EDN: EPGZKX
Authors & Affiliations
Romadinov A. M., Fedorov I. A., Yufereva V. A.
N.A. Dollezhal Research and Development Institute of Power Engineering, Moscow, Russia
Romadinov A.M. – Engineer 2nd Category. Сontacts: 1, bldg. 3, pl. Akademika Dollezhalya, Moscow, Russia, 107140. Tel.: +7 (919) 963-41-33; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Fedorov I.A. – Engineer 2nd Category.
Yufereva V.А. – Head of Group.
Abstract
Currently, R&D is underway in respect of the design of the fast reactor with a heavy liquid metal coolant and mixed nitride uranium-plutonium fuel with the implementation of a closed nuclear fuel cycle. To justify the safety of using the nitride fuel, verify codes, establish a balance of fission materials and fission products, accumulate potentially radiation-hazardous nuclides, appropriate experimental studies are required. In order to verify the MCU-NR precision code in terms of isotope kinetics and nuclear fuel burnup, computational simulation of irradiation of oxide nuclear fuel samples in the capillary tube in the BN-350 reactor and samples of mixed nitride uranium-plutonium fuel in the combined experimental fuel assembly in the BN-600 reactor was carried out. To justify the possibility of transferring the results of the error estimation of the concentration of actinides during the burnup of the mixed nitride fuel in the core in the spectrum of the BN-600 reactor (combined experimental fuel assembly, CEFA) to the lead-cooled reactor, a computational test of the burnup of CEFA fuel elements in the BREST-OD-300 reactor was developed. A computational study was carried out to justify the possibility of using a cell model in calculating burnup and isotope kinetics in fast reactors. A comparative analysis of the results of calculating the burnup of nuclear fuel in a full-scale model of a fast reactor with a heavy liquid metal coolant and mixed nitride uranium-plutonium fuel and in the corresponding cell models was carried out. In the course of computational simulation of irradiation of nuclear fuel samples in fast reactors at rated power, maximum errors of calculation of nuclide composition change using the MCU-NR code were determined.
Keywords
computational modeling, irradiation experiments, verification, MNIT fuel, MOX fuel, isotopic kinetics, nuclear fuel burnup, precision code, Monte Carlo method
Article Text (PDF, in Russian)
References
- Adamov E.O., Dzhalavyan A.V., Lopatkin A.V., Molokanov N.A., Muravyev E.V., Orlov V.V., Kalyakin S.G., Rachkov V.I., Troyanov V.M., Avronin E.N., Ivanov V.B., Aleksakhin R.M. Conceptual framework of a strategy for the development of nuclear power in Russia to 2100. At Energy, 2012, vol. 112, issue 6, pp. 391–403. DOI: https://doi.org/10.1007/s10512-012-9574-x.
- Troyanov M.F. The BN-350 reactor: development, design and operating experience. Proc. of the International conference on nuclear power and its fuel cycle. Vienna, May 2–13, 1997. Report IAEA-CN-36/594, vol. 1, pp. 775–787.
- Vasilyev B.A., Kryukov A.N., Farakshin M.R., Belov S.B., Sheryakov V.S., Kuznetsov A.E., Filin I.A. Ispol'zovanie MOKS-topliva v reaktore BN-800 [Use of MOX Fuel in BN-800 Reactor]. Izvestiya vuzov. Yadernaya Energetika. 2025, no. 3, pp. 45–59. DOI: https://doi.org/10.26583/npe.2025.3.03.
- Andrushechko S.A., Afrov A.M., Vasilyev B.Yu., Generalov V.N., Kosourov K.B., Semchenkov Yu.M., Ukraincev V.F. AES s reaktorom tipa VVER-1000. Ot fizicheskikh osnov ekspluatatsii do evolyutsii proekta [Nuclear power plant with VVER-1000 reactor. From fundamental operational physics to evolution of the project]. Moscow, Logos Publ., 2010. 604 p.
- Gurevich M.I., Shkarovsky D.A. Raschet perenosa neytronov metodom Monte-Karlo po programme MCU. Uchebnoe posobie [Calculation of neutron transport by the Monte Carlo method using the MCU Program. Textbook]. Moscow, NRNU MEPhI Publ., 2012. 156 p.
- Kahler A.C., MacFarlane R.E., Mosteller R.D., Kiedrowski B.C., Frankle S.C., Chadwick M.B., McKnight R.D., Lell R.M., Palmiotti G., Hiruta H., Herman M., Arcilla R., Mughabghab S.F., Sublet J.C., Trkov A., Trumbull T.H., Dunn M. ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments. Nuclear Data Sheets, 2011, vol. 112, issue 12, pp. 2997–3036.
- FACT-BR (versiya 1.1). Certification passport of the software no. 433 dated 02.27.2018.
- Nuclide kinetics, radioactivity and decay heat code. Version 2.1 (BPSD/V2.1). Certification passport of the software no. 516 dated 30.01.2021.
- Seleznev E.F. Kinetika reaktorov na bystrykh neytronakh. Pod red. akad. RAN A.A. Sarkisova [Fast reactor kinetics. Ed. Academician of the RAS Sarkisov A.A.]. Moscow, Nauka Publ., 2013. Pp. 178–198.
- Manturov G.N., Nikolaev M.N., Polyakov A.Yu., Tsibulya A.M. Notes on the CONSYST code. Voprosy atomnoy nauki i tekhniki. Seriya: Yadernye konstanty – Problems of Atomic Science and Technology. Series: Nuclear Constants, 1999, issue 2. Available at: https://vant.ippe.ru/images/pdf/1999/2-15-en.pdf (accessed 12.01.2026).
- Oshkankov N.N., Potapov O.A., Govorov P.P. Otsenka effektivnosti raboty energobloka s reaktorom na bystrykh neytronakh BN-600 Beloyarskoy AES [Operational efficiency evaluation of the BN-600 sodium-cooled fast reactor]. Izvestiya vuzov. Yadernaya Energetika, 2005, no. 1, pp. 3–9.
- Alekseev S.V., Zaytsev V.A. Nitridnoe toplivo dlya yadernoy energetiki [Nitride fuel for atomic energy]. Moscow, Tekhnosfera Publ., 2013.
- Grachev A.F., Zabudko L.M., Zvir E.A., Ivanov Yu.A., Krukov F.N., Marinenko E.E., Nikitin O.N., Porollo S.I., Skupov M.V. The results of the research of the BN-600 fuel elements with mixed nitride uranium-plutonium fuel and ChS68-ID steel cladding. At Energy, 2019, vol. 126, issue 3, pp. 182–190. DOI: https://doi.org/10.1007/s10512-019-00534-0.
- Momotov V.N., Volkov A.Yu., Kupriyanov V.N., Khamdeev M.I., Tikhonova D.E., Shadrin A.Yu., Khomyakov Yu.S.. Radiokhimicheskiy i elementnyy analiz smeshannogo uran-plutonievogo topliva, obluchennogo v reaktore BN-600 [Radiochemical and elemental analysis of mixed uranium-plutonium fuel irradiated in the reactor]. Radiokhimiya – Radiochemistry, 2022, vol. 64, no. 1, pp. 53–59.
- Adamov E.O., Kaplienko A.V., Orlov V.V., Smirnov V.S., Lopatkin A.V., Lemekhov V.V., Moiseev A.V. Brest Lead-Cooled Fast Reactor: From Concept to Technological Implementation. At Energy, 2021, vol. 129, issue 4, pp. 179–187. DOI: https://doi.org/10.1007/s10512-021-00731-w.
UDC 621.039.51
Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2026, no. 1, 1:8
