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Nuclear Energy Conference & Expo (NECX)
September 8–11, 2025
Atlanta, GA|Atlanta Marriott Marquis
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Hash Hashemian: Visionary leadership
As Dr. Hashem M. “Hash” Hashemian prepares to step into his term as President of the American Nuclear Society, he is clear that he wants to make the most of this unique moment.
A groundswell in public approval of nuclear is finding a home in growing governmental support that is backed by a tailwind of technological innovation. “Now is a good time to be in nuclear,” Hashemian said, as he explained the criticality of this moment and what he hoped to accomplish as president.
Yu-Chih Ko, Ching-Hui Wu, Min Lee
Nuclear Technology | Volume 155 | Number 1 | July 2006 | Pages 22-33
Technical Paper | Reactor Safety | doi.org/10.13182/NT06-A3743
Articles are hosted by Taylor and Francis Online.
Probabilistic safety assessment (PSA) uses a systematic approach to estimate the reliability and risk of a nuclear power plant (NPP). Over the past few years, severe accident management guidance (SAMG), which delineates the mitigation actions of core melt accidents of an NPP, has been developed to support operators and staff in the technical support center in dealing with those misfortunes. It can be expected that the implementation of SAMG will lower the containment failure frequency and reduce the amount of radionuclides released to the environment during the accident. The plant studied is the Maanshan NPP of Taiwan Power Company, which employs a Westinghouse-designed three-loop pressurized water reactor (PWR) with large dry containment.The containment system event trees and containment phenomenological event trees of the Level-2 PSA model are modified to incorporate the new mitigation actions specified in SAMG. The HCR (Human Cognitive Reliability) and THERP (Technique for Human Error Rate Prediction) models are used to quantify the human error probability (HEP) of all the actions in the Level-2 PSA model. The MAAP4 (Module Accident Analysis Program version 4) code is used to perform thermohydraulic calculations to determine the demand time required in the HEP analysis.The results show that the frequency of most of the source term categories is reduced except the one in which both the reactor pressure vessel and containment are intact. The containment failure frequency is reduced by 14.8% after the implementation of SAMG. The frequency of containment early failure is reduced by 16.2%. Most of the reduction in the containment early failure frequency comes from the reduction in the induced steam generator tube rupture (STGR). The frequency of induced SGTR was reduced from 2.3 × 10-7/reactoryr to 1.0 × 10-8/reactoryr.