Estimation of conservatism of present embrittlement criteria for Zircaloy fuel cladding under LOCA
Creators
- 1. Tokai Research Establishment, Japan Atomic Energy Research Inst., Ibaraki-ken
Description
To estimate the degree of conservatism of the present embrittlement criteria, the following examinations have been conducted. Zircaloy-4 tubes were oxidized at temperatures ranging from 1173 to 1573 K, and then compressed at 373 K to determine the embrittlement due to oxygen absorption. The simulated Zircaloy-clad fuel rods were burst and oxidized in steam at temperatures within the range 1200 to 1500 K. The segment sectioned from each burst cladding was also compressed to determine the effect of hydrogen absorption on cladding embrittlement. The results obtained from these experiments have revealed that the additional embrittlement due to hydrogen absorption is found to be significant in burst cladding. The simulated Zircaloy clad fuel rods were burst and exposed to steam at isothermal temperatures ranging from 1173 to 1573 K and then bottom-flooded at the oxidation temperature or 773 K with water, after the specimen was restrained at the end of exposure time. The specimens were also bottom-flooded under a nonrestrained condition to determine the effect of restraining stress on thermal-shock-failure characteristics. The presence of restraining stress limits the permissible oxidation condition to lower temperature and shorter exposure time. Even the estimated thermal-shock-failure boundary under a restrained condition is located above the 15% oxidation line which is adopted by the present criteria
Additional details
Publishing Information
- Publisher
- American Society for Testing and Materials.
- Imprint Place
- Philadelphia, PA (USA)
- Imprint Title
- Zirconium in the nuclear industry
- Journal Page Range
- p. 734-746.
Conference
- Title
- 6. international conference on zirconium in the nuclear industry.
- Dates
- 28 Jun - 1 Jul 1982.
- Place
- Vancouver, British Columbia (Canada).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 16063441
- Subject category
- S36: MATERIALS SCIENCE; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- ALLOY-ZR98SN-4; BWR TYPE REACTORS; EMBRITTLEMENT; FUEL CANS; FUEL RODS; HYDROGEN ADDITIONS; LOSS OF COOLANT; MEDIUM TEMPERATURE; OXIDATION; OXYGEN; PHYSICAL RADIATION EFFECTS; PWR TYPE REACTORS; REACTOR MATERIALS; REACTOR SAFETY; SIMULATION; THERMAL SHOCK; THERMAL STRESSES; TIME DEPENDENCE; VERY HIGH TEMPERATURE
- Descriptors DEC
- ACCIDENTS; ALLOYS; CHEMICAL REACTIONS; CHROMIUM ADDITIONS; CORROSION RESISTANT ALLOYS; ELEMENTS; ENRICHED URANIUM REACTORS; FUEL ELEMENTS; HEAT RESISTING ALLOYS; IRON ADDITIONS; MATERIALS; NONMETALS; POWER REACTORS; RADIATION EFFECTS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SAFETY; STRESSES; THERMAL REACTORS; TIN ALLOYS; WATER COOLED REACTORS; WATER MODERATED REACTORS; ZIRCALOY; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS