Evaluation of Advanced Thermal-Hydraulic System Codes for Safety Analysis of Integral Type PWR

Abstract

The integral Pressurized Water Reactor (PWR) concept, which contains the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high possibility for near-term deployment. An IAEA International Collaborative Standard Problem (ICSP) on “Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Primary System and Containment during Accidents” has been conducted since 2010. Oregon State University of USA has offered their experimental facility, which was built to demonstrate the feasibility of Multi-Application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven IAEA Member States have been participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiment. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena including the coupling of primary system, high pressure containment and cooling pool are expected to occur in this transient. The ICSP has been conducted in three phases: pre-test (with designed initial & boundary conditions before the conduction of the experiment), blind (with real initial & boundary conditions after the conduction of the experiment) and open simulation (after the observation of real experimental data). Most advanced thermal-hydraulic system analysis codes like TRACE, RELAP5-3D and MARS have been assessed against experiments conducted at MASLWR test facility.