Transient simulation code development of primary coolant system of Chinese Experimental Fast Reactor

Abstract Chinese Experimental Fast Reactor (CEFR) is a 25 MWe sodium cooled, pool type reactor, which was built at the China Institute of Atomic Energy in Beijing as the forerunner to the first-stage of Chinese fast reactor development plans. In order to understand the response of the Primary Coolant System (PCS) to various transients and train the operators a dynamic model using basic energy and momentum equations was developed with some assumptions. Heat transfer models for reactor core and intermediate heat exchanger were also included. Subroutines were developed to calculate the thermal properties, friction coefficients and heat transfer coefficients of liquid sodium. Gear’s method was applied to solve the dynamic model. A transient analysis code named THPCS (Thermal–Hydraulic code of PCS) was developed and is independent of the design and safety analysis codes. Three typical events, such as loss of one primary pump, unprotected transient overpower and accidental closure of primary pump check valves were chosen and investigated. The prediction results of the code agree well with those of the final safety analysis report of CEFR. A fourth postulated accident, station blackout without scram and loss of all heat sink, was also analyzed to show the ability of the code, which is more serious than the former. The transient simulation code developed in this paper will be useful for the safety operation of CEFR.