strongly acidic zone of the section to reextract Np, Zr and Tc.

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