MAIN RESULTS OF HYDRA-IBRAE/LM/V1 CODE VERIFICATION DURING BN-600 RP EXPERIMENTAL MODES

Authors & Affiliations

Anfimov A.M.¹, Gorbunov V.S.¹, Kuznetsov D.V.¹, Osipov S.L.¹, Ivanov E.N.², Klimonov I.A.², Kudashov I.G.², Mosunova N.A.², Usov E.V. ^2
1Afrikantov OKB Mechanical Engineering, Nizhny Novgorod, Russia
²Nuclear Safety Institute of the Russian Academy of Science, Moscow, Russia

Anfimov A.M. – Design Engineer 1 category, Afrikantov OKB Mechanical Engineering. Contacts: 15 Burnakovsky proezd, Nizhny Novgorod, Russia, 603074. Tel.: (831)246-94-40; e-mail:This email address is being protected from spambots. You need JavaScript enabled to view it..
Gorbunov V.S. – Lead Design Engineer, Cand. Sci. (Tech.), Afrikantov OKB Mechanical Engineering.
Kuznetsov D.V. – Design Engineer 3 categories, Afrikantov OKB Mechanical Engineering.
Osipov S.L. – Head of Department, Cand. Sci. (Tech.), Afrikantov OKB Mechanical Engineering.
Ivanov E.N. – Engineer, Nuclear Safety Institute of the Russian Academy of Science.
Klimonov I.A. – Engineer, Nuclear Safety Institute of the Russian Academy of Science.
Kudashov I.G. – Engineer, Nuclear Safety Institute of the Russian Academy of Science.
Mosunova N.A. – Head of Department, Cand. Sci. (Phys.-Math.), Nuclear Safety Institute of the Russian Academy of Science.
Usov E.V. – Head of Laboratory, Cand. Sci. (Tech.), Nuclear Safety Institute of the Russian Academy of Science.

Abstract

The thermo-hydraulic system code HYDRA-IBRAE/LM/V1 is under development at IBRAE RAN within the particular project “New Generation Codes” of the “Proryv” project area and is directed toward simulation of liquid metal cooled fast reactor behavior under normal operation, abnormal operation and emergency modes. IBRAE RAN and JSC “Afrikantov OKBM” carry out code verification as applied to the sodium coolant during BN-600 reactor plant (RP) experimental modes. To perform the analysis, the BN-600 nodalization scheme was developed to take account of fast neutron RP main features: non-uniformity of fuel element temperature state along the core height and radius; heat ex-change in the core space between subassemblies; heat transfer between cold and hot reactor cham-bers, formed by reactor internals; non-uniformity of coolant temperature distribution along the upper mixing chamber height and radius, etc. This paper deals with code verification for the modes of forced coolant circulation loss and change over to the natural circulation in RP sodium circuits. The analysis results for the following modes are given: emergency cooling using natural circulation in case of reactor power decrease from 19 down to 1 % of nominal value; emergency cooling using natural circulation from reactor power of ~ 50 % of nominal value. The comparative analysis confirmed the capability of HYDRA-IBRAE/LM/V1 code to simulate sufficiently the transients which are characteristic for fast neutron RPs at the moment of natural circulation development in the plant circuits and associated with deterioration of heat removal from the core and heat exchange equipment.

Keywords
natural circulation, thermo-hydraulic system code, sodium coolant, core, experimental modes, verification, emergency cooling

Article Text (PDF, in Russian)

References

UDC 004.415.53:621.039.586

Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2016, issue 4, 4:20