SEDL / STP / STP681-EB / STP36681S

Analysis of Pressurized Water Reactor Fuel Pin Length Changes

Harbottle, JE
Research officers, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire,

Cornell, RM
Research officers, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire,

Pages: 11    Published: Jan 1979

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Abstract

Textured Zircaloy fuel cans subjected to a net external coolant pressure will in general show a change in length due to anisotropic irradiation creep if mechanical interaction between the cladding and the fuel are negligible. Based on some recent measurements, the length increases observed in the pins of a pressurized water reactor fuel element are described to within 10 percent by a model based on irradiation growth and anisotropic creep. The model provides an adequate, quantitative explanation of the differences found between pressurized and unpressurized pins taking into account the change of hoop stress with fission gas buildup. The diametral measurements indicate that ratcheting is probably not a dominant mode of deformation in the majority of the cases.