Development of a thermohydraulic model of the European Sodium Fast Reactor (ESFR) using the system code TRACE

One of the main goals of the Generation IV International Forum (GIF) nuclear energy systems is to excel in safety and reliability. To pursue such objective, the development of computational tools able to simulate operation conditions that may be critical for the safety of these innovative reactor concepts is essential. This paper presents how a thermohydraulic model of the ESFR plant has been developed using the best-estimate system code TRACE. The model simulates the primary, secondary and tertiary system. The primary system includes the reactor core with point kinetic neutronics implemented and it is able to analyse different accident transient scenarios. The work presented in the paper provides the first steps for the development of the transient part of an integrated safety analysis platform with capabilities to perform detailed simulations of the reactor dynamic thermohydraulic-neutronic coupling techniques.

LAZARO CHUECA Aurelio;  AMMIRABILE Luca;  MARTORELL S.;  VERDU G.; 

2014-09-15

European Nuclear Society

JRC75035

978-92-95064-14-0,   

http://www.euronuclear.org/events/enc/enc2012/transactions/ENC2012-Transactions-Advanced-Reactors.pdf,    https://publications.jrc.ec.europa.eu/repository/handle/JRC75035,