Fuel-cladding mechanical interaction in PCI-resistant LWR fuel designs during normal operation and power ramping
Journal Article
·
· Nucl. Technol.; (United States)
OSTI ID:7144794
The fuel cladding mechanical interaction behavior of developmental fuel rods irradiated in the Halden Boiling Water Reactor was evaluated based primarily on rod elongation measurements made during steadystate and power-ramping irradiation. The developmental fuel rod designs were selected based on attributes that were expected to reduce pellet-cladding interaction (PCI) failures during irradiation. Testing results were compared to a nonpressurized reference design with dished-pellet fuel. For the reference rods, there was a relationship between thermal feedback and fuelcladding mechanical interaction during steady-state irradiation. Significant cladding stresses developed in both the axial and hoop directions in the reference rods during power ramping. During power ramping the general cladding stress distribution in fuel rods with annular fuel pellets was primarily axial while cladding stresses in rods with sphere-pac fuel were mostly in the hoop direction. These results are indicative of superior PCI resistance in the annular pellet fuel rod designs when compared to the reference and sphere-pac rods.
- Research Organization:
- Pacific Northwest Laboratory, P.O. Box 999, Richland, Washington 99352
- OSTI ID:
- 7144794
- Journal Information:
- Nucl. Technol.; (United States), Journal Name: Nucl. Technol.; (United States) Vol. 63:1; ISSN NUTYB
- Country of Publication:
- United States
- Language:
- English
Similar Records
Fuel-cladding mechanical interaction in PCI-resistant LWR fuel designs during normal operation and power ramping
FUEL PERFORMANCE IMPROVEMENT PROGRAM Power-Ramp Testing and Postirradiation Examination of PCI-Resistant LWR Fuel Rod Designs
Fuel performance improvement program. Quarterly/annual progress report, October 1978-September 1979
Journal Article
·
Sat Oct 01 00:00:00 EDT 1983
· Nucl. Technol.; (United States)
·
OSTI ID:6477602
FUEL PERFORMANCE IMPROVEMENT PROGRAM Power-Ramp Testing and Postirradiation Examination of PCI-Resistant LWR Fuel Rod Designs
Technical Report
·
Wed Sep 01 00:00:00 EDT 1982
·
OSTI ID:1084094
Fuel performance improvement program. Quarterly/annual progress report, October 1978-September 1979
Technical Report
·
Mon Oct 01 00:00:00 EDT 1979
·
OSTI ID:5605431
Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
AXIAL SYMMETRY
BWR TYPE REACTORS
DESIGN
EVALUATION
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
IRRADIATION
REACTOR COMPONENTS
REACTORS
STEADY-STATE CONDITIONS
STRESSES
SYMMETRY
WATER COOLED REACTORS
WATER MODERATED REACTORS
210100* -- Power Reactors
Nonbreeding
Light-Water Moderated
Boiling Water Cooled
AXIAL SYMMETRY
BWR TYPE REACTORS
DESIGN
EVALUATION
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
IRRADIATION
REACTOR COMPONENTS
REACTORS
STEADY-STATE CONDITIONS
STRESSES
SYMMETRY
WATER COOLED REACTORS
WATER MODERATED REACTORS