Thermal-hydraulic (T/H) analyses are used to support the level 1 Probabilistic Risk Assessment (PRA) success criteria and the manual operation time. To address the multiple-failure accident scenarios that are considered in the PRA, usually numerous T/H analyses were performed. So it is meaningful to develop a relative simple T/H model with acceptable accuracy for level 1 PRA T/H analyses. To achieve this object, the core modeling effects on the core damage progression were studied according to ASME/ANS RA-Sa-2009. Two types of core modeling methods were studied, including single channel core modeling and multi-channel core modeling. For the single channel core modeling, the study was focused on the axial nodes number effect. For the multi-channel core modeling, the cross-flow effects were studied. Several cases were calculated on a 3-loop PWR medium size break LOCA core damage scenario with Relap5/MOD3.2. Some key parameters related to the core state, such as peak cladding temperature (PCT), core water level and coolant inventory, were compared and analyzed. A kind of core modeling for level 1 PRA T/H analyses was suggested at the end.
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2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5581-2
PROCEEDINGS PAPER
RELAP5 Core Modeling Study for Level 1 PRA Thermal-Hydraulic Analyses
Changjiang Yang
Changjiang Yang
China Nuclear Power Engineering Co., Ltd., Beijing, China
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Changjiang Yang
China Nuclear Power Engineering Co., Ltd., Beijing, China
Paper No:
ICONE21-16125, V004T09A074; 6 pages
Published Online:
February 7, 2014
Citation
Yang, C. "RELAP5 Core Modeling Study for Level 1 PRA Thermal-Hydraulic Analyses." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 4: Thermal Hydraulics. Chengdu, China. July 29–August 2, 2013. V004T09A074. ASME. https://doi.org/10.1115/ICONE21-16125
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