Fracture predictions in Zircaloy fuel cladding
Creators
- 1. Department of Mechanical Engineering, Auburn University, Auburn, AL
Description
Predictions of the maximum initial allowable temperature required to achieve a 40-year life in dry storage are made for Zircaloy clad spent fuel. Maximum initial temperatures of 360 to 4050C for irradiated spent fuel cladding (wet pool storage) are predicted. The technique utilized in this work is based on the deformation and fracture map methodology. Maps are presented for temperatures between 50 and 8500C and stresses between 5 and 500 MPa. These maps are then combined with both the known temperature history (an exponentially decaying one) of Zircaloy fuel cladding in dry storage and a life fracture rule to predict the rupture life of the cladding in dry storage. Predictions of the deformation and fracture map methodology are shown to be in good agreement with constant stress-constant temperature data
Additional details
Publishing Information
- Publisher
- American Society for Testing Materials.
- Imprint Place
- Philadelphia, PA (USA)
- Imprint Title
- Effects of radiation on material. STP 870 Vol. 2
- Journal Page Range
- p. 642-655.
Conference
- Title
- 12. international symposium on effects of radiation on materials.
- Dates
- 18-20 Jun 1984.
- Place
- Williamsburg, VA (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 17049930
- Subject category
- S36: MATERIALS SCIENCE; S42: ENGINEERING; S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- DEFORMATION; DRY STORAGE; FORECASTING; FRACTURE MECHANICS; FRACTURES; FUEL CANS; FUEL-CLADDING INTERACTIONS; HIGH PRESSURE; HIGH TEMPERATURE; IRRADIATION; MAPS; MECHANICAL TESTS; MEDIUM PRESSURE; MEDIUM TEMPERATURE; PHYSICAL RADIATION EFFECTS; SPENT FUELS; STRESSES; TEMPERATURE DEPENDENCE; VERY HIGH PRESSURE; ZIRCALOY
- Descriptors DEC
- ALLOYS; ENERGY SOURCES; FAILURES; FUELS; MATERIALS; MATERIALS TESTING; MECHANICS; NUCLEAR FUELS; RADIATION EFFECTS; REACTOR MATERIALS; STORAGE; TESTING; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS