Authors & Affiliations
Bolnov V.A., Bogdanova E.V., Malkin S.A.
Afrikantov OKBM JSC, Nizhny Novgorod, Russia
Bolnov V.A. – Head of Department.
Bogdanova E .V. – Head of Design Group.
Malkin S .A. – Lead Design Engineer, Cand. Sci. (Tech.). Contacts: 115 Burnakovsky proyezd, Nizhny Novgorod, Russia, 603074. Tel.: +7 (987) 111-54-93; e-mail:
The BURAN program is used to calculate the normal and emergency modes of BN-type installations at JSC Afrikantov OKBM. In 2021, her attestation passport expired. The program was upgraded, after which it was submitted for recertification to the SEC NRS. In the report on the justification of the applicability of the BURAN program, all available experimental data were used. This article presents the results of validation of the BURAN program by comparing the calculation results with experimental data obtained only at the BN-600 reactor plant. These are modes with a planned change in power, shutdown of the power unit at a power of 95 % Nnom, complex testing of the emergency cooling system and removal of residual heat due to heat losses from the reactor pressure vessel and from the secondary circuit in the absence of circulation in the third circuit. Based on the results of comparison of calculated and experimental data, the possibility of using the BURAN program for calculating transient processes in installations with fast neutron reactors with sodium coolant in normal operation modes, in modes with disruption of normal operation, requiring an emergency transfer of the installation to reduced power levels, is substantiated. Transient processes can be with long-term cooling down due to heat losses from the reactor vessel, from equipment and pipelines of the secondary circuit, with a complete failure of forced heat removal systems.
validation, program, BURAN, reactor facilities, fast, BN-600, experiment, dynamics, temperature, deviations, results
1. Buksha Yu.K., Bagdassarov Yu.E., Kiryushin A.I. et al. Operation experience of the BN-600 fast reactor. Nuclear Engineering and Design, 1997, vol. 173, pp. 67–79.
2. Klimonov I.A., Usov E.V., Dugarov G.A. et al. HYDRA-IBRAE/LM/V1 Thermohydraulic Code Verification Based on BN-600 Experiments. Atomic Energy, 2017, vol. 122, pp. 311–318. DOI: https://doi.org/10.1007/s10512-017-0272-6.
3. Alipchenkov V.M., Boldyrev A.V., Veprev D.P. et al. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification. Thermal Engineering, 2018, vol. 65, pp. 627–640. DOI: https://doi.org/10.1134/S004060151809001X.
4. Paholkov V.V., Rogozhkin S.A., Fadeev I.D., Shepelev S.F. Analiz rezul'tatov teplogidravlicheskikh ispytaniy sistemy avariynogo raskholazhivaniya, vypolnennykh na etape vvoda v ekspluatatsiyu RU BN-800 [Analysis of the results of thermal-hydraulic tests of the emergency cooling system performed at the stage of commissioning of the BN-800 reactor plant]. Sbornik trudov odinnadcatoj mezhdunarodnoj nauchno-tekhnicheskoj konferencii AO “Koncern Rosenergoatom” na temu “Bezopasnost', effektivnost' i ekonomika atomnoj energetiki” (MNTK-2018) [Proc. of the eleventh international scientific and technical conference of Rosenergoatom Concern JSC on the topic “Safety, efficiency and economics of nuclear energy” (IRTC-2018)]. Moscow, 2018, pp. 648–651.