Code analysis on transient behavior of LWR MOX fuel during the test-irradiation in Halden reactor

Suzuki, Motoe; Nagase, Fumihisa 

The behavior of MOX fuels which were base-irradiated in PWR and test-irradiated in the Halden reactor was analyzed by the latest version of fuel performance code FEMAXI-7. For the calculation conditions, linear heat rate history, power density profile, and coolant condition etc. were given consistently from the base- to test-irradiation to predict the fuel temperature, fission gas release rate, and cladding deformation, etc. Comparison of calculated values with the measured data during the test-irradiation shows a reasonable agreement in thermal analysis results such as fuel temperatures and fission gas release rates, while the cladding deformation, which is involved with various interactions, suggests that it is still necessary to analyze and consider an optimum combination of models and their parameters to obtain a satisfactory prediction.