Authors & Affiliations
Anfimov A.M., Kuznetsov D.V.
Afrikantov Experimental Design Bureau for Mechanical Engineering, Nizhny Novgorod, Russia
Anfimov A.M. – Head of Group. Contacts: 15, Burnakovsky proezd, Nizhny Novgorod, 603074. Tel.: +7 (831) 246-94-40; e-mail:
Kuznetsov D.V. – Design Engineer 1st Category, Department for Physical and Thermal-Hydraulic Analyses of Stationary Reactor Plants.
Abstract
The article presents the results of computational studies on a heat removal accident of a three-circuit and three-loop BN-type reactor plant with an integral layout of the primary circuit, taking into account the implementation of various management measures. Successful organization of accident management measures is considered from the point of view of fulfilling the main safety functions: bringing the reactor into subcritical state, ensuring core cooling, and limitation of radioactive releases into the environment.
To analyze the course of the emergency mode, the characteristics of the BN reactor plant were selected that are close to those of medium-sized power plants. To identify optimal combinations of organizational and engineering measures to manage a heat removal accident, various scenarios of its development have been analyzed: (1) a heat removal accident with failing means of active reactivity control and automatic emergency cooldown system connection; (2) a heat removal accident with failures of automatic emergency cooldown system connection and automatic safety trip.
The computational analysis of various accident management measures was conducted using the SOKRAT-BN code certified for such calculations.
The results of studies on a heat removal accident, taking into account the implementation of various measures for its management, have shown that:
1) PSR-H actuation confines the core damage extent to breach of fuel cladding. To reduce the number of breached fuel rods, it is necessary to increase the efficiency of the PSR-H rods or to increase the speed of their insertion into the reactor core.
2) If automatic ECDS connection is unavailable, the personnel can start the ECDS loops manually one by one. In this case, residual heat removal is ensured through natural circulation in the primary circuit and circuits of all the three ECDS loops.
Keywords
safety analysis, reactor plant, sodium, heat removal accident, sodium-cooled fast neutron reactor (BN), emergency cooldown system (ECDS), passive safety rod (PSR)
Article Text (PDF, in Russian)
UDC 621.039.586
Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2024, no. 1, 1:5