Accession Number DE2012-1054496
Title Status Report on Improved Understanding of Creep-Fatigue Damage in Advanced Materials.
Publication Date May 2011
Media Count 33p
Personal Author D. Rink K. Natesan M. Li S. Majumdar W. K. Soppet
Abstract The overall objective of the Advanced Materials Performance Criteria and Methodology 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 licensing of the liquid metal 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 LMR 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 LMR were identified and presented in earlier reports (Natesan et al. 2008, 2009). 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
Licensing
Liquid metal fast reactors
Methodology
Performance evaluation
Pressure vessels
Reactor components
Research and development
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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