Development of new cladding types for nuclear fuel: IOP Conference Series: Materials Science and Engineering

Research output: Contribution to report/book/conference proceedingsIn-proceedings paper

Authors

Institutes & Expert groups

  • MTA - Centre for Energy Research
  • KAERI - Korea Atomic Energy Research Institute - Korea
  • Czech Technical University in Prague - Faculty of Nuclear Sciences and Physical Engineering

Documents & links

Abstract

Three different cladding types were tested for nuclear fuel in traditional light water reactors and generation IV gas-cooled fast reactors. Cr coated Zr cladding was tested in steam atmosphere up to 1200 °C to demonstrate moderate oxidation and hydrogen production in accident conditions. 15-15Ti stainless steel alloy and SiCf/SiC cladding tube samples were treated in helium atmosphere with different impurities for several hours at 1000 °C. Additional mechanical testing and microstructure examinations were carried out with as-received samples and with specimens after high temperature treatments. The experiments results indicated the applicability of the tested materials for reactor conditions in the investigated range of parameters.

Details

Original languageEnglish
Title of host publicationIOP Conference Series: Materials Science and Engineering
Subtitle of host publication12th Hungarian Conference on Materials Science (HMSC12) 13-15 October 2019, Balatonkenese, Hungary
Pages1-9
Number of pages9
Volume903
DOIs
Publication statusPublished - 4 Dec 2020

Publication series

NameIOP Conference Series: Materials Science and Engineering
PublisherIOP Publishing Ltd
Volume903
ISSN (Print)1757-899X

Keywords

  • Cladding, Coating, Chromium, Oxidation

ID: 6987834