RU2242812C2 - Isotope reduction of regenerated uranium - Google Patents

Abstract

FIELD: recycling nuclear power-generating materials.

SUBSTANCE: proposed method intended to enable reuse of uranium extracted from spent nuclear fuel in fuel cycle of light-water reactors by reducing its isotope involves increase of ²³⁵ U isotope content in regenerated uranium to 2.0 - 7.0 mass percent at the same time reducing relative concentration of isotopes ²³² U, ²³⁴ U, and ²³⁶ U including direct enrichment of raw regenerated uranium hexafluoride in gas-centrifuge isotope-separating cascade; raw regenerated uranium is enriched by ²³⁵ U isotope to content of 21.0 - 90.0 mass percent in double cascade during extraction of reduced fuel material with ²³⁵ U isotope content not over 50.0%, preferably 2.0 - 36.0 mass percent, through waste flow of second ordinary cascade. Enrichment in first ordinary cascade is conducted up to ²³⁵ U content of 3.5 - 10.0 mass percent or 21.0 - 35.0 mass percent. This method provides for attaining desired enrichment of regenerated uranium by ²³⁵ U isotope at low content of noxious isotopes.

EFFECT: enhanced quality of regenerated uranium reduction due to absolute and relative control of noxious isotope content.

7 cl, 3 dwg, 5 tbl

Description

The invention relates to technology for the recycling of nuclear energy materials and can be used to return uranium extracted from spent nuclear fuel to the fuel cycle of light-water reactors.

Uranium regenerated from spent nuclear fuel (SNF) is a valuable source for reuse in light-water reactors (LWR), since it contains a fissile isotope ²³⁵ U in an amount not less than natural uranium (Table 1) and allows saving the latter. For this purpose, the regenerated uranium should be enriched in the ²³⁵ U isotope to a content of 2.0-5.0 wt.%, And the mass of the obtained fuel material should not exceed the load of the LWR core, from where SNF with raw uranium regenerate is unloaded.

However, raw uranium regenerate is different from natural uranium. Firstly, it contains the residual amount of radionuclides formed as a result of nuclear transformations: transuranium elements (neptunium, plutonium), fission products (ruthenium-106, cerium-144, antimony-125, etc.), technetium-99. Temporary exposure, repeated radiochemical purification, and the conversion of regenerated uranium to hexafluoride make it possible to lower the content of these radionuclides to an insignificant level. Secondly, there are significant differences in isotopic compositions. In raw uranium regenerate, in addition to the difference in the content of ²³⁵ U, significant quantities of harmful isotopes ²³² U, ²³⁴ U and ²³⁶ U are present in significant quantities, which cannot be separated during chemical processing and the increased content of which leads to the main difficulties in using the regenerated uranium for the production of reactor fuel.

An increase in the content of the ²³⁴ U isotope is associated with a deterioration in the radiation situation due to internal exposure of personnel. The ²³² U isotope with its daughter elements forms the bulk of the external radiation exposure of personnel at all stages of the processing of regenerated uranium and the manufacture of fuel. The ²³⁶ U isotope is a neutron absorber and degrades the performance of reactor fuel. The presence of this isotope requires expenditures for the enrichment of regenerated uranium with the ²³⁵ U isotope by 0.2-0.6 (on average 0.28) fractions of the content of ²³⁶ U.

The content of harmful isotopes of uranium depends on the recycling scheme, the number of cycles of use of regenerated uranium and other parameters.

Here, the authors draw the attention of the examination to the following:

1) in the text, the term "content" or "concentration" the authors understand the mass content (concentration) of a particular uranium isotope only in a mixture of uranium isotopes. These terms do not refer to the chemical form of uranium,

2) the terms "nuclear (reactor) fuel", "fuel material" are understood, depending on the context, either hexafluoride of a mixture of uranium isotopes, or products (uranium dioxide powders, tablets, fuel rods, fuel assemblies), which can be further obtained from the above hexafluoride.

As is known, the enrichment of raw uranium regenerates with the ²³⁵ U isotope can be carried out by two methods: either by adding this uranium isotope of higher enrichment (for example, 10-20%), or by enriching, or by enriching ²³⁵ U to a higher content on isotope-separating uranium factories, or direct enrichment.

It is known the experimental use of regenerated uranium in the LWR cycle by direct enrichment of raw uranium regenerates up to 3.0-4.0% by gas diffusion (Rice T.G. Production of oxide fuel from regenerated uranium in existing plants -Atomic engineering abroad 1994, No. 12, p . 19) (analogue). The method leads to the concentration of harmful isotopes in enriched regenerated uranium and, accordingly, a significant increase in their negative impact. For this reason, restrictions are introduced on the initial maximum concentration of harmful isotopes in uranium hexafluoride directed to the separation of isotopes. US Standard ASTM C 787-90 provides, for example, the following values:

²³² U 5 · 10 ^-7 %; ²³⁴ U 4.8 · 10 ^-2 %; ²³⁶ U 8.4 · 10 ^-1 %.

Evaluation of the components of the external radiation dose in the manufacture of fuel for LWR from regenerated uranium after direct enrichment showed (Rice T.G. Production of oxide fuel from regenerated uranium in existing plants. -Atomic Engineering Abroad, 1994, No. 12, p.21), which During the operations with containers containing enriched regenerated uranium (in the form of hexafluoride, oxides, etc.), personnel receive 75% of the effective dose of radiation, and to reduce the dose rate, access to containers must be limited, radiation shielding is installed newly etc.

On the other hand, from calculations of the burning of the fissile ²³⁵ U isotope and the accumulation of harmful isotopes during fuel irradiation in the WWER-type LWR, it follows that with such a reduction, regenerated uranium degrades as reactor fuel due to absolute accumulation of ²³⁶ U (5-6%) to the third cycle of use, and the relative content of ²³² U after the second cycle will exceed the permissible level. This is one of the reasons why the recovered uranium has not yet been recycled on an industrial scale, but is stored and regarded as a national reserve (Le Baster J. Recycling and preparation of mixed oxide fuels: Achievements in France and Belgium. - Nuclear Technology Abroad, 1995, No. 11, p. 7).

It is known that the raw uranium regenerate from VVER spent fuel is isotopically enriched by mixing with medium- and highly enriched regenerated uranium (SOU, HEU) obtained from reprocessing of spent nuclear reactors and other spent nuclear fuel of high enrichment (A. Polyakov and others. Status and prospect technologies for spent fuel processing. - Nuclear energy, vol. 89, issue 4, p. 286, Fig. 1) (analogue). The method does not require enrichment of raw uranium regenerate in isotope-separation uranium plants.

The method is economically justified for raw uranium regenerates with a ²³⁵ U isotope content of more than 1% and when enriched to 2.2-2.6%. Under other conditions, the mass of fuel material exceeds the mass of raw uranium regenerate. In addition, since the regenerates of SOU and HEU also contain harmful isotopes, additional enrichment does not lead to the purification of regenerated uranium from radiation-hazardous nuclides and their content increases. After double recycling, it is necessary to develop special radiation protection measures at all stages of treatment of regenerated uranium (A comparative analysis of the radiation situation in the manufacture of nuclear fuel from regenerated uranium // Prep. / All-Russian Research Institute of Inorganics, material. - 1997, No. 1, pp. 1-21 (Abstract Journal: Technology of Inorganic Substances and Products, 1998, No. 17, p. 22, Abstract 17L273)).

A significant disadvantage of the analogue is the limited amounts of SOU and HEU, as well as raw uranium regenerate with ²³⁵ U isotope content of more than 1%, extracted from SNF. The involvement of weapons-grade HEU in mixing makes this method economically unprofitable due to a significant loss of separation work (the main component of the cost in terms of preparing reactor fuel for LWR (Sinev I.M., Baturov B.B. Economics of atomic energy: Fundamentals of technology and Economics of Nuclear Fuel. - M.: Energoatomizdat, 1984, p. 379) contained in the weapon HEU.

Highly enriched uranium (HEU) is uranium with ²³⁵ U enrichment above 20%; enriched uranium (СОУ) from 5 to 20%; low enriched uranium (LEU) from 1.5 to 5%; raw uranium of natural origin 0.711%; depleted uranium less than 0.711%.

A method for restoring the suitability of regenerated uranium for reuse in LWR reactors (Patent No. 2113022 Russia, IPC 6 G 21 C 19/42) (analogue) is proposed, which consists in simultaneously reducing the concentrations of ²³² U, ²³⁴ U and ²³⁶ U isotopes relative to ²³⁵ U by mixing in the liquid and / or gas phases of three components: the first is raw uranium regenerate hexafluoride; the second is natural uranium hexafluoride with a concentration of ²³⁵ U isotope from 0.15 to 0.7115%; and the third is natural uranium hexafluoride of natural origin containing 3.6-100% ²³⁵ U. Continuous addition and mixing of the components continues until the content of uranium isotopes is reduced in an economically feasible (acceptable) range of concentrations (patent No. 2110855 Russia, IPC 6 G 21 C 19/42), wt.%: ²³⁵ U 1-10; ²³² U 4.9 · 10 ^-7 - 3.8 · 10 ^-9 ; ²³⁴ U 1.7 · 10 ^-1 -7.3 · 10 ^-3 ; ²³⁶ U 8.0 · 10 ^-1 –6.4 · 10 ^-3 ; ²³⁸ U and other impurities - the rest. The relative deviation of the content from the nominal values is from -35 to + 35%. Uranium in the form of hexafluoride is mixed. Mixing in the liquid and / or gas phases ensures the maximum possible homogeneity of the nuclear fuel to be made from this hexafluoride.

The analogue method allows to reduce the content of harmful isotopes in the regenerated uranium to almost any level, however, the cost of such reactor fuel and the consumption of natural reducing agents will increase proportionally. In addition, the mass of fuel material exceeds the mass of raw uranium regenerate.

The closest in technical essence is the proposed method of isotopic reduction of regenerated uranium, which allows to stabilize the content of ²³² U and ²³⁶ U isotopes in reactor fuel and consists in diluting the raw uranium regenerate with natural uranium before the direct isotope enrichment operation in an ordinary isotope-separation cascade (ordinary cascade - cascade with a single supply stream) (Lebedev V.M. Fuel for nuclear power plants. Production and economics. - Obninsk: GTsIPK, 1996, p. 28). The required mass of natural diluent is at least 8-10 parts per 1 part of regenerated uranium with the required enrichment of ²³⁵ U isotope to 5-7%. With this dilution of the raw uranium regenerate, the concentration ratio of ²³⁵ U to ²³² U in the enriched product does not exceed (1.0-2.5) · 10 ^-7 , the absolute content of the ²³⁶ U isotope is not higher than 1.2-1.3%. This method is selected as a prototype.

Although uranium enrichment in the prototype is proposed to be carried out by gas diffusion or gas centrifugation methods, uranium enrichment is more cost-effective in cascades of gas centrifuges. The effective dose rate during gas centrifugal (gas centrifugal) enrichment is also significantly lower, because unlike gas diffusion enrichment, there are no repair operations for separation equipment, and the isotopic separation inside centrifuge rotors is carried out with low gas filling. The continuous operation resource of industrial gas centrifuges currently exceeds 15 years. In addition, the gas centrifuge enrichment technology makes it possible to direct uranium regenerates with ²³⁵ U isotope content below 0.85-0.95% for enrichment and to form various schemes for constructing isotope-separation cascades.

The disadvantages of the prototype method is that, firstly, to enrich the diluted raw uranium regenerate at the isotope-separation uranium plant, it is necessary to allocate 8-10 times greater separation power than in the case of enrichment of only regenerated uranium. Secondly, with each new cycle (multiple use of the same burned out mixture of uranium isotopes in LWR), raw uranium regenerate has to mix an increasing amount of natural uranium in order to introduce concentrations of ²³² U, ²³⁴ U and ²³⁶ U into the specification requirements. To enrich the increasing mass of diluted regenerated uranium in 2-3 cycles, the available separation capacities of uranium plants will not be enough. Thirdly, dilution is accompanied by a significant loss of separation work contained in the regenerated uranium, since the initial content of ²³⁵ U in the raw uranium regenerate is 70% higher than in the natural diluent. Fourth, in the prototype method there is no possibility of regulating the content of harmful isotopes during enrichment, their concentration only increases.

The present invention is aimed at solving the following problems:

- achieving the required enrichment of regenerated uranium with the ²³⁵ U isotope at a low (predetermined) content of harmful isotopes;

- improving the quality of recovery of regenerated uranium due to the absolute and relative regulated reduction in the content of harmful isotopes;

- minimization of separation capacities intended for isotope enrichment of regenerated uranium;

- optimization of the mass of regenerated uranium directed to the preparation of reactor fuel.

The above objectives are achieved by a technical solution, the essence of which is that in the method of isotope reduction of regenerated uranium in the fuel cycle, which consists in increasing the content of the ²³⁵ U isotope in regenerated uranium to 2.0-7.0% with a decrease in the absolute and relative concentration of isotopes ²³² U, ²³⁴ U and ²³⁶ U, comprising a direct enrichment of uranium hexafluoride feed gas centrifuge to reclaim isotope separation cascade, reclaimed raw uranium enriched in the isotope ²³⁵ U in a content of 21,0-90,0% in double askade the selection of the reduced fuel flow through the material of the second blade ordinary cascade.

The content of the ²³⁵ U isotope in the dump stream of the second ordinary cascade is 2.0-36.0%. The enrichment in the first ordinary cascade leads to the content of the isotope ²³⁵ U 3.5-10.0% or 21.0-36.0%.

In addition, the dump stream of the second ordinary cascade is diluted with natural uranium hexafluoride to a mass not exceeding the mass of raw uranium regenerate hexafluoride. In this case, dilution is carried out by uranium with an isotope content of ²³⁵ U 0.1-5.0%. Dilution is carried out at identical (close or equal) contents of the ²³⁵ U isotope in regenerated uranium and uranium of natural origin.

The main distinguishing feature of the method is the enrichment of the raw uranium regenerate with the ²³⁵ U isotope to a content of 21.0-90.0% in a double cascade when selecting reduced fuel material with a content of ²³⁵ U isotope not more than 50.0% through the dump stream of the second ordinary cascade (double cascade represents a system of two ordinary cascades in which the selection of the product of the first relative to the external power of the cascade serves as the power of the second cascade). With such enrichment values in the selection and dump, the second ordinary cascade performs the functions of a ²³² U and ²³⁴ U cascade purifying with respect to external power. Enriched regenerated uranium with the required or economically feasible (acceptable) relative concentrations of ²³² U and ²³⁴ U is obtained in the dump stream of the second ordinary cascade. The optimum concentration of ²³⁵ U in the dump stream of the second ordinary cascade is determined by the price of separation work and is in the range of 2.0-36. 0%

When enrichment in a double cascade to 21.0-90.0%, two optimal variants of enrichment of raw uranium regenerate in the first ordinary cascade are possible - up to 3.5-10.0% or 21.0-36.0% at ²³⁵ U. When Such enrichment in the first ordinary cascade purifies the raw uranium regenerate from the ²³⁶ U isotope, which goes to the dump of the cascade together with the ²³⁸ U isotope: in the first variant, 1.3-1.5 times, in the second variant, 1.5-2 times. That is, the first ordinary cascade performs the functions of a cleaning cascade with an adjustable cleaning coefficient for the ²³⁶ U isotope with respect to external power.

When enrichment in the dump stream of the second ordinary cascade exceeds 7.0%, enriched regenerated uranium is diluted with natural uranium to a mass not exceeding the mass of raw uranium regenerate hexafluoride. Dilution further reduces the relative content of harmful isotopes in the reduced fuel material. Depending on the amount of enrichment of regenerated uranium and the required dilution ratio, low enriched uranium with 0.711-5.0% ²³⁵ U, natural raw uranium or dump uranium with ²³⁵ U less than 0.711% (0.1-0.3%) are used for dilution, having referring also to the different isotopic composition of the natural diluent in these cases. The use of raw natural uranium saves the work of separation into the preparation of a natural diluent. The use of dump uranium provides additional savings in raw natural uranium. The choice of one or another natural diluent is based on customer requirements for the isotopic composition of reduced regenerated uranium, the costs of separation and raw natural uranium.

Dilution of low enriched regenerated uranium with low enriched natural uranium is carried out at an identical (close or equal) content of the ²³⁵ U isotope due to the requirements of minimizing the loss of separation work achieved by such a mixture.

Separate enrichment of raw uranium regenerate and natural uranium makes it possible to separate a relatively small double gas centrifuge cascade for enriching regenerated uranium in the isotope separation uranium plant, and all other plant capacities are still used in normal enrichment of natural uranium. The packing (desublimation) of enriched regenerated uranium in a transport container is combined with the operation of diluting natural origin with uranium, which minimizes the effective dose rate during subsequent handling of the containers. Naturally, in this case, the dilution occurs by mixing the regenerated uranium hexafluoride with natural uranium hexafluoride.

Figure 1-3 shows according to the proposed method, variants of block diagrams of a double isotope separation gas centrifuge cascade 1-2 for enrichment of raw uranium regenerate and ordinary cascade 3 for enrichment of natural diluent in accordance with their generally accepted designation (see: Sinev I.M ., Baturov BB Economics of nuclear energy: Fundamentals of technology and economics of nuclear fuel. - M.: Energoatomizdat, 1984, p.204, fig. 8.1, p.262-264, fig. 9.5-9.6). The flowcharts show the directions of supply flows 4, 5, flows 6, 7 and 8 of enriched no ²³⁵ U selection (selection) and flows 9, 10 and 11 of the depleted allotment (dump) along the ²³⁵ U isotope fractions.

In addition, figure 1-3 shows the installation location of the consumable washers 12 and 13 for controlled mixing of streams of enriched regenerated uranium and natural diluent, as well as containers 14, 15, 16 and 17, designed for packing uranium hexafluoride.

Raw uranium regenerate enters the uranium plant in the form of oxides. An additional purification from fission products is carried out at a uranium plant by radiochemical technology, after which the regenerated uranium is converted into the form of hexafluoride. Uranium hexafluoride containers are transported to a gas centrifuge isotope separation unit.

To supply 1-2 cogenerated hexafluoride of regenerated uranium to the double cascade, they are transferred to the gas phase (gasified) by heating the original containers above 323 K, and then they are sent to the power collector in the form of stream 4. At the output of the ordinary cascade 1, a selection stream 6 is received in the form of LEU-СОУ hexafluoride with 3.5-10.0% ²³⁵ U (see FIGS. 1 and 2), or SOU-HEU with 21.0-51.0% (mainly 21.0-36.0%) ²³⁵ U (cm Fig. 3), and the dump stream 9 containing the ²³⁸ U isotope with 0.1-0.3% ²³⁵ U and a different (predetermined) ²³⁶ U content.

The selection stream 6 is directed to the supply of an ordinary cascade 2 with the output of the selection stream 7 containing up to 21.0-90.0% ²³⁵ U, in which light isotopes ²³² U and ²³⁴ U are concentrated, as well as a dump stream 10 with 2.0 -36.0% ²³⁵ U, purified from ²³² U and ²³⁴ U.

At the same time, a natural diluent is being produced at the uranium plant according to the standard scheme. Uranium hexafluoride of natural origin is enriched in an ordinary cascade 3 with a supply stream 5 with 0.711% ²³⁵ U. At the cascade output, a selection stream 8 containing 2.0-5.0% ²³⁵ U and a dump stream 11 from 0.1-0 are obtained. 3% ²³⁵ U.

In the embodiment of the scheme of FIG. 1, the dump stream 10 from the cascade 2 by desublimation at 233-253K is packaged into a transport container 14 for delivery to the customer as recovered fuel material. In the variants of the schemes of FIGS. 2 and 3, the dump stream 10 from cascade 2 is preliminarily diluted with a sampling stream 8 from cascade 3. In the diagram of FIG. 2, the enriched regenerated uranium is diluted with enriched natural uranium at identical (equal or proximity) concentrations of ²³⁵ U in streams 10 and 8 selections. Figure 3 also shows the options for using, as a natural diluent, a feed stream 5 or a dump stream 11 from the cascade 3. The dilution ratio is maintained by consumable washers 12 and 13. After dilution, the uranium hexafluoride is packed into a transport container 14. As a result, reduced fuel material for LWR is obtained .

Dumps 9 and 11 from cascades 1 and 3 into containers, respectively, 15 and 17 are packaged in the same way, after which they are sent for temporary storage until the need for the ²³⁸ U isotope occurs.

Uranium hexafluoride with an isotopic mixture of stream 7, in which isotopes ²³² U and ²³⁴ U are concentrated, is packed into container 16 and sent for long-term storage until gamma activity decreases.

Due to pressure fluctuations of uranium hexafluoride during dilution according to the scheme of figure 2, the flow rate of gases during mixing may vary slightly. Therefore, to obtain the same concentration of ²³⁵ U in the volume, the contents of the container 14 are transferred to the liquid-phase state by heating in the autoclave 18 to a temperature above 337.17 K (up to 353 K at a pressure in the container of 238-243 kPa). Stirring of the melting uranium hexafluoride leads to equalization of the isotopic composition. This requirement is sometimes made by customers to recover fuel material.

Specific examples of the isotopic reduction of regenerated uranium are given in tables 2-5. Hereinafter in the tables: R is the amount of separation work spent on enriching uranium with the ²³⁵ U isotope, in units of thousand SWU; tails - ²³⁵ U content in cascade dump streams.

Reductive enrichment of raw uranium regenerate hexafluoride (RepU) in an amount of 100 tons of uranium with 0.95% ²³⁵ U was carried out in a double isotope-separation gas centrifuge cascade with ²³⁵ U enrichment in the first ordinary cascade to 5.0% (see Table 2 ), up to 10.0% (see table. 3), up to 21.0% (see table. 4) and 36.0% (see table. 5). The enrichment in the second ordinary cascade, the mass selection and dump streams were organized in such a way that the ²³² U isotope content in the selection stream was 1.00 · 10 ^-3 %, and the dump stream corresponded to the concentration of raw uranium regenerate.

In the first case, uranium hexafluoride from the dump stream of the second ordinary cascade was used directly as reduced fuel material. In other cases, it was additionally diluted with natural uranium hexafluoride, respectively, with an enrichment of 2.7%, raw uranium hexafluoride, and dump with 0.3% ²³⁵ U. The dilution ratio and the resulting mass of the recovered fuel material are shown in the tables.

In all cases, the recovered fuel material met the requirements of the specification (customer) and contained the isotopes ²³² U, ²³⁴ U and ²³⁶ U in relative (to fissile isotope ²³⁵ U) or absolute values less than the raw uranium regenerate. When diluting with uranium of natural origin of another enrichment, other, in principle any, concentrations of harmful isotopes are possible.

The mass of uranium hexafluoride, which concentrated 80-90 wt.% High background ²³² U, in all variants of isotopic reduction amounted to 14-16 kg and does not present any special problems during long-term storage.

The optimal isotope reduction option is selected based on the isotopic composition of the raw uranium regenerate, customer requirements, as well as the ratio of the prices of raw natural uranium and separation work.

The proposed method is suitable for the isotopic reduction of regenerated uranium during repeated recycling of spent nuclear fuel and for the production of fuel material with an extremely low content of harmful isotopes.

In all the considered options, the absolute need for a natural diluent for the recovery of 100 tons of regenerated uranium was significantly less than in the prototype.

It is understood that the invention is not limited to the examples given. There are other possible examples of examples within the scope of the proposed claims.

The proposed method for the isotopic reduction of regenerated uranium can be organized in existing isotope-separation uranium plants and provides a fuel material at the outlet of the plant while minimizing separation capacity and the absence or minimum consumption of natural uranium diluent, the content of harmful isotopes in which is in the required concentrations, regardless of the burnup depth initial SNF. Quantitative purification from the high-level ²³² U isotope and the absence of degradation of the raw uranium regenerate as reactor fuel as a result of partial purification from the ²³⁶ U isotope significantly saves the consumption of natural uranium in the fuel cycle of light-water reactors.

Claims (7)

1. The method of isotope reduction of regenerated uranium, which consists in increasing the content of the ²³⁵ U isotope in regenerated uranium to 2.0-7.0 wt.% While reducing the absolute and relative concentration of ²³² U, ²³⁴ U and ²³⁶ U isotopes, including direct enrichment of raw hexafluoride uranium regenerate in a gas centrifuge isotope separation cascade, characterized in that the raw uranium regenerate is enriched in the ²³⁵ U isotope to a content of 21.0-90.0 wt.% in a double cascade when selecting reduced fuel material with an isotope content of ²³⁵ U not more than 50.0 wt.% Through the dump stream of the second ordinary cascade. 2. The method according to claim 1, characterized in that the content of the ²³⁵ U isotope in the dump stream of the second ordinary cascade is 2.0-36.0 wt.%. 3. The method according to claim 1 or 2, characterized in that the enrichment in the first ordinary cascade is carried out to a content of ²³⁵ U of 3.5-10.0 wt.%. 4. The method according to claim 1 or 2, characterized in that the enrichment in the first ordinary cascade is carried out to a content of ²³⁵ U of 21.0-36.0 wt.%. 5. The method according to claim 1, characterized in that the dump stream of the second ordinary cascade is diluted with natural uranium hexafluoride to a mass not exceeding the mass of raw uranium regenerate hexafluoride. 6. The method according to claim 5, characterized in that the dilution is carried out by uranium with an isotope content of ²³⁵ U 0.1-5.0 wt.%. 7. The method according to claim 5 or 6, characterized in that the dilution is carried out at an identical (close or equal) content of the ²³⁵ U isotope in the regenerated uranium and uranium of natural origin.

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