For a severe accident (a core melting accident) of nuclear power plants, a heat-up of the molten core might cause a overpressurizing of containment building to be damaged, if there couldn’t be given a proper cooling and/or a depressurizing strategy. In order to depressurize containment building and also to minimize the release of radioactive materials, filtered containment venting system (FCVS) might be used for a one of possible options. For a wet-type FCVS, radioactive aerosol released from molten core could be decontaminated by water pool, which is called pool scrubbing effect. The objective of this study is to find out regulatory insights for evaluating a wet-type FCVS for Korean nuclear power plant, APR1400. MELCOR, which is a severe accident analysis code developed by Sandia National Laboratories, was used for simulating postulated accidents. A full-plant scale calculation was performed considering the accident conditions such as temperature, pressure flow rate from containment to the pool of FCVS, behavior of radioactive materials and decontamination factors (DFs) for them. FCVS was operated with containment pressure set points. The decrease thermal margin between containment atmosphere and the pool of the FCVS influenced the DF, because the decreased amount of the steam due to the lowered thermal margin interrupted the radioactive aerosols and steam condensed.
- Fluids Engineering Division
Pool Scrubbing Efficiency of Hypothesized Filtered Containment Venting System for APR1400 Nuclear Power Plants
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Seo, S, Lee, J, & Cho, Y. "Pool Scrubbing Efficiency of Hypothesized Filtered Containment Venting System for APR1400 Nuclear Power Plants." Proceedings of the ASME 2017 Fluids Engineering Division Summer Meeting. Volume 1B, Symposia: Fluid Measurement and Instrumentation; Fluid Dynamics of Wind Energy; Renewable and Sustainable Energy Conversion; Energy and Process Engineering; Microfluidics and Nanofluidics; Development and Applications in Computational Fluid Dynamics; DNS/LES and Hybrid RANS/LES Methods. Waikoloa, Hawaii, USA. July 30–August 3, 2017. V01BT09A002. ASME. https://doi.org/10.1115/FEDSM2017-69387
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