Abstract
UO 2 fuel in a Pressurised Heavy Water Reactor (PHWR) fuel pin undergoes noticeable micro-structural changes during irradiation. The restructuring of fuel observed during post irradiation examination (PIE) of irradiated fuel is indicative of fuel centre temperature, radial temperature profile, redistribution of the fissile nuclides, fission products and the mechanism of release of fission gases. The size of various restructured regions observed in the pellet is used to estimate the fuel centre temperature existing during reactor operation. A method has been developed for estimating the fuel temperature and fission gas release fraction in a PHWR fuel pin using the fuel restructuring observed in the cross section of the irradiated fuel. This method has been applied to estimate the maximum fuel centre temperature experienced by the fuel during irradiation and fission gas release in the fuel pins of a PHWR fuel bundle irradiated to a high burn-up. The radial profile of retained fission gas in the pellet has been calculated and compared with the fission product distribution observed in beta gamma auto-radiograph of the fuel. Calculated fission gas release and retained gas radial profile show good match with experimental observations.
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