Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding. [PWR]

Technical Report · Fri Feb 01 00:00:00 EST 1985

DOI:https://doi.org/10.2172/6021014· OSTI ID:6021014

Barner, J. O.; Fitzsimmons, D. E.

In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) 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.

• Research Organization: Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 6021014
• Report Number(s): NUREG/CR-3999; PNL-5245; ON: TI85007915; TRN: 85-006261
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
FUEL CANS
PHYSICAL RADIATION EFFECTS
FUEL PELLETS
FUEL-CLADDING INTERACTIONS
PWR TYPE REACTORS
FUEL RODS
FUEL SYSTEMS
SIMULATION
SYSTEM FAILURE ANALYSIS
TESTING
ZIRCALOY
ALLOYS
FUEL ELEMENTS
PELLETS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
SYSTEMS ANALYSIS
TIN ALLOYS
WATER COOLED REACTORS
WATER MODERATED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
210200* - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled