Interaction between UO/sub 2/ pellets and cladding in water reactor fuel pins

Journal Article · Sat Jan 01 00:00:00 EST 1977 · J. Am. Ceram. Soc.; (United States)

DOI:https://doi.org/10.1111/j.1151-2916.1977.tb15516.x· OSTI ID:6355069

Waddington, J S; Roberts, G; Dickson, I K

The contribution of mechanical interaction between fuel and cladding to the change in length of a fuel pin has been identified from an analysis of data from the Steam-Generating Heavy-Water Reactor. A model which accounts for this effect uses the creep properties of fuel and cladding.

• Research Organization: Central Electricity Generating Board, Berkeley, Eng.
• OSTI ID: 6355069
• Journal Information: J. Am. Ceram. Soc.; (United States), Vol. 60:9-10, Issue 9-10
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
36 MATERIALS SCIENCE
SGHWR REACTOR
FUEL-CLADDING INTERACTIONS
URANIUM DIOXIDE
PHYSICAL RADIATION EFFECTS
ZIRCALOY
CREEP
CRYSTAL STRUCTURE
ELONGATION
FUEL CANS
FUEL PINS
ACTINIDE COMPOUNDS
ALLOYS
CHALCOGENIDES
ENRICHED URANIUM REACTORS
FUEL ELEMENTS
HEAVY WATER MODERATED REACTORS
MECHANICAL PROPERTIES
OXIDES
OXYGEN COMPOUNDS
POWER REACTORS
RADIATION EFFECTS
REACTOR COMPONENTS
REACTORS
THERMAL REACTORS
TIN ALLOYS
URANIUM COMPOUNDS
URANIUM OXIDES
WATER COOLED REACTORS
ZIRCONIUM ALLOYS
ZIRCONIUM BASE ALLOYS
210400* - Power Reactors
Nonbreeding
Otherwise Moderated or Unmoderated
360106 - Metals & Alloys- Radiation Effects