• Nuclear Engineering and Technology
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• 제38권1호
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• pp.1-10
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• 2006

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability.

• Nuclear Engineering and Technology
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• 제47권1호
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• pp.1-10
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• 2015

After the Fukushima Daiichi nuclear power plant (NPP) accident, new regulatory requirements were enforced in July 2013 and a backfit was required for all existing nuclear power plants. It is required to take measures to prevent severe accidents and mitigate their radiological consequences. The Regulatory Standard and Research Department, Secretariat of Nuclear Regulation Authority (S/NRA/R) has been conducting numerical studies and experimental studies on relevant severe accident phenomena and countermeasures. This article highlights fission product (FP) release and hydrogen risk as two major areas. Relevant activities in the S/NRA/R are briefly introduced, as follows: 1. For FP release: Identifying the source terms and leak mechanisms is a key issue from the viewpoint of understanding the progression of accident phenomena and planning effective countermeasures that take into account vulnerabilities of containment under severe accident conditions. To resolve these issues, the activities focus on wet well venting, pool scrubbing, iodine chemistry (in-vessel and ex-vessel), containment failure mode, and treatment of radioactive liquid effluent. 2. For hydrogen risk: because of three incidents of hydrogen explosion in reactor buildings, a comprehensive reinforcement of the hydrogen risk management has been a high priority topic. Therefore, the activities in evaluation methods focus on hydrogen generation, hydrogen distribution, and hydrogen combustion.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2498-2503
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• 2023

When the reactor vessel is penetrated in a severe accident of light water reactor, the molten fuel-coolant interaction including the jet breakup occurs and the jet breakup length becomes one of the important parameters. Most numerical studies on jet breakup process have been carried out using dedicated computer codes. Some researchers are trying to apply commercial CFD codes to their investigations on comprehensive jet breakup process. However, the complexity of the phenomena limits the CFD application only to hydrodynamic aspects. In the present study, numerical analysis of jet breakup under vapor generation is pursued using the STAR-CCM + code. The obtained CFD prediction of the MATE09 experiment shows jet breakup progression patterns consistent to the images taken in the experiment. Further, the predicted positions of leading head, which determine the jet breakup length, are in good agreement with the MATE 09 data. The investigation of hydrodynamic effects on the jet breakup with higher jet velocity results in a stronger shear force and earlier jet breakup process even though there exists the vapor pocket around the corium jet. In future studies, the effect of vapor intensity on the jet breakup length would be investigated further by changing other parameters.

• 시스템엔지니어링학술지
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• 제19권2호
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• pp.32-45
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• 2023

Implementing Severe Accident Management (SAM) strategies is crucial for enhancing a nuclear power plant's resilience and safety against severe accidents conditions represented in the analysis of Station Blackout (SBO) event. Among these critical approaches, the In-Vessel Retention (IVR) through External Reactor Vessel Cooling (IVR-ERVC) strategy plays a key role in preventing vessel failure. This work is designed to evaluate the efficacy of the IVR strategy for a high-power density reactor APR1400. The APR1400's plant is represented and simulated under steady-state and transient conditions for a station blackout (SBO) accident scenario using the computer code, ASYST. The APR1400's thermal-hydraulic response is analyzed to assess its performance as it progresses toward a severe accident scenario during an extended SBO. The effectiveness of emergency operating procedures (EOPs) and severe accident management guidelines (SAMGs) are systematically examined to assess their ability to mitigate the accident. A group of associated key phenomena selected based on Phenomenon Identification and Ranking Tables (PIRT) and uncertain parameters are identified accordingly and then propagated within DAKOTA Uncertainty Quantification (UQ) framework until a statistically representative sample is obtained and hence determine the uncertainty bands of key system parameters. The Systems Engineering methodology is applied to direct the progression of work, ensuring systematic and efficient execution.