RU2727140C1 - Irradiated nuclear fuel extraction processing method - Google Patents

Abstract

FIELD: nuclear physics and equipment.SUBSTANCE: invention relates to extraction technology of irradiated nuclear fuel processing. Uranium, plutonium and associated elements are extracted from a nitrate solution, the extract is washed, re-extracting plutonium with a portion of uranium, washing the uranium extract from plutonium, neptunium and technetium residues, correcting the plutonium re-extract based on nitric acid content and oxidising plutonium (III) to plutonium (IV), repeatedly extracting plutonium with part of uranium, washing the extract of repeated extraction and repeatedly re-extracting plutonium. First re-extraction of plutonium with a portion of uranium is carried out with an aqueous stream containing reducers for plutonium (IV) and technetium (VII) at high ratio up to O:B = 35. Final purification of the uranium extract is carried out with a slightly acid complexone solution. Used complexone solution is delivered for washing of uranium-plutonium extract, concentrated nitric acid is added to the plutonium-technetium re-extract to a content of not less than 3 mol/l and passed through a catalytic oxidation column with a carbon catalyst. Treated re-extract is directed to affinity block. Plutonium re-extract with part of uranium is removed from process, and organic flow is directed to head extractor.EFFECT: invention reduces volume of liquid radioactive wastes.6 cl, 2 dwg, 3 tbl

Description

The invention relates to a technology for processing irradiated nuclear fuel (SNF), specifically - to an extraction technology for processing solutions obtained by dissolving spent nuclear fuel and passed the clarification stage.

Extraction processing of such solutions using tributyl phosphate (TBP) in the form of a 30% by volume solution in a hydrocarbon diluent (HCR) is generally accepted.

Processing, as a rule, is implemented in the form of three extraction cycles, each of which has its own circulating extractant circuit with an intracycle regeneration unit. In the first cycle, joint extraction and separate reextraction of uranium and plutonium is carried out, then uranium and plutonium are processed at their own refining cycles (see, for example, A.A. Kopyrin, A.I. Karelin, V.A.Karelin. Production technology and radiochemical reprocessing of nuclear fuel. M., "Publishing house Atomenergoizdat", 2006)

Disadvantages of this method:

- intracycle regeneration of the extractant of the plutonium refining cycle is a significant source of saline LRW of the average activity level;

- to exclude the ingress of plutonium into the carbonate-alkaline regenerates, an increased consumption of the stripping agent is required, which leads to the dilution of the plutonium stripping.

The closest is the method of reprocessing spent nuclear fuel, known as the SOEX-process (P. Baron, B. Dingch, M. Masson, F. Drain, J.-L. Emen. A method of regenerating spent nuclear fuel and producing mixed uranium-plutonium oxide. Patent RF No. 2431896, IPC G21C 19/46, 2011). The method provides for the oxidative treatment of plutonium strip obtained by separating plutonium from uranium, and subsequent processing of the strip according to the refining scheme (repeated extraction of plutonium and part of uranium, washing of the extract and stripping of plutonium and part of uranium). To carry out these operations, a circulating extractant of the head uranium-plutonium cycle is used.

Thus, the extraction refining of plutonium operates in the subcycle mode, i.e. cycle without its own extractant circuit. The organic stream after the plutonium re-stripping operation is added to the uranium-plutonium extract before the first plutonium stripping. This solution removes the requirement for completeness of repeated stripping, which greatly simplifies this operation.

In the description of the invention to the patent, options are given that slightly differ from the above (other reducing reagents, options for the formation of streams), but everywhere the sequence of operations for transferring plutonium from the head cycle to its own subcycle and removing plutonium from the extraction processing process in the form of a reextract subcycle is preserved.

In technical essence, this method is the closest to the claimed method and is chosen as a prototype.

The disadvantages of the prototype are:

1. Low content of plutonium (4.2 g / l) in the stream supplied for re-extraction, and, as a consequence, a large flow of re-extraction raffinate 2.59 m ³ / tТМ (tТМ - ton of heavy metals, U + Pu), requiring disposal.

2. Low ratio of flows of organic (O) and aqueous (B) phases on repeated extraction of plutonium, O: B = 1: 4.35. With such a ratio, losses of plutonium with raffinate are possible with slight deviations of the extraction process from the routine mode.

Objective: development of a method for SNF reprocessing, which ensures reprocessing of concentrated plutonium solutions at the refinery in a mode with an increased relative flow rate of the extractant flow.

The technical result of the proposed invention is to reduce the volume of liquid radioactive waste (LRW) generated by the technological process, and to increase the technological stability of the process.

This result is achieved in the method of SNF extraction reprocessing, according to which uranium, plutonium and accompanying neptunium, technetium and zirconium are extracted from a nitric acid solution, the extract is washed, plutonium is reextracted with a part of uranium, the uranium extract is washed from plutonium residues, neptunium and technetium are reextracted, plutonium is corrected content of nitric acid and oxidize plutonium (III) to plutonium (IV), re-extract plutonium with a part of uranium, wash the extract of repeated extraction and re-extract plutonium, while the first re-extraction of plutonium with a part of uranium is carried out by processing the extract with an aqueous stream containing reducing agents for plutonium ( IV) and technetium (VII) with a high ratio of organic (O) and water (B) streams up to O: B = 35, the uranium extract is purified with a weakly acidic complexone solution, mainly diethylenetriaminepentaacetic acid (DTPA), the spent complexone solution is sent to uranium washing -plu tonium extract, concentrated nitric acid is added to the plutonium-technetium reextract to a content of at least 3 mol / l, after which the reextract is passed through a catalytic oxidation column with a carbon catalyst, the reextract, which has undergone oxidative treatment, is sent to the refining unit, where the operations of uranium and plutonium extraction are carried out, washing the extract, reducing extraction of plutonium and, if necessary, washing the re-extract with a circulating extractant, then the plutonium re-extract with a part of uranium is removed from the process, and the organic flow leaving the refining unit and containing excess uranium and a small amount of plutonium is sent to the head extractor in the zone of intensive mass exchange ...

FIG. 1 shows a schematic diagram of SNF extraction processing according to the proposed method (operations of uranium re-extraction and extractant regeneration are omitted). The feed flow enters stage 17 of the head extraction unit 1. The circulating extractant can enter unit 1 both in one flow (stage 8) and in two (stages 8 and 15); 2 593 831, publ. 10.08.216, bull. No. 22).

An extraction column with the location of the inlet nozzles and the length of various mass transfer zones similar to the block of discrete extractors can be used as the head extractor.

The extract leaving unit 1 enters the mass-transfer separator 2, where the extract is purified from emulsion entrainment and primary purified from fission products. The partially washed extract goes to the washing unit 3, where the extract is purified from zirconium and the predominant part of neptunium. The spent wash solution is either washed from uranium and plutonium with a circulating extractant (optionally), or processed at the head unit together with the feed stream.

The next operation is the concentrating stripping of plutonium and technetium, carried out using a mass transfer separator 4. The ratio of the flows of the organic (O) and aqueous (B) phases in this operation, O: B = 25-35, which is not available for other mass transfer apparatus used in the extraction technologies (V. I. Volk, S. N. Veselov, A. A. Zherebtsov, V. N. Rubisov. Radiochemistry, 2010, vol. 52 No. 5 pp. 428-432). The stripping agent contains a balanced mixture of reducing agents, carbohydrazide (CG) (N ₂ H ₃ ) ₂ CO and diformylhydrazine (DPG) N ₂ H ₂ (COH) ₂ , the first of which has high kinetic parameters for the reduction of Pu (IV) → Pu (III) , the second in terms of reduction Tc (VII) → Tc (IV).

It is possible to include in the composition of each of the stripping agents an amino acid, glycine (GL) H ₂ NCH ₂ COOH, the presence of which reduces the effective concentration of nitric acid due to binding to the amino group, which increases the efficiency of the reducing agents.

In preliminary experiments, the concentrations of the reagents were determined, the excess of which no longer led to a noticeable increase in the operational yield of plutonium and technetium in the re-extract. It was 99.5-99.9% for technetium and 93-97% for plutonium, while 99% of the plutonium in the organic stream leaving separator 4 is plutonium (III), which is associated with a high ratio of O: B fluxes. This circumstance makes it possible to easily wash the uranium extract from the remains of plutonium and neptunium (from the technetium, the extract is washed off already at separator 4) with a weakly acidic solution of a complexing agent, DTPA, at unit 5.

At block 3, plutonium (III) is re-oxidized, which is supplied with the spent washing solution from stage 42 of block 5 to stage 41 of block 3. In addition to oxidation with nitric acid, plutonium (III) is oxidized by neptunium (IV) supplied with the extract to stage 25 of block 3. As a result, neptunium is localized in the raffinate of Unit 3, while the loss of plutonium with raffinate is less than 0.01% of the incoming one.

Thus, the technological interconnection of the washing unit 3, separator 4 and washing unit 5 provides the following indicators:

- Unit 3 performs the task of purifying the extract of uranium, plutonium and technetium from zirconium and, in conjunction with Unit 5, provides almost complete discharge of neptunium into the raffinate;

- separator 4 performs the task of obtaining a concentrated plutonium reextract with a quantitative yield of technetium into the plutonium stream;

- Unit 5 solves the problem of purifying the uranium extract from plutonium and neptunium, while the unit raffinate is excluded from the LRW volumes.

The uranium extract leaving Unit 5 has been purified from plutonium, neptunium and technetium. And, depending on the achieved purification from fission products, it is sent either for additional washing followed by uranium stripping, or immediately for uranium stripping (not shown in Fig. 1).

The first plutonium-technetium re-extract (stream 84) is acidified with concentrated nitric acid (stream 88) to an acid content of 3.25 ± 0.25 mol / L and enters the catalytic oxidation column (CCO) 6, where the excess of reducing agents is destroyed and Pu ( III) to Pu (IV) and Tc (IV) to Tc (VII), after which it is sent to stage 57 of the refining unit 7. At unit 7, extraction operations (stages 49-57), two different acid washings (stages 58-64) and repeated plutonium strip extraction (stages 68-70 in the variant of washing the strip from uranium and stages 65-70 in the variant of obtaining uranium-plutonium strip).

Example 1. A mathematical simulation of the process was performed in accordance with the scheme shown in FIG. 1. The simulation results are summarized in Table 1.

Shown in FIG. 2 and in table. 1 operating mode demonstrates the production of block 3 - mass transfer separator 4 - block 5 of uranium extract, free from neptunium and problematic fission products - zirconium and technetium.

In the considered regime, the final product of the plutonium branch of the scheme is the second plutonium re-extract, which is practically uranium-free. This option is provided for the separate production of uranium and plutonium oxides with their subsequent mixing in the production of various types of uranium-plutonium fuel.

When obtaining a joint uranium-plutonium oxide with a plutonium content at the level of 50-60% of the total metals (the so-called master product, the production of which also includes the prototype method by complex and prolonged manipulations with the second plutonium reextract), it is possible to obtain a joint uranium plutonium reextract with the required ratio of uranium and plutonium with some adjustment of the operating mode of Unit 7.

Example 2. Mathematical modeling of the operation of the refining unit 7 in the mode of obtaining a uranium-plutonium reextract was performed (Fig. 2). The supply of circulating extractant 94, which washes out uranium from the strip stream 95, is excluded. Glycine (streams 98 and 99) is additionally fed to the reduction strip section as a process stabilizer. The extraction and washing operations of the extract are similar to those shown in FIG. 1 and tab. 1.

The simulation results are presented in table. 2.

As follows from the data table. 3, insignificant changes in the mode of the second plutonium strip extraction make it possible to obtain in the plutonium branch of the scheme a uranium-plutonium strip extract already corresponding to the master product (50.4% plutonium in option 1 and 57.8% in option 2), which favorably distinguishes the proposed method from the method -prototype.

Comparable indicators of the proposed method and the prototype method are summarized in table. 3.

From the data table. 3 it follows that in the proposed method the specific volumes of flows requiring subsequent disposal (flows 95 and 96) are significantly less than in the prototype method.

It is important that in the proposed method, the specific volume of the highly concentrated first reextract of plutonium is 4 times less than in the prototype method. This circumstance has the consequence that the operation of the repeated extraction of plutonium in the proposed method is carried out at a ratio of flows of phases O: B = 1: 1.43, and in the prototype method at O: B = 1: 4.35, i.e. in the proposed method in each stage the value of the extraction coefficient is three times higher than that of the prototype method, which significantly increases the technological stability of the process in the proposed method.

Thus, a comparison of the indicators of the prototype method and the proposed method fully demonstrates the advantages of the proposed method.

Claims (6)

1. A method for SNF extraction processing, including the extraction of uranium, plutonium and associated neptunium, technetium and zirconium, washing the extract, stripping plutonium with a part of uranium, washing the uranium extract from residues of plutonium, neptunium and, if necessary, technetium, adjusting the composition of the stripping plutonium with a part uranium and subsequent oxidative treatment of the reextract, re-extraction of plutonium with a part of uranium, washing of the re-extraction extract and re-re-extraction of plutonium, characterized in that the first re-extraction of plutonium with a part of uranium is carried out by processing the extract with an aqueous stream containing reducing agents for plutonium (IV) and technetium (VII ) with a high ratio of organic (O) and water (B) streams, up to O: B = 35, the uranium extract is purified with a weakly acidic solution of a complexone, mainly diethylenetriaminepentaacetic acid, the spent solution of a complexone is sent to the washing of the uranium-plutonium extract, to the pluto strip extract nium-technetium add concentrated nitric acid to a content of at least 3 mol / l, after which the re-extract is passed through a catalytic oxidation column with a carbon catalyst, the re-extract that has undergone oxidative treatment is sent to the refining unit, where the operations of uranium and plutonium extraction, extract washing, reductive extraction are carried out plutonium and, if necessary, washing the re-extract with a circulating extractant, then the plutonium re-extract with part of the uranium is removed from the process, and the organic flow leaving the refining unit and containing excess uranium and a small amount of plutonium is sent to the head extractor in the zone of intensive mass transfer. 2. The method according to claim 1, characterized in that the concentrating strip extraction of plutonium and technetium with a part of uranium is carried out in a mass transfer separator. 3. The method according to PP. 1, 2, characterized in that the plutonium and technetium stripping agent supplied to the separator contains up to 0.45 mol / L of carbohydrazide, up to 0.15 mol / L of diformylhydrazine and up to 0.45 mol / L of nitric acid. 4. The method according to claim 1, characterized in that carbon hemosorbent VNIITU-1 is used as the carbon catalyst. 5. The method according to claim 1, characterized in that the plutonium stripping agent supplied for repeated stripping contains up to 0.65 mol / l of carbohydrazide and up to 0.65 mol / l of nitric acid. 6. Method according to pp. 1, 3, 5, characterized in that the stripping agents additionally contain up to 1 mol / l glycine.

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