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Accession Number DE2012-1054495
Title Improved Creep-Fatigue Models on Advanced Materials for SFR Applications. Final Report.
Publication Date Sep 2011
Media Count 51p
Personal Author D. Rink K. Natesan M. Li S. Majumdar W. K. Soppet
Abstract The overall objective of the Materials Performance Criteria and Methodology work project is to evaluate the key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of advanced structural materials in support of the design and performance of sodium-cooled fast reactors. Advanced materials are a critical element in the development of fast reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced fast reactors. Qualification and licensing of advanced materials are prominent needs for the development and implementation of advanced fast reactor technologies. Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel (B&PV) Code Section III (Rules for Construction of Nuclear Facility Components), and the NRC grants licensing. As the SFRs will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Section III Subsection NH (Class 1 Components in Elevated Temperature Service). A number of technical issues relevant to materials performance criteria and high temperature design methodology in the SFR were identified and presented in earlier reports. A viable approach to resolve these issues and the R&D priority were also recommended. The development of mechanistically based creep-fatigue interaction models for life prediction and reliable data extrapolation was chosen to be the central focus in near-term efforts.
Keywords Creep
Design
Fatigue
Implementation
Methodology
Performance evaluation
Pressure vessels
Reactor components
Research and development
Sodium-cooled fast reactors
Structural materials
Water cooled reactors


 
Source Agency Technical Information Center Oak Ridge Tennessee
NTIS Subject Category 77H - Reactor Engineering & Nuclear Power Plants
71 - Materials Sciences
Corporate Author Argonne National Lab., IL. Nuclear Engineering Div.
Document Type Technical report
Title Note N/A
NTIS Issue Number 1308
Contract Number DE-AC02-06CH11357

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