Published 1975 | Version v1
Book Metadata-only

Numerical simulation of tensile behaviour of nuclear fuel cladding

Creators

  • 1. Atomic Energy of Canada Ltd., Chalk River, Ontario. Chalk River Nuclear Labs.

Description

The Zircaloys are the most widely used alloys for nuclear fuel cladding and an accurate description of their response is required for modelling fuel behaviour. A numerical model has been developed which describes the uniaxial true stress-true strain behaviour of irradiated Zircaloy. (Auth.)

Additional details

Publishing Information

Publisher
North-Holland.
Imprint Place
Amsterdam, The Netherlands
Imprint Title
Structural mechanics in reactor technology
Imprint Pagination
v. 1 p. C2/7 1-10.

Conference

Title
3. international conference on structural mechanics in reactor technology.
Dates
1 Sep 1975.
Place
London, UK.

INIS

Country of Publication
Netherlands
Country of Input or Organization
Netherlands
INIS RN
8286431
Subject category
S36: MATERIALS SCIENCE; S99: GENERAL AND MISCELLANEOUS;
Resource subtype / Literary indicator
Conference
Descriptors DEI
DEFORMATION; DISLOCATIONS; FUEL CANS; MATHEMATICAL MODELS; PHYSICAL RADIATION EFFECTS; SIMULATION; STRAIN HARDENING; STRAINS; STRESSES; TENSILE PROPERTIES; ZIRCALOY
Descriptors DEC
ALLOYS; CRYSTAL DEFECTS; CRYSTAL STRUCTURE; HARDENING; LINE DEFECTS; MECHANICAL PROPERTIES; RADIATION EFFECTS; TIN ALLOYS; TRANSITION ELEMENT ALLOYS; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS