Reassessment of the basis for NRC fuel damage criteria for reactivity transients

Conference · Sat Oct 01 00:00:00 EDT 1994

OSTI ID:54619

McCardell, R K ^[1]

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1. Idaho National Engineering Lab., Idaho Falls, ID (United States)

The present basis for NRC Fuel Damage Criteria was obtained from experiments performed in the Special Power Excursion Reactor Test (SPERT) IV Reactor Capsule Driver Core (CDC) at the Idaho National Engineering Laboratory (INEL) between 1967 and 1970. Most of the CDC test fuel rods were previously unirradiated and the failure threshold for these unirradiated fuel rods was measured to be about 200 calories per gram of UO{sub 2} radially averaged fuel enthalpy at the axial peak.

• Research Organization: US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research
• OSTI ID: 54619
• Report Number(s): NUREG/CP-0139; CONF-9410216-; ON: TI95001469; TRN: 95:012653
• Resource Relation: Conference: 22. water reactor safety information meeting, Bethesda, MD (United States), 24-26 Oct 1994; Other Information: PBD: Oct 1994; Related Information: Is Part Of Transactions of the twenty-second water reactor safety information meeting; PB: 129 p.
• Country of Publication: United States
• Language: English

Related Subjects

21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS
22 NUCLEAR REACTOR TECHNOLOGY
FUEL ELEMENTS
DAMAGE
NUCLEAR FUELS
REACTIVITY
NUCLEAR POWER PLANTS
REACTOR SAFETY
BURNUP