Published 1984
| Version v1
Book
Modelling of stress corrosion cracking behavior in zirconium-lined barrier cladding
Creators
- 1. Dept. of Materials Science and Engineering, Stanford Univ., Stanford, CA 94305
Description
The SCCIG model has been modified to predict the stress corrosion cracking behavior of zirconium-lined barrier cladding. The modifications made to the original SCCIG model are very simple and follow the design differences that exist, between barrier and standard cladding, namely that in the SCCIG model 10 percent of the wall thickness at the inner surface of barrier cladding consists of pure zirconium, which is softer and more ductile than Zircaloy substrate
Additional details
Publishing Information
- Publisher
- National Association of Corrosion Engineers.
- Imprint Place
- Houston, TX (USA)
- Imprint Title
- Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors
- Journal Page Range
- p. 851-875.
Conference
- Title
- Environmental degradation of materials in nuclear power systems-water reactors.
- Dates
- 22-24 Aug 1983.
- Place
- Myrtle Beach, SC (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 17022858
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS; S36: MATERIALS SCIENCE; S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BUILDINGS; CORROSION; CRACK PROPAGATION; DUCTILITY; FORECASTING; FUEL CANS; IODINE; MATHEMATICAL MODELS; POWER REACTORS; REACTOR MATERIALS; STRESS CORROSION; SURFACES; THICKNESS; ZIRCALOY; ZIRCONIUM
- Descriptors DEC
- ALLOYS; CHEMICAL REACTIONS; DIMENSIONS; ELEMENTS; HALOGENS; MATERIALS; MECHANICAL PROPERTIES; METALS; NONMETALS; REACTORS; TENSILE PROPERTIES; TRANSITION ELEMENTS; ZIRCONIUM ALLOYS; ZIRCONIUM BASE ALLOYS