Implications of the Rosa/AP600 High- and Intermediate-Pressure Test Results

Abstract

Westinghouse has submitted the new AP600 reactor design for certification under the U.S. Nuclear Regulatory Commission (NRC) regulations for standard designs. The NRC has performed confirmatory testing in the Rig of Safety Assessment (ROSA) - V Large-Scale Test Facility run by the Japan Atomic Energy Research Institute (JAERI). The ROSA/AP600 test results as provided by JAERI are reviewed in terms of their implications for the original technical concerns that were to be addressed by the testing program. Implications for computer code capabilities are also discussed. Since gravity-driven natural circulation flow in a complicated piping network was a key concern, the review concentrates on this aspect of the ROSA/AP600 test results. In particular, it looks at the possible effect of system interactions at high pressure and during early depressurization. It identifies those ROSA/AP600 test occurrences that point to processes that could delay automatic depressurization system initiation or in-containment reactor water storage tank injection. Since most of the tests run were small-break loss-of-coolant accidents (SBLOCAs), the review focusses on this type of scenario. A comparison of several SBLOCA tests is presented.