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An estimation of an operator's action time by using the MARS code in a small break LOCA without a HPSI for a PWR
Authors:Seok-Jung Han   Ho-Gon Lim  Joon-Eon Yang  
Affiliation:aIntegrated Safety Assessment Division, Korea Atomic Energy Research Institute, P.O. Box 105, Yuseung, Daejeon 305-600, Republic of Korea
Abstract:
To estimate the success criteria of an operator's action time for a probabilistic safety/risk assessment (PSA/PRA) of a nuclear power plant, the information from a safety analysis report (SAR) and/or that by using a simplified simulation code such as the MAAP code has been used in a conventional PSA. However, the information from these is often too conservative to perform a realistic PSA for a risk-informed application. To reduce the undue conservatism, the use of a best-estimate thermal hydraulic code has become an essential issue in the latest PSA and it is now recognized as a suitable tool. In the same context, the ‘ASME PRA standard’ also recommends the use of a best-estimate code to improve the quality of a PSA. In Korea, a platform to use a best-estimate thermal hydraulic code called the MARS code has been developed for the PSA of the Korea standard nuclear power plant (KSNP). This study has proposed an estimation method for an operator's action time by using the MARS platform. The typical example case is a small break loss of coolant accident without the high pressure safety injection system, which is one of the most important accident sequences in the PSA of the KSNP. Under the given accident sequence, the operator has to perform a recovery action known as a fast cooldown operation. This study focuses on two aspects regarding an operator's action; one is how they can operate it under some restrictions; the other is how much time is available to mitigate this accident sequence. To assess these aspects, this study considered: (1) the operator's action model and (2) the starting time of the operation. To show an effect due to an operator's action, three kinds of control models (the best-fitting, the conservative, and the proportional-integral) have been assessed. This study shows that the developed method and the platform are useful tools for this type of problem and they can provide a valuable insight related to an operator's actions.
Keywords:ADV, atmosphere dump valve   ASC, aggressive secondary cooldown   DBA, design basis accident   EOP, emergency operation procedure   FSAR, final safety analysis report   HPSI, high pressure safety injection   ISI, in-service inspection   KSNP, Korean standard nuclear power plant   LHGR, linear heat generation rate   LOCA, loss of coolant accident   LPSI, low pressure safety injection   MCR, main control room   MSIS, main steam isolation signal   MSIV, main steam isolation valve   MSSV, main steam safety valve   PCT, peak cladding temperature   PSA, probabilistic safety assessment   PRA, probabilistic risk assessment   PTS, pressurized thermal shock   PWR, pressurized water reactor   RCP, reactor coolant pump   RCS, reactor coolant system   RIA, risk-informed application   SIAS, safety injection actuation signal   TBV, turbine bypass valve
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