• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.11-21
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• 2020

This study aims to develop and advance the evaluation technology for assessing PHWR safety. For this purpose, the complete loss of AC power or station blackout (SBO) was selected as a target accident scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes the main features of the primary heat transport system with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was achieved successfully by running the present model to check out the stable convergence of the key parameters. Subsequently, through the SBO transient analyses two cases with and without the coolant leakage via the PHTS pumps were simulated and the behaviors of the major parameters were compared. The sensitivity analysis on the amount of the coolant leakage by varying its flow area was also performed to investigate the effect on the system responses. It is expected that the results of the present study will contribute to upgrading the evaluation technology of the detailed thermal hydraulic analysis on the SBO transient of the operating PHWRs.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.1
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.