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Japan Atomic Energy Agency, Ibaraki 311-1393, Japan


The purpose of the present study is to develop a methodology to evaluate fuel discharge through the CRGT (control-rod guide tube) during CDAs (core-disruptive accidents) of SFRs (sodium-cooled fast reactors), since fuel discharge will decrease the core reactivity and CRGTs have a potential to provide an effective discharge path. Fuel discharge contains multi-component fluid dynamics with phase changes, and, in the present study, the SFR safety analysis code SIMMER (Sn, implicit, multifield, multicomponent, Eulerian recriticality) was utilized as a technical basis. First, dominant phenomena affecting fuel discharge through the CRGT are identified based on parametric calculations by the SIMMER code. Next, validations on the code models closely relating to these phenomena were carried out based on experimental data. It was shown that the SIMMER code with some model modifications could reproduce the experimental results appropriately. Through the present study, the evaluation methodology for the molten-fuel discharge through the CRGT was successfully developed.


Sodium-cooled fast reactor, core disruptive accident; molten-fuel discharge, FBR (fast breeder reactor) safety analysis code SIMMER.

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