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Report No.

Analysis of MOX fuel behavior in reduced-moderation water reactor by fuel performance code FEMAXI-RM

Suzuki, Motoe; Saito, Hiroaki*; Iwamura, Takamichi

To assess the feasibility of the 31percentPu-MOX fuel rod design of reduced-moderation boiling water reactor in terms of thermal and mechanical behaviors, a single rod which is assumed to be irradiated in the core of RMWR up to 106 GWd/tHM has been analyzed by a fuel performance code FEMAXI-RM which is an extended version of FEMAXI-6 code. In the analysis, design specifications of fuel rod and irradiation conditions have been input, and available models of both MOX fuel and UO$$_{2}$$ fuel have been used complementally. The results are: FGR is several tens of percent, rod internal pressure does not exceed the coolant pressure, and the highest fuel center temperature is 2400K, while cladding diameter increase caused by pellet swelling is within 1percent strain. These predictions suggest that the MOX fuel rod integrity will be held during irradiation in RMWR, though actual behavior of MOX pellet swelling requires to be investigated in detail.



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Category:Nuclear Science & Technology



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