This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 18019and 21440. Revision 18019 represents the previous official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 21440. Along with the newly updated MELCOR Users’ Guide [2] and Reference Manual [3], users are aware and able to assess the new capabilities for their modeling and analysis applications.
This document details the development of modeling and simulations for existing plant security regimes using identified target sets to link dynamic assessment methodologies by leveraging reactor system level modeling with force-on-force modeling and 3D visualization for developing table-top scenarios. This work leverages an existing hypothetical example used for international physical security training, the Lone Pine nuclear power plant facility for target sets and modeling.
This document details the development of modeling and simulations for existing plant security regimes using identified target sets to link dynamic assessment methodologies by leveraging reactor system level modeling with force-on-force modeling and 3D visualization for developing table-top scenarios. This work leverages an existing hypothetical example used for international physical security training, the Lone Pine nuclear power plant facility for target sets and modeling.
The ADAPT software allows for the examination of aleatory and epistemic uncertainties in complex system transients using the Dynamic Event Tree (DET) methodology. This manual outlines the principles of operation of ADAPT and provides directions for its use. Future plans for the code are briefly outlined.
Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.