A review of aging effects in alloy 617 for Gen IV nuclear reactor applications

Weiju Ren, Robert Swimdeman

Research output: Chapter in Book/Report/Conference proceedingConference contributionpeer-review

8 Scopus citations

Abstract

The literature was reviewed of aging and aging effects in Alloy 617 to determine the supplementary data needed to understand the response of the alloy to long-time exposure conditions being considered for structural components in Gen IV nuclear reactors. Most of the data were produced in connection with the international research effort supporting High Temperature Gas-Cooled Reactor (HTGR) projects in the 1970s and 1980s. Topics considered included microstructural changes, hardness, tensile properties, toughness, creep-rupture, fatigue, and crack growth. It became clear that, for the longtime, very high temperature conditions of the Gen IV reactors, a significant effort would be needed to fully understand and characterize property changes. Several topics for further research were recommended.

Original languageEnglish
Title of host publicationProceedings of 2006 ASME Pressure Vessels and Piping Division Conference - ASME PVP2006/ICPVT-11 Conference - Pressure Vessel Technologies for the Global Community
DOIs
StatePublished - 2006
EventASME PVP2006/ICPVT-11 Conference - Vancouver, BC, Canada
Duration: Jul 23 2006Jul 27 2006

Publication series

NameAmerican Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP
Volume2006
ISSN (Print)0277-027X

Conference

ConferenceASME PVP2006/ICPVT-11 Conference
Country/TerritoryCanada
CityVancouver, BC
Period07/23/0607/27/06

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