Fukushima Daiichi Unit 1 Ex-Vessel Prediction: Core-Concrete Interaction

Journal Article · Mon Oct 31 00:00:00 EDT 2016 · Nuclear Technology

DOI:https://doi.org/10.13182/NT16-48· OSTI ID:1346626

Robb, Kevin R. ^[1]; Farmer, Mitchell T. ^[2]; Francis, Matthew W. ^[1]

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1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2. Argonne National Lab. (ANL), Argonne, IL (United States)

Lower head failure and corium-concrete interaction were predicted to occur at Fukushima Daiichi Unit 1 (1F1) by several different system-level code analyses, including MELCOR v2.1 and MAAP5. Although these codes capture a wide range of accident phenomena, they do not contain detailed models for ex-vessel core melt behavior. However, specialized codes exist for the analysis of ex-vessel melt spreading (e.g., MELTSPREAD) and long-term debris coolability (e.g., CORQUENCH). On this basis, in this paper an analysis was carried out to further evaluate ex-vessel behavior for 1F1 using MELTSPREAD and CORQUENCH. Best-estimate melt pour conditions predicted by MELCOR v2.1 and MAAP5 were used as input. MELTSPREAD was then used to predict the spatially dependent melt conditions and extent of spreading during relocation from the vessel. The results of the MELTSPREAD analysis are reported in a companion paper. This information was used as input for the long-term debris coolability analysis with CORQUENCH. For the MELCOR-based melt pour scenario, CORQUENCH predicted the melt would readily cool within 2.5 h after the pour, and the sumps would experience limited ablation (approximately 18 cm) under water-flooded conditions. Finally, for the MAAP-based melt pour scenarios, CORQUENCH predicted that the melt would cool in approximately 22.5 h, and the sumps would experience approximately 65 cm of concrete ablation under water-flooded conditions.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE)
• Contributing Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Grant/Contract Number: AC05-00OR22725
• OSTI ID: 1346626
• Alternate ID(s): OSTI ID: 1351266
• Journal Information: Nuclear Technology, Journal Name: Nuclear Technology Journal Issue: 3 Vol. 196; ISSN 0029-5450
• Publisher: American Nuclear Society (ANS)
• Country of Publication: United States
• Language: English

Related Subjects

42 ENGINEERING
99 GENERAL AND MISCELLANEOUS
Fukushima Daiichi
MCCI
severe accident