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Title: A REVIEW OF AGING EFFECTS IN ALLOY 617 FOR GEN IV NUCLEAR REACTOR APPLICATIONS

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 long-time, 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.

Authors:
 [1];  [1]
  1. ORNL
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1093003
DOE Contract Number:  
DE-AC05-00OR22725
Resource Type:
Conference
Resource Relation:
Conference: 2006 ASME Pressure Vessels and Piping Division Conference, Vancouver, British Columbia, Can, Canada, 20060723, 20060727
Country of Publication:
United States
Language:
English

Citation Formats

Ren, Weiju, and Swindeman, Robert W. A REVIEW OF AGING EFFECTS IN ALLOY 617 FOR GEN IV NUCLEAR REACTOR APPLICATIONS. United States: N. p., 2006. Web.
Ren, Weiju, & Swindeman, Robert W. A REVIEW OF AGING EFFECTS IN ALLOY 617 FOR GEN IV NUCLEAR REACTOR APPLICATIONS. United States.
Ren, Weiju, and Swindeman, Robert W. Sun . "A REVIEW OF AGING EFFECTS IN ALLOY 617 FOR GEN IV NUCLEAR REACTOR APPLICATIONS". United States. doi:.
@article{osti_1093003,
title = {A REVIEW OF AGING EFFECTS IN ALLOY 617 FOR GEN IV NUCLEAR REACTOR APPLICATIONS},
author = {Ren, Weiju and Swindeman, Robert W},
abstractNote = {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 long-time, 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.},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Sun Jan 01 00:00:00 EST 2006},
month = {Sun Jan 01 00:00:00 EST 2006}
}

Conference:
Other availability
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