The meeting of the European User Group of MELCOR and MACCS codes used to model and predict severe accident progression in the reactors and spent fuel pools was held in Brugg (Switzerland) in early April.
The participants are representatives of 15 countries, including the expert NPP Thermohydraulic and Probabilistic Safety Analysis Department in the State Scientific and Technical Center for Nuclear and Radiation Safety Oleksandr Kotsuba, presented 25 reports.
These presentations described experience in using these codes by the European organizations ad specifying the approaches to modelling ad revealing further areas for their implementation.
SSTC NRS exert presented a report on the experience in using different versions of MELCOR code for accident modeling severe in the spent fuel pool. During the presentation, he noted the need to consider individual recommendations regarding approaches to model severe accidents in the spent fuel pool using MELCOR code. Oleksandr Kotsuba also presented an analysis of the impact of differences in using various versions of MELCOR code, and causes of these differences, which are recommended to be avoided when modeling transients in the reactor and spent fuel pool using different codes and their versions.
As Oleksandr Kotsuba noted, the participation in this meeting allowed familiarization with the international experience related to the approaches to transient modeling at nuclear facilities, as well as presentation of own experience in analyzing this challenge:
“Summarizing the information obtained from the presentations and discussions with other meeting participants, one may conclude that the approaches used by SSTC NRS in severe accident modeling in the spent fuel pool using MELCOR code correspond to the practice of the leading European organizations and institutions performing research and development on this topic. The meeting organizers noted that individual recommendations proposed by SSTC NRS for severe accident modeling in the spent fuel pool using MELCOR code are useful in further research of transients in the reactor and spent fuel pool using computer codes”.