A design was developed for leading thermocouples from a high-temperature, pressurized water reactor-coolant system of such integrity that no reactor shutdowns were caused by its use. Using this design, measurements of the fuel-element-cladding temperature and its variation with time were made in three tests on elements clad in type X-8001 aluminum alloy. The following conclusions were reached from the test results: (1) the cladding temperature of a fuel element operated at low heat flux in high bulk-outlet temperature water did not increase with time and was slightly lower than predicted by the Sieder-Tate equation; (2) cladding temperatures of fuel elements operated at high heat flux in either high bulk-inlet or outlet temperature water increased 40 C higher than predicted by the Sieder-Tate equation with initial temperatures equal to the predicted temperatures; and (3) the rate of temperature increase appeared dependent only on fuel-element heat flux and location with respect to the front and rear faces of the reactor.

Limitations on the service life of a fuel element imposed by degradation of the fissile core during irradiation have been a matter of great concern. Limitations imposed by changes in cladding properties during irradiation should be evaluated with equal care. Zircaloy-2, stainless steel, and aluminum alloys have been irradiated in the form of cladding on metallic and ceramic fuel elements. Several aspects of fuel behavior as influenced by these clad materials will be discussed. All observations related to irradiation behavior in this paper have been made on fuel specimens irradiated in water coolant.