Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

The probabilistic safety of the supercritical-water cooled fast reactor (SC FR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no re circulation loops in the once-through direct cycle system, which is the mos t important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system conf iguration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency c ore flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure c oolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordi ngly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that o f the Japanese BWR when Japanese data are employed for initiating event fre quencies. Four alternatives to the safety system configurations are also ex amined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a onc e-through coolant system, the CDF is not high due to the diversity of feedw ater systems as the direct cycle characteristics. (C) 1999 Elsevier Science Ltd. All rights reserved.