Published 2012
| Version v1
Book
Effect of steam oxidation on the tensile properties of PHWR fuel cladding
Creators
- 1. Post Irradiation Examination Division, Bhabha Atomic Research Centre, Mumbai (India)
Description
Fuel cladding tubes made of Zirconium alloys (Zircaloy-2 earlier and Zircaloy-4 presently) are being used in pressurized heavy water reactors (PHWR). The experimental details and the salient features of the methodology which were specifically developed for estimating the transverse tensile properties of the fuel cladding tubes shall be discussed in the paper
Additional details
Publishing Information
- Publisher
- Nuclear Fuel Complex
- Imprint Place
- Hyderabad (India)
- Imprint Title
- Proceedings of the international conference on characterization and quality control of nuclear fuels: inaugural and technical sessions and abstracts
- Imprint Pagination
- 181 p.
- Journal Page Range
- p. 102-103
Conference
- Title
- 4. international conference on characterization and quality control of nuclear fuels
- Acronym
- CQCNF-2012
- Dates
- 27-29 Feb 2012
- Place
- Hyderabad (India)
INIS
- Country of Publication
- India
- Country of Input or Organization
- India
- INIS RN
- 43034422
- Subject category
- S36: MATERIALS SCIENCE;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- DUCTILITY; FUEL-CLADDING INTERACTIONS; PHWR TYPE REACTORS; STRESSES; ZIRCONIUM ALLOYS
- Descriptors DEC
- ALLOYS; HEAVY WATER COOLED REACTORS; HEAVY WATER MODERATED REACTORS; MECHANICAL PROPERTIES; REACTORS; TENSILE PROPERTIES; TRANSITION ELEMENT ALLOYS