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A. S. Bain
Nuclear Technology | Volume 3 | Number 4 | April 1967 | Pages 240-244
Technical Paper and Note | doi.org/10.13182/NT67-A27763
Articles are hosted by Taylor and Francis Online.
UO2 fuel elements, sheathed in Zircaloy or stainless steel, were irradiated under controlled conditions to study the transfer of heat across the fuel-to-sheath interface. Variables studied were diametral clearance, heat-transfer medium, duration of irradiation, and power rating. After irradiation, fractured and polished cross sections and β autoradiographs were examined to determine the temperature distribution in the UO2. The heat-transfer coefficient h increases with increasing power per unit length. For a specified power, h increased with lower initial clearances. The use of helium instead of argon increased h especially with large clearances, but by a factor much less than the ratio of the thermal conductivities of the gases. Values of h varied widely with lead-bonding; in some positions, h was very large, whereas in others its values were less than for the argon-filled elements. Metallographic examination showed that the lead had moved from some areas of the interface, leaving gaps with poor heat transfer. In the loop elements the grain-growth pattern indicated that some of the heat passed through the lead that had flowed between the pellets. Elements evacuated just before final sealing had values of h equal to or higher than those for argon-filled elements. This is tentatively attributed to the release of natural gases (mainly hydrogen) from the U02 pellets during irradiation, as observed in auxiliary experiments.