SIMULATION OF NATURAL CIRCULATION FLOW DURING POWER MANEUVERING IN MASLWR TEST FACILITY WITH COMPUTER CODE RELAP-5/MOD3.2
Simulation of transient behavior of a nuclear power plant is the primary objective of system thermal hydraulic code's calculation precision which is ensured by 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 which characterize the thermal hydraulic phenomena during both steady state and transient conditions. The validation of the best estimate thermal hydraulic system code RELAP-5/MOD3.2 and also, in-house computer code 'SPTS' against the Multi-Application Small Light-Water Reactor (MASLWR) design, and is the topic of the present paper. The current work relates to the validation 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 modelling and predicting capability of state-of-the-art thermal hydraulic code in simulating NC flow within integral type reactor.
MASLWR Test Facility is a system-level test facility to examine natural circulation phenomena of importance to small modular concept evolved for integral MASLWRs and is scaled at 1:3 length scale, 1:254 volume scale and 1:1 time scale and designed for full pressure and temperature of 11.4 MPa and 590 K prototype operation respectively. This facility includes Reactor Pressure Vessel (RPV), Pressurizer and Automatic Depressurization System (ADS) blow-down/vent lines; secondary circuit along-with an integrated helical coil SG and containment structure. MASLWR test facility is extensively instrumented to capture complete transient behaviour.
This paper covers the analytical simulation of MASLWR test facility to investigate RPV natural circulation fluid flow rate through the primary loop in normal operating conditions at different power levels using computer code RELAP-5/MOD3.2.
It is found that the predicted results were in excellent agreement with experimental results within the accuracy of the experimental facility. It is observed that overall transient of key identified thermal hydraulic parameters such as RPV pressure, temperature at various location of RPV are well predicted by the code and mainly governed by rate of rise of core heater power and SG feed water flow rate. All phenomena are captured well by the code which has enhanced the confidence in performing accident analysis of light water reactors including natural circulation reactors.