Large-break LOCA, in-reactor fuel bundle Materials Test MT-6A

Technical Report · Wed Sep 01 00:00:00 EDT 1993

DOI:https://doi.org/10.2172/10186908· OSTI ID:10186908

Wilson, C L; Hesson, G M; Pilger, J P; King, L L; Panisko, F E

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuel bundle is cooled.

• Research Organization: Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC06-76RL01830
• OSTI ID: 10186908
• Report Number(s): PNL-8829; ON: DE94000499; TRN: 93:022992
• Resource Relation: Other Information: PBD: Sep 1993
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
PWR TYPE REACTORS
LOSS OF COOLANT
REACTOR SAFETY
HEAT TRANSFER
HYDRAULICS
NRU REACTOR
SIMULATION
TESTING
FUEL RODS
FUEL CANS
RUPTURES
220900
210200
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED