The objective of the test discussed in this report is to verify further the corrosion rate of irradiated tubular fuel elements under conditions of high specific power and central core temperatures. The fuel will be the inner component of an NPR fuel element assembly. A proposal to charge two tubular fuel elements into the GEH-P7 Loop has been made. The upper fuel element of the two has been irradiated to 1660 MWD/T in a Hanford reactor. A hydraulic cylinder will be attached to the upper fuel element. The hydraulic cylinder will be used to move a chisel along the outer surface of the upper fuel element and shear off a cap which has been welded over a hole through the cladding. The lower fuel element of the two will be used to add heat to the loop so that the desired loop water temperatures can be maintained.

This report provides some temperature and swelling data on Zircaloy-2 clad, coextruded fuel elements irradiated in the KER loops and in the GEH-3 {times} 3P7 loop at the ETR. Data on both rod cluster fuel elements and tubular fuel elements are included. Density changes, where density measurements were made, have also been included. These data are preliminary and are subject to change.