Simulation of loss of feed water transient in MASLWR test facility with computer code RELAP-5/MOD 3.2

Accurate simulation of transient behavior of a nuclear power plant is the primary objective of system thermal hydraulic code calculations, and the evaluation of a code's calculation precision is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled-down model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The validation and assessment of the best estimate thermal hydraulic system code RELAP-5/MOD3.2 against the Multi-Application Small Light-Water Reactor (MASLWR) design is an innovative effort, and is the topic of the present paper. Specifically, the current work relates to the assessment and validation process of RELAP-5/MOD3.2 code against the natural circulation (NC) database developed in the MASLWR test facility through NPCIL participation in an International Atomic Energy Agency (IAEA) International Collaborative Standard problem (ICSP). The purpose of this paper is to present the predicting capability of state-of-the-art thermal hydraulic code in simulating NC flow within integral type reactor and the assessment of passive safety systems. This paper presents the analytical simulation of loss of feed water transient in MASLWR test facility. The experimental data available includes the blow-down phase as well as the coupling of the primary system to containment systems and the long term cooling using cooling pool vessel (CPV) through ADS operation and subsequent to natural circulation. (author)