Journal Issue: Nuclear Energy and Technology (NUCET)
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Volume
2025-11
Number
1
Issue Date
Journal Title
Nuclear Energy and Technology (NUCET)
Journal ISSN
2452-3038
Том журнала
Том журнала
Nuclear Energy and Technology (NUCET)
Nuclear Energy and Technology (NUCET) (2025-11)
Статьи
Публикация
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Assessing the suitability of two-group cross-sections and diffusion coefficients derived from SERPENT-2 for small modular reactor ACP-100
(НИЯУ МИФИ, 2025) Md. Abidur Rahman Ishraq; Kruglikov, A. E.
The focus of this work is to analyse the suitability of two-group diffusion coefficients and macro constants generated from SERPENT using out-scattering approximation (OSA), transport correction (TRC) and cumulative migration methods (CMM) for fuel and non-fuel materials. For this purpose, various assembly and core models of ACP-100 SMR were designed. Assessment of these constants was conducted using COMSOL Multiphysics. For six distinct fuels, the best models were proposed with the least error margin in keff. Fuel material affects the group constants of non-fuel components except for radial reflectors. The sufficiency of two-group calculation was justified through spectrum analysis. Additional analysis revealed that MOX-RG has the hardest spectrum among all the fuels. Moreover, the effectiveness of boric acid to control excess reactivity was observed. Subcriticality was achieved for all fuel types except MOX-RG at a boric acid concentration of 4500 ppm. The influence of variation of boric acid concentrations on group constants was investigated using TRC and OSA. The reactivity difference between SERPENT and COMSOL was determined. It was found that OSA generates the most accurate results for MOX-RG with maximum 863 pcm error, while TRC produces higher accuracy with maximum error of approximately 250 pcm for other fuels.
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Neutronic peculiarities of the MTRR-SCW reactor as an experimental base for testing advanced light-water reactor technologies
(НИЯУ МИФИ, 2025) Lapin, A. S.; Blandinsky, V. Yu.; Nevinitsa, V. A.; Pustovalov, S. B.; Sedov, A. A.; Subbotin, S. S.; Fomichenko, P. A.
The nuclear power system has faced a challenging issue of significantly improving the characteristics of nuclear fuel breeding while maximizing the advantages of the technology of vessel-type pressurized water reactors used extensively in nuclear power. This is possible via switching to supercritical coolant parameters. An increase in production of fissile nuclides, as compared with traditional pressurized water reactors, is achieved by switching to a harder neutron spectrum due to reducing greatly the coolant density and using a dense fuel lattice. A necessary condition for the VVER-SKD design development is the establishment of an experimental base. A multipurpose test research reactor, MTRR-SCW, is the potential testing ground for the reactor technology, and for new structural and fuel materials and fuel rods. The paper presents the key characteristics of the MTRR-SCW reactor, as well as the potential MTRR-SCW applications at different stages of its operation (testing and research). A concept is proposed at the initial stage for the reactor phased rise to power, which will make it possible to justify the efficiency of the MTRR-SCW fuel with increased linear loads through experiments in the independent central loop channel. This concept also involves phased validation and study of the joint operation of the reactor plant and the steam turbine plant as part of the MTRR-SCW nuclear power plant. At the research stage of operation, safe operating limits need to be determined, and the choice of normal operating modes for the VVER-SKD power reactor justified, and experimental studies need to be undertaken to investigate the behavior of structural materials and fuel compositions as part of experimental fuel rods for the advanced light-water reactor cores with different neutron spectra. Long-term irradiation of experimental fuel rods is planned to be carried out in the MTRR-SCW’s independent peripheral loop channel, and experimental simulation of emergency processes to be performed in the reactor’s central loop channel. This paper deals with the issues to be addressed prior to starting the VVER-SKD power reactor design. Issues have been identified that can be fully or partially solved at effective facilities, as well as the applications for the MTRR-SCW prototype reactor. The paper presents the key characteristics of the MTRR-SCW reactor, and describes in detail the concept for the phased development of the research reactor capabilities and phased rise to power.
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Closed nuclear fuel cycle of thermal and fast reactors with fuel self-sufficiency
(НИЯУ МИФИ, 2025) Shmelev, A. N.; Apse, V. A.; Kulikov, E. G.; Kulikov, G. G.; Glebov, V. B.; Глебов, Василий Борисович; Куликов, Евгений Геннадьевич; Куликов, Геннадий Генрихович; Апсэ, Владимир Александрович
The paper presents the results obtained in numerical evaluations of a possibility to reach self-sufficiency of fissile materials in the joint system of fast and thermal reactors. These studies considered the joint system consisting of thermal light-water reactors of VVER-type and fast lead-cooled reactors of BREST-type, which are operated within the frames of the closed (Th-U-Pu) fuel cycle. It was assumed that fast reactors (FR) of BREST-type used the mixed thorium-plutonium nitride fuel, while thermal reactors (TR) of VVER-type used the mixed oxide fuel of natural uranium and 233U. Uranium isotope 233U was produced in the Th-fraction of FR fuel for further introduction into the fresh composition of TR fuel, while plutonium was produced in the natural uranium fraction of TR fuel for further introduction into the fresh composition of FR fuel. The numerical studies resulted in the determination of the conditions necessary to provide fuel self-sufficiency in the joint TR-FR system. The following key findings of the research may be noted:
– It is demonstrated the possibility to create the joint TR-FR system with inherent fuel self-sufficiency;
– Involvement of thorium and 233U in the closed NFC of the joint TR-FR system can arrange an optimal regime for production and consumption of main FM;
– Purposeful change of thermal power and introduction of natural uranium in the fuel composition of BREST-type FR made it possible to reach fuel self-sufficiency of the joint TR-FR system;
– Application of radiogenic lead instead of natural lead allowed us to reduce necessary values of thermal power and content of natural uranium nitride in the fuel of BREST-type FR.
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Tungsten wires with steel matrix composite: wetting and infiltration by steel melt
(НИЯУ МИФИ, 2025) Popov, N.; Suchkov, A.; Zharkov, M.; Kirillova, V.; Bazhenov, A.; Dzhumaev, P.; Vertkov, A.; Sevryukov, O.
Lithium-based capillary porous systems (CPS) made of tungsten mesh are a part of the most prospective approach to plasma facing components. Currently, tungsten mesh as a part of the CPS is mounted directly on the experimental assembly without a proper joining with a substrate. Tungsten mesh filled with steel could be used as a base structure for the CPS. The experiment considers the wetting of tungsten by steel melt and the features of short- and long-term interaction between the two materials. Wetting was studied by improved sessile drop experiment. The results show that an average contact angle is 69° for SS316LN and 83.2° for SS420 melt on tungsten substrate with a temperature of 500–650 °C. Tungsten-steel composite was manufactured by infiltration of tungsten mesh with a steel melt. As a result of an active dissolution of tungsten in steel melt, (Fe,Cr)7W6 interaction layer with a thickness up to 10 µm forms around tungsten. Optimal structure with the thinnest intermetallic layer is obtained in the zones with the lowest temperature and the highest cooling speed.
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Radiation characteristics of reactor grade platinum group metals
(НИЯУ МИФИ, 2025) Kovalev, N. V.; Prokoshin, A. M.; Davydova, P. V.; Korolev, V. A.
The paper examines the radiation characteristics of noble platinum group metals (PGMs) extracted from spent nuclear fuel (SNF) of the VVER-1000 reactor. These are the so-called reactor-grade ruthenium, rhodium and palladium. PGMs are radioactive when extracted from SNF, but after the decay cooling of ruthenium for about 27 years, and of rhodium for about 13 years, they can be used in unlimited quantities. There is no sense in decay cooling of reactor grade palladium due to its radioactive isotope 107Pd having a half-life of 6.5 million years. As specified by regulatory documents, such palladium can be freely used only in quantities of up to 34 g. Pd is a soft beta emitter with a maximum beta particle energy of 34 keV. The calculation results show that the mean free path of beta particles from 107Pd in palladium metal is 0.8 μm, so reactor-grade palladium emits only from the surface layer, and other electrons are absorbed in the material itself. The mean free path of electrons with an energy of 34 keV in biological tissue is about 20 μm, which does not exceed the thickness of the skin epidermiscorneous layer. Calculations have shown that the equivalent dose rate (EDR) on the surface of reactor-grade palladium is 0.04 μSv/h, which is below the public EDR value. As a result, a conclusion is made that reactor-grade palladium does not pose a danger in the event of external contact.