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# MELCOR-ASTEC Crosswalk of the Accident at Fukushima-Daiichi Unit 1: Phase I Analysis (NUREG/IA-0510)

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## Publication Information

**Manuscript Completed:** May 2018**Date Published:** June 2019

**Prepared by:** N. Andrews1, C. Faucett1, S. Belon2, C. Bouillet2, and H. Bonneville2

1Sandia National Laboratories Albuquerque, NM 871852Institut de Radioprotection et de Surete Nucleaire Cadarache, France, BP 3- 13115 St-Paul-Lez-Durance Cedex

D. Algama, NRC Project ManagerDivision

Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

**Prepared as part of:** The Agreement on Research Participation and Technical Exchange Under the Cooperative Severe Accident Research Program (CSARP)

**Published by:** U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 

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## Abstract

This analysis compares Sandia National Laboratories' (SNL) MELCOR results for the first phase of the Modular Accident Analysis Program (MAAP)-MELCOR Crosswalk to the Accident Source Term Evaluation Code (ASTEC), developed by the Institut de Radioprotection et de Sureté Nucléaire (IRSN), results for the same accident scenario. Similar to the original MAAP-MELCOR Crosswalk, this analysis integrates system response of both containment and reactor pressure vessel (RPV), core degradation behavior, lower plenum behavior and lower head failure, and finally hydrogen behavior and generation. 

The accident scenario developed by the Electric Power Research Institute (EPRI) and SNL for this analysis is stylized after accident progression of Fukushima Daiichi Unit 1 to better highlight areas of similarity and differences in the two computer codes studied. Hence, this work is not appropriate for extrapolation to the area of Fukushima forensic study. The behavior of the main steam line isolation valve, control rod drive mechanism, feedwater system, safety relief valve, and isolation condenser were made constant between the two codes. The MELCOR simulation was run to 16 hours, while the ASTEC simulation was run a slightly shorter amount of time before to the point of lower head failure. Ex-vessel behavior was not examined in this analysis.

Key differences in the system response were found to result from differing thermal-hydraulic models, how the two codes treat in-vessel core relocation, and how the codes treat debris generated. MELCOR treats the core debris primarily as particulate debris, whereas ASTEC treats debris in a single phase – called "magma" – which often resembles a molten pool. This has significant importance in predicting the total amount of hydrogen generated and the total amount of convective heat transfer away from degraded core materials. Key differences were also found in the total amount of core debris relocating to the lower plenum and then ex-vessel during the scenario, with ASTEC predicting significantly more core debris relocating. 

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