The accident progression and melt behavior at the Fukushima Daiichi Units 1 and 2 were investigated using MELCOR 2.1. In the modeling the lower head failure mechanism by penetration tube rupture and ejection was modeled. In the modeling of Unit 2, according to the latest findings by TEPCO investigation, the possibilities of torus room flooding, RCIC piping leakage and thermal stratification in suppression pool were taken into account. The analysis results indicate that for Unit 1 when considering penetration tube failure, a part of debris still remained in the lower head after debris discharge; otherwise all the debris discharged out. The present MELCOR modeling of Unit 2 well reproduced the RPV and PCV pressure. A part of the core was damaged and the debris that slumped into the lower head was sufficiently cooled down. The pressure vessel kept intact.
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2017 25th International Conference on Nuclear Engineering
July 2–6, 2017
Shanghai, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5784-7
PROCEEDINGS PAPER
Investigation on Accident Progression and Melt Behavior at the Fukushima Daiichi Units 1 and 2 Using MELCOR Code
Shan Zheng,
Shan Zheng
China Datang Corporation Science and Technology Research Institute Northwest, Xi’an, China
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Akifumi Yamaji,
Akifumi Yamaji
Waseda University, Tokyo, Japan
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Daotong Chong,
Daotong Chong
Xi’an Jiaotong University, Xi’an, China
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Junjie Yan
Junjie Yan
Xi’an Jiaotong University, Xi’an, China
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Gen Li
Xi’an Jiaotong University, Xi’an, China
Shan Zheng
China Datang Corporation Science and Technology Research Institute Northwest, Xi’an, China
Akifumi Yamaji
Waseda University, Tokyo, Japan
Daotong Chong
Xi’an Jiaotong University, Xi’an, China
Junjie Yan
Xi’an Jiaotong University, Xi’an, China
Paper No:
ICONE25-66389, V006T08A029; 7 pages
Published Online:
October 17, 2017
Citation
Li, G, Zheng, S, Yamaji, A, Chong, D, & Yan, J. "Investigation on Accident Progression and Melt Behavior at the Fukushima Daiichi Units 1 and 2 Using MELCOR Code." Proceedings of the 2017 25th International Conference on Nuclear Engineering. Volume 6: Thermal-Hydraulics. Shanghai, China. July 2–6, 2017. V006T08A029. ASME. https://doi.org/10.1115/ICONE25-66389
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