RU2261493C1 - Method for building-up americium-242m nuclei in fast reactor radiator and irradiating device for building-up 242-m americium nuclei - Google Patents

Abstract

FIELD: nuclear engineering.

SUBSTANCE: proposed method for building-up 242m-americium nuclei in fast reactor reflector includes placement of americium-241 nuclei containing material in reflector and its irradiation with neutrons. Neutron spectrum is formed by sequential passage of neutrons through neutron moderator and neutron filter made of material absorbing thermal neutrons. Used as neutron moderator is beryllium oxide or boron carbide enriched with respect to boron-11 isotope, or metal hydrides; used as substances absorbing thermal neutrons are gadolinium or cadmium. Irradiating device for building-up americium-242m nuclei in fast reactor reflector has at least one irradiation assembly provided with neutron moderator with at least one coolant channel accommodating container with target made of starting material. This container is placed in filter made of thermal neutron absorbing material; used as starting material is that having americium-241 nuclei in its composition.

EFFECT: enhanced building-up efficiency.

16 cl, 5 dwg

Description

The invention relates to the field of nuclear engineering and can be used in fast-neutron nuclear reactors for the production of americium-242m nuclei.

The interest in americium-242m is due to the very small critical mass of this isotope (for reactors with a thermal spectrum - only a few tens of grams), which makes it possible to create a compact nuclear reactor based on this isotope, which can be used as a neutron source, in particular, in neutron capture therapy.

There is a known method for producing americium-242m ( ^242m Am) nuclei in the reflector of a fast neutron reactor (Y. Ronen, M. Aboudy, D. Regev "Breeding of 242m Am in Fast Reactor". Proceedings of the 21 ^st Conference of the Nuclear Societies in Israel. May 22-23, 2002. Haifa, Israel), selected as a prototype. The known method consists in the fact that instead of the usual uranium compounds, approximately 10,000 kg of americium-241 dioxide ( ²⁴¹ AmO ₂ ) are loaded into the reflector of a fast reactor and, as a result of radiation capture of neutrons by americium-241 nuclei ( ²⁴¹ Am), ^242m Am nuclei are formed.

The disadvantage of this method is the very ^low efficiency of the production of ^242m Am cores (for 180 days of reactor operation, approximately 0.07% of the number of starting material cores). The second disadvantage is the low concentration of this isotope in irradiated america and, as a consequence, the high cost of the produced material, which includes the cost of the starting material, the cost of enriching the irradiated material to the desired concentrations of ^242m Am and the cost of environmental protection as the main components. Moreover, it is impossible to solve this problem by increasing the time of irradiation of the starting material, because, firstly, ^242m Am burns ^out in parallel with the operating time and, secondly, with an increase in the time of irradiation of the starting material, the residual energy release of the irradiated material increases significantly, which makes it is practically impossible to process it for enrichment in an industrial environment without additional temporary exposure of irradiated americium.

A known irradiation device for producing radioactive isotopes in the reflector of a fast neutron reactor (RF Patent No. 2076362 "Method for producing radioactive isotopes in a fast neutron reactor and a nuclear fast neutron reactor". IPC G 21 G 1/02), selected as a prototype. This irradiating device has the external geometric dimensions of the fuel assembly (FA) of the fast reactor and is equipped with a neutron moderator and at least one channel for the coolant flow, as well as a target from the starting material placed inside the neutron moderator.

A disadvantage of the known irradiation device is that as the operating time occurs, the burned-out ^242m Am intensively burns out due to the very large cross-section of fission of ^242m Am by thermal neutrons, which makes the industrial operating time of ^242m Am practically impossible.

The technical problem solved by the invention was to create a method and an irradiation device for efficiently operating ^242m Am on an industrial scale with minimal material and financial costs.

The essence of the invention in relation to the method of operating the americium-242m nuclei in the reflector of a fast reactor consists in placing a substance containing americium-241 nuclei in the reflector and its irradiation with neutrons. The difference of the proposed method is that the neutron spectrum is formed by sequential transmission of neutrons through a neutron moderator and a neutron filter made of a substance that absorbs thermal neutrons.

In particular, substances containing beryllium (Be) nuclei are used as a neutron moderator, in particular, beryllium oxide (BeO) can be this substance.

In addition, boron carbide (B ₄ C) enriched in the boron-11 isotope ( ¹¹ V) is used as a neutron moderator. In particular, enrichment by ¹¹ V isotope is at least 99%.

In addition, metal hydrides are used as a neutron moderator.

In particular, gadolinium (Gd) is used as a neutron-absorbing substance, in particular gadolinium-157 ( ¹⁵⁷ Gd) enriched in the isotope to a value of at least 70%.

In addition, cadmium (Cd) is used as a neutron absorbing substance.

The invention with respect to an irradiation device for producing americium-242m nuclei in a fast reactor reflector is that it includes at least one irradiation assembly equipped with a neutron moderator with at least one channel for the coolant flow, and also equipped with a target from the starting material, which is placed inside the neutron moderator.

The difference between the irradiation device is that the target from the starting material is placed inside the filter made of a material absorbing thermal neutrons, and the material containing americium-241 nuclei is used as the starting material.

In particular, the neutron moderator can be made of a material containing beryllium nuclei, in particular, beryllium oxide can be used as such a material.

In particular, the neutron moderator can be made of boron carbide enriched in the boron-11 isotope, while enrichment can be at least 99%. In particular, the neutron moderator can be made of metal hydride. In particular, the filter can be made of gadolinium, which can be enriched by the isotope gadolinium-157, and the enrichment can be at least 70%.

In particular, the filter may be made of cadmium. In particular, the filter thickness is in the range from 0.005 to 0.12 cm. In addition, the thickness of the neutron moderator between the outer side surface of the irradiation assembly and the outer side surface of the filter is in the range from 1.5 to 4.7 cm.

In addition, the mass of americium-241 nuclei in the target does not exceed 110 g. The technical result of using this invention is to increase the efficiency of the production of ^242m Am nuclei, in which it is possible to produce up to 50 grams of ^242m Am from 1 kg ^241m Am for approximately one microcompany of the BN-600 reactor (160-180 effective days). Or, by irradiating the starting material for two to three micro- ^companies, it is possible to increase the content of ^242m Am in the Am isotopic composition to 13-14%, and the residual heat release of the irradiated target material will allow it to work with standard equipment after temporary exposure typical for irradiated fuel assemblies fast reactor.

The achievement of the technical result became possible due to the fact that the set of essential features of the invention provides the formation of a special spectrum of epithermal neutrons from the neutron spectrum characteristic of a fast reactor reflector, which allows you to produce ^242m Am at a speed significantly higher than its burnout rate.

Figure 1 shows a diagram of a cartogram loading a fast reactor, which schematically shows the active zone 1, reflector 2 of the reactor, fuel assembly of the first row of reflector 3, irradiation assembly 4.

Figure 2 shows a cross-sectional diagram of an irradiation assembly 4 with a neutron moderator 5, with a channel 6 for passing the coolant, in the inner region of which a container with a target 7 of starting material is placed, which in turn is located inside the filter 8 absorbing thermal neutrons.

Figure 3 shows the graphs of the operating time of ^242m Am in irradiation assemblies with various types of neutron moderator and thermal neutron absorber depending on the exposure time. Curve 9 corresponds to an irradiation assembly with boron carbide as a neutron moderator and ¹⁵⁷ Gd as a thermal neutron absorber; curve 10, respectively, with zirconium hydride and ¹⁵⁷ Gd; curve 11, respectively, with boron carbide and gadolinium of a natural isotopic composition; curve 12, respectively, with boron carbide and cadmium.

Figure 4 shows graphs of the fraction of ^242m Am in irradiated Am in irradiation assemblies with various types of neutron moderator and thermal neutron absorber versus irradiation time. Curve 13 corresponds to an irradiation assembly with boron carbide as a neutron moderator and ¹⁵⁷ Gd as a thermal neutron absorber; curve 14, respectively, with zirconium hydride and ¹⁵⁷ Gd; curve 15, respectively, with boron carbide and gadolinium of natural isotopic composition; curve 16, respectively, with boron carbide and cadmium.

Figure 5 shows graphs of the residual energy release of fuel assemblies, spent in the medium enrichment zone of the BN-600 reactor, and the irradiation assembly after irradiation, depending on the exposure time. Curve 17 corresponds to a fuel assembly, and curve 18 to an irradiation assembly with a neutron moderator of ¹¹ V ₄ C and a neutron absorber of ¹⁵⁷ Gd.

The method is carried out, and the device operates as follows. In the fast neutron reactor, among the standard fuel assemblies of the second row of the reflector 2, the irradiation assemblies 4 are placed. During operation of the reactor, neutrons with the energy spectrum characteristic of the fuel assemblies of the second row of the reactor reflector fall into the irradiation assembly 4, slow down in the moderator layer 5, and pass through the filter 8 absorbing thermal neutrons, and hit the target 7, where they are captured by ^242m Am nuclei with the emission of a gamma quantum, as a result of which ^242m Am nuclei are formed.

The irradiation assembly 4 has the external geometric dimensions of the fuel assembly of the fast reactor reflector, inside which a neutron moderator 5 is placed. Inside the neutron moderator 5, a channel 6 is made for the passage of the coolant. In the space between the shell of the irradiation assembly and the neutron moderator, an additional channel for the passage of the coolant can be located. Channel 6 contains a container with a target 7 of starting material. The container is placed in the filter from a material that absorbs thermal neutrons.

As a neutron moderator can be used: beryllium; beryllium oxide; boron carbide with a natural isotopic composition of boron, as well as with boron enriched in the isotope boron-11, and enrichment in boron-11 can be at least 99%; metal hydrides, for example zirconium hydride. The thickness of the neutron moderator layer can range from 1.5 to 4.7 cm.

As a material that absorbs thermal neutrons, the following can be used: gadolinium of natural isotopic composition and gadolinium-157 enriched in the isotope, and gadolinium-157 enrichment can exceed 70%; cadmium. Filter thickness can range from 0.005 cm to 1.2 cm.

Am ₂ O ₃ or AmO ₂ diluted with an inert material can be used as starting material. The amount of americium-241 in one irradiation assembly, as a rule, should not exceed 110 g.

In cases where it is necessary to place the irradiation assembly in the first row of the reflector fuel assembly or to minimize the neutron field distortion of the reactor core near the irradiation assembly, an additional thermal neutron absorber is placed in the space between the irradiation assembly housing and the neutron moderator. An additional neutron absorber practically does not affect the penetration of the reactor neutron spectrum into the irradiation assembly, but absorbs neutrons that have slowed down and exit the irradiation assembly.

In cases where it is necessary to optimize the neutron field of the irradiation assembly located in the second row of the reflector fuel assembly, it is surrounded by assemblies similar to the irradiation assembly, in which starting material may be absent.

The industrial applicability of the invention is confirmed by the results of model-computational studies. As model and computational studies show on the example of the BN-600 fast neutron reactor, the application of the proposed invention allows arranging ^242m Am in this reactor in the amount of hundreds of grams for one or two micro-companies with a high content (more than 10%) of the required isotope in America.

The model of the BN-600 reactor in the mid-stationary state was adopted as the basis for the calculation model. Based on it, calculations were performed with an irradiation device (OS), which was located in the second row of the lateral reproduction zone.

The neutron-physical characteristics were calculated using the TRIGEX program (Seregin A.S., Kislitsyna T.S. Annotation of the TRIGEX-CONSYST-BNAB-90 program package // Preprint FEI - 2655. 1997) in the 26-group approximation taking into account burnup only Shelter zones. Since the TRIGEX calculations use a simplified version of burnout, the CARE program was used separately (A. Kochetkov, “The CARE program for calculating the isotope kinetics, radiation and environmental characteristics of nuclear fuel during its irradiation and aging” // Preprint FEI 2431. 1995) (ORIGEN-S / OWHermann. RMWestfall. ORIGEN-S: SCALE system module to calculate fuel depletion, actinide transmutation, fission product buldup and decay, and Association source terms // NUREG / CR-0200. Revision 4. Vol.2 . Section F7. 1995). The calculation procedure was as follows: "TRIGEX - CARE -.... CARE". From TRIGEX calculations, 26-group flows and blocked cross sections were fed to CARE, where new actinide concentrations for the next time step were calculated, which were again fed to TRIGEX. For those isotopes that are not blocked, single-group cross sections were obtained by averaging 26-group flows on the data of the ABBN-93 library (Manturov G.N., Nikolaev M.N. Tsibulya A.M. System of group constants BNAB-93. Part 1. Nuclear constants for calculating neutron and photon radiation fields. // Questions of Atomic Science and Technology. Ser. Nuclear constants, issue 1. 1996).

The results of model studies are shown in figures 3-5. From them follows:

1. ZrH ₂ , as a moderating material, compared to ¹¹ V ₄ C, strongly “eats up” the neutron flux in the target region and ^moreover, it takes more time to irradiate to achieve an optimum operating time of ^242m Am.

2. Cd plays Gd like a heat group filter. It slows down not only the thermal group, but also suprathermal groups, which reduces the operating time of ^242m Am. The best of the gadolinium isotopes is ¹⁵⁷ Gd.

3. In the second row of the inner reproduction zone, using an op-amp with a moderator of ZrH ₂ and ¹¹ V ₄ C and a filter of ¹⁵⁷ Gd, it is possible to produce 20-30 grams of ^242m Am of 1 kg of ²⁴¹ Am. The content of ^242m Am in the isotopic composition of Am will be 13-14%.

4. When no more than 110 g is placed in the target starting material in terms of the ²⁴¹ Am isotope, the residual energy release of the irradiated target does not exceed the residual energy release of the spent fuel assembly of the secondary enrichment zone.

Claims (16)

1. A method for producing americium-242m nuclei in a reflector of a fast reactor, comprising placing a substance containing americium-241 nuclei in a reflector and irradiating it with neutrons, characterized in that the neutron spectrum is formed by sequential transmission of neutrons through a neutron moderator and a neutron filter made of matter absorbing thermal neutrons, while beryllium oxide, or boron carbide enriched in the boron-11 isotope, or metal hydrides are used as a neutron moderator, and as substances that absorb lovye neutrons using gadolinium or cadmium. 2. The method according to claim 1, characterized in that the enrichment of boron by boron-11 isotope is at least 99%. 3. The method according to claim 1, characterized in that gadolinium is used enriched in the gadolinium-157 isotope to a value of at least 70%. 4. An irradiating device for producing americium-242m nuclei in a fast reactor reflector, comprising at least one irradiation assembly equipped with a neutron moderator with at least one channel for the coolant flow with a container with a target from the starting material inside it, characterized in that in the channel for the flow of coolant, a container with a target from the starting material is placed in a filter made of a material that absorbs thermal neutrons, and used as a starting material material containing americium-241 nuclei. 5. The irradiation device according to claim 4, characterized in that the neutron moderator is made of a material containing beryllium nuclei. 6. The irradiation device according to claim 5, characterized in that beryllium oxide is used as the material containing beryllium nuclei. 7. The irradiation device according to claim 4, characterized in that the neutron moderator is made of boron carbide enriched in the boron-11 isotope. 8. The irradiation device according to claim 7, characterized in that the enrichment of boron by boron-11 isotope is at least 99%. 9. The irradiation device according to claim 4, characterized in that the neutron moderator is made of metal hydride. 10. The irradiation device according to claim 4, characterized in that the filter is made of gadolinium. 11. The irradiation device according to claim 10, characterized in that the gadolinium is enriched in the isotope gadolinium-157. 12. The irradiation device according to claim 11, characterized in that the gadolinium-157 isotope enrichment is at least 70%. 13. The irradiation device according to claim 4, characterized in that the filter is made of cadmium. 14. The irradiation device according to claim 4, characterized in that the thickness of the neutron moderator is in the range of 1.5 - 4.7 cm 15. The irradiation device according to claim 4, characterized in that the thickness of the filter is in the range of 0.005 - 0.12 cm 16. The irradiation device according to claim 4, characterized in that the mass of americium-241 nuclei in the target does not exceed 110 g.

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