Abstract
To construct advanced non-light water reactors (ANLWRs) operating in the temperature range above that for the traditional light water reactors (LWRs), Alloy 316H is one of the candidate materials because of its inexpensiveness, significant service experience, and qualification for nuclear applications by the American Society of Mechanical Engineers (ASME). However, during the life span at temperatures expected for the ANLWRs, the alloy is likely to experience thermal embrittlement that has not been a concern for the traditional LWRs. To prepare for the development, the possibility of adverse thermal embrittlement effects on Alloy 316H performance in the ANLWRs must be evaluated and a technical basis regarding thermal embrittlement, if necessary, must be established for structural integrity analysis to provide reasonable assurance of adequate nuclear safety protection. In this paper, current technical basis for nuclear applications of Alloy 316H deterioration from thermal aging is briefly introduced. The likelihood of adverse thermal embrittlement effects on Alloy 316H performance is evaluated through historical data on microstructural and mechanical property evolution. Characterization of thermal embrittlement is then discussed, followed by a review of predictive models and trend curves for alloy embrittlement. Based on the review and evaluation, technical gaps for addressing thermal embrittlement issues are identified and gap-filling actions are recommended for establishing a technical basis to enable adequate consideration of thermal embrittlement in Alloy 316H applications to the ANLWRs.
| Original language | English |
|---|---|
| Title of host publication | Materials and Fabrication |
| Publisher | American Society of Mechanical Engineers (ASME) |
| ISBN (Electronic) | 9780791858981 |
| DOIs | |
| State | Published - 2019 |
| Event | ASME 2019 Pressure Vessels and Piping Conference, PVP 2019 - San Antonio, United States Duration: Jul 14 2019 → Jul 19 2019 |
Publication series
| Name | American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP |
|---|---|
| Volume | 6B-2019 |
| ISSN (Print) | 0277-027X |
Conference
| Conference | ASME 2019 Pressure Vessels and Piping Conference, PVP 2019 |
|---|---|
| Country/Territory | United States |
| City | San Antonio |
| Period | 07/14/19 → 07/19/19 |
Funding
Boiler and Pressure Vessel Code Clinch River Breeder Reactor Ductile-to-Brittle Transition Temperature Department of Energy Electric Power Research Institute Intergranular Light Water Reactor Minimum Creep Rate National Institute of Materials Science Nuclear Regulatory Commission Oak Ridge National Laboratory Power Reactor Innovative Small Module Pressurized Thermal Shock Reduction of Area Reactor Pressure Vessel Power Reactor Innovative Small Module Total Elongation Uniform Elongation United States Ultimate Tensile Strength Yield Strength This manuscript has been authored by UT-Battelle, LLC, under contract DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a nonexclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. DOE will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).