Development of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling

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dc.contributor.authorKim, Jong Hyunko
dc.contributor.authorJin, Dong Sikko
dc.contributor.authorChang, Soon-Heungko
dc.identifier.citationNUCLEAR ENGINEERING AND DESIGN, v.263, pp.241 - 254-
dc.description.abstractThe new safety analysis methodology for the CANDU-6 nuclear power plant (NPP) moderator system failure has been developed by using the coupling technology with the thermalhydraulic code, CATHENA and reactor core physics code, RFSP-IST. This sophisticated methodology can replace the legacy methodology using the MODSTBOIL and SMOKIN-G2 in the field of the thermalhydraulics and reactor physics, respectively. The CATHENA thermalhydraulic model of the moderator system can simulate the thermalhydraulic behaviors of all the moderator systems such as the calandria tank, head tank, moderator circulating circuit and cover gas circulating circuit and can also predict the thermalhydraulic property of the moderator such as moderator density, temperature and water level in the calandria tank as the moderator system failures go on. And these calculated moderator thermalhydraulic properties are provided to the 3-dimensional neutron kinetics solution module - CERBRRS of RFSP-IST as inputs, which can predict the change of the reactor power and provide the calculated reactor power to the CATHENA. These coupling calculations are performed at every 2 s time steps, which are equivalent to the slow control of CANDU-6 reactor regulating systems (RRS). The safety analysis results using this coupling methodology reveal that the reactor operation enters into the self-shutdown mode without any engineering safety system and/or human interventions for the postulated moderator system failures of the loss of heat sink and moderator inventory, respectively. (c) 2013 The Authors. Published by Elsevier B.V. All rights reserved.-
dc.titleDevelopment of safety analysis methodology for moderator system failure of CANDU-6 reactor by thermal-hydraulics/physics coupling-
dc.citation.publicationnameNUCLEAR ENGINEERING AND DESIGN-
dc.contributor.localauthorChang, Soon-Heung-
dc.contributor.nonIdAuthorKim, Jong Hyun-
dc.contributor.nonIdAuthorJin, Dong Sik-
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