Published October 1994
| Version v1
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Reassessment of the basis for NRC fuel damage criteria for reactivity transients
Description
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 UO2 radially averaged fuel enthalpy at the axial peak
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Additional details
Publishing Information
- Imprint Title
- Transactions of the twenty-second water reactor safety information meeting
- Imprint Pagination
- 129 p.
- Journal Page Range
- p. 113-114.
- Report number
- NUREG/CP--0139
Conference
- Title
- 22. water reactor safety information meeting.
- Dates
- 24-26 Oct 1994.
- Place
- Bethesda, MD (United States).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 26067116
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BURNUP; FUEL ELEMENTS; NUCLEAR FUELS; NUCLEAR POWER PLANTS; REACTIVITY; REACTOR SAFETY
- Descriptors DEC
- ENERGY SOURCES; FUELS; MATERIALS; NUCLEAR FACILITIES; POWER PLANTS; REACTOR COMPONENTS; REACTOR MATERIALS; SAFETY; THERMAL POWER PLANTS
Optional Information
- Secondary number(s)
- CONF-9410216--.