Series: Nuclear and Reactor Constants

since 1971

Русский (РФ)

ISSN 2414-1038 (online)

DOI: 10.55176/2414-1038-2020-2-96-103

Authors & Affiliations

Dolgikh V.P., Zabrodskaya S.V., Lebedeva O.М., Popov E.P., Smykov V.B.
A.I. Leypunsky Institute for Physics and Power Engineering, Obninsk, Russia

Belov A.A.1 – Researcher, Nuclear Safety Institute of the Russian Academy of Sciences.
Bikeev A.S.2 – Head of Software Laboratory.
Daichenkova Yu.S.2 – Engineer. Contacts: 1, Akademika Kurchatova pl., Moscow, Russia, 123182. Tel: +7(499)196-91-49; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Shkarovskiy D.A.2 – Head of the Division of Benchmark Calculations of Nuclear Reactors, Cand. Sci. (Tech.)


At present, the BN-600 and BN-800 reactors are used for tests of experimental fuel assemblies containing MOX and MNUP fuels. For transportation of spent fuel assemblies of BN-600 with mixed nitride uranium-plutonium fuel is planned to use a regular transport container containing 35 fuel assemblies. The purpose of the work is to develop a calculation model of the transport container containing 35 fuel assemblies and radiation safety evaluation of the transport container during carting spent MNUP fuel of BN-600 using the MCU-FR Monte Carlo code. The developed model of the experimental fuel assembly is integrated into the developed calculation model of the transport container. The resulting model was used to calculate the equivalent dose rate from sources of ionizing radiation in the fuel. The azimuthal and radial distributions of the equivalent dose rate in the plane of the FA core center were considered. In the course of radiation safety evaluation, it was shown that under normal transportation conditions, the maximum total calculated radiation levels on the container surface are determined by primary photon radiation and do not exceed the established limit of 10 mSv/h under exclusive use. The developed model of the transport container and the calculation results obtained by means of the MCU-FR code will be used to prepare the validation base of the ODETTA calculation code.

MCU-FR, Monte Carlo method, computer modelling, mixed nitride uranium-plutonium fuel, spent nuclear fuel, radiation safety, transport container

Article Text (PDF, in Russian)


UDC 621.039.54:519.85

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2020, issue 2, 1:10