Exploratory experiment with the H2O:Zr:UO2 system at high temperature

Three tests were performed using short, single zircalloy-clad simulated fuel rods submerged in stagnant water. Using power gradients believed to be representative of Three Mile Island-2 Reactor, the maximum cladding temperature was 7230C and there was no rod damage. Severe cladding oxidation occurred in the second test using a slightly larger maximum gradient without subsequent power decay. In the third test the power gradient was monotonically increased to 9 kW/m, and the upper region of the rod was destroyed. The lower intact segment of the rod was examined metallographically