Abstract
The Materials Management and Minimization (M3) Program intends to qualify a new high‑density low‑enriched‑uranium (LEU) U–Mo monolithic fuel to enable conversion of five US high‑performance research reactors (USHPRRs). This thesis presents the preliminary results and discussions related to post-irradiation blister anneal studies and fission product release scoping studies performed on U–Mo monolithic fuel plates.
Blister anneal testing on irradiated fuel plates is a temperature‑resolved failure‑threshold measurement technique historically used to assess fuel plate stability under off-normal operating conditions. The effects of fuel composition, geometry, fission density, and irradiation conditions are presented herein as parameters that were investigated for their impact on blister‑threshold temperatures. The fission‑product‑transport scoping study successfully characterized the release, transport and temperature‑resolved deposition behavior of iodine and cesium. Two failure temperatures were evaluated: 600 and 1250C. Testing was performed in the main hot cell at the Materials and Fuels Complex located at Idaho National Laboratory.