ICONE23-1560 DEVELOPMENT OF FAST REACTOR CONTAINMENT SAFETY ANALYSIS CODE, CONTAIN-LMR : (2) VALIDATION STUDY OF SODIUM FIRE MODEL IN CONTAIN-LMR ICONE23-1560 DEVELOPMENT OF FAST REACTOR CONTAINMENT SAFETY ANALYSIS CODE, CONTAIN-LMR : (2) VALIDATION STUDY OF SODIUM FIRE MODEL IN CONTAIN-LMR
Access this Article
The CONTAIN-LMR code is being developed in the Japan Atomic Energy Agency (JAEA) since 1980's for the purpose to utilize for the quantitative assessment of accident consequences considered in sodium-cooled fast reactor (SFR) plant. It is a best-estimate integrated analysis tool to predict lots of thermal- hydraulic behaviors and radioactive materials transfer to the environment in the case of hypothetically postulated ex-vessel severe accident progression. Out of various physical and chemical behaviors treated in the code, the present paper describes sodium fire related study issues such as computational modeling and its validation activities with focusing on important evaluating targets. Major objective in using sodium leak and fire modules in the evaluation of SFR severe accident is not necessarily to clarify the detailed or localized phenomena, but rather to quantify the longer-term or overall dominant thermal consequences affecting to the plant building. The related computational models implemented in CONTAIN-LMR are therefore typical ones to reproduce sodium spray and pool fires, which include relatively simplified model for eliminating higher computational cost. The authors present in this paper representative code validation practices for fundamental sodium combustion and consequential heat and mass transfer behavior models through the numerical analyses of sodium leak and fire experiments performed in the SAPFIRE facility.
- The Proceedings of the International Conference on Nuclear Engineering (ICONE)
The Proceedings of the International Conference on Nuclear Engineering (ICONE) 2015.23(0), _ICONE23-1-_ICONE23-1, 2015
The Japan Society of Mechanical Engineers