Electrically Heated Ex-Reactor Pellet-Cladding Interaction (PCI) Simulations Utilizing Irradiated Zircaloy Cladding.

摘要

A series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory primarily to determine the susceptibility of irradiated pressurized-water reactor Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain measuring device, and test matrix. Test results are presented and discussed.