Differing system response behavior resultant from diverging core degradation models found within the Melcor-Astec crosswalk

​This analysis compares the MELCOR results for the first phase of the Modular Accident Analysis Program (MAAP)-MELCOR Crosswalk to Accident Source Term Evaluation Code (ASTEC) results for the same accident scenario. Similar to the original MAAP MELCOR Crosswalk this analysis contains an analyses of system response of both the containment and RPV, core degradation behavior, lower plenum behavior and lower head failure, and finally hydrogen behavior and generation.

 

The accident scenario developed by EPRI and SNL for this analysis is stylized after accident progression of Fukushima Daiichi Unit 1. However, this accident scenario is for the purpose of code comparison and not for Fukushima Daiichi forensic efforts. The behavior of the main steam line isolation valve, control rod drive mechanism, feedwater system, safety relief valve and the isolation condenser behavior were made constant between the two codes. The MELCOR simulation was run to 16 hours, while the ASTEC simulation was run to the point of lower head failure.

 

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 field – "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.