Properties of fuel-element materials in the steam-generating heavy water reactor

Features of the steam-generating heavy water reactor (SGHWR) design and fuel-element operating experience are reviewed. Up to over half the design burnup, no systematic fuel defects that can be related to design criteria or materials selection or specification have occurred. Monitoring examination of irradiated fuel elements has not shown evidence of any impending problems, but has indicated the need to pay particular attention to nodular corrosion of the Zircaloy-2 cladding as a possible life-limiting feature. As a result of this examination, fuel pin axial growth was found to be exceeding design predictions and corrective action was taken. Factors governing the choice of cold-worked and stress-relieved Zircaloy-2 as the cladding alloy for SGHWR are discussed together with the influence of strength level on the ductility of irradiated cladding and the possible advantages of fully annealed Zircaloy-2. Reasons for the choice of high-density pelleted U02 were based on conductivity, moisture content, and stability. This choice is justified by experience, but it is suggested that the specified pore size and grain size should be reviewed in the light of densification behaviour and the focussing of attention on performance of fuel pins in accidents. Environmental aspects and their influence on fuel performance are exemplified by power cycling and power ramp experience, which are governed by the effects of temperature and neutron flux on the stability of both cladding and U02, and by the fission product environment inside the fuel pin. Work carried out to determine the swelling behaviour of fuel pins in a loss-of-coolant accident shows the importance of fission gas release from U02 and changes in cladding properties during the temperature transient. Design maximum rating is currently set by criteria applied to ensure that there will be no temperature run-away during such an accident. (author)