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Title: Quantitative measurement and modeling of sensitization development in stainless steel

The state-of-the-art to quantitatively measure and model sensitization development in austenitic stainless steels is assessed and critically analyzed. A modeling capability is evolved and validated using a diverse experimental data base. Quantitative predictions are demonstrated for simple and complex thermal and thermomechanical treatments. Commercial stainless steel heats ranging from high-carbon Type 304 and 316 to low-carbon Type 304L and 316L have been examined including many heats which correspond to extra-low-carbon, nuclear-grade compositions. Within certain limits the electrochemical potentiokinetic reactivation (EPR) test was found to give accurate and reproducible measurements of the degree of sensitization (DOS) in Type 304 and 316 stainless steels. EPR test results are used to develop the quantitative data base and evolve/validate the quantitative modeling capability. This thesis represents a first step to evolve methods for the quantitative assessment of structural reliability in stainless steel components and weldments. Assessments will be based on component-specific information concerning material characteristics, fabrication history and service exposure. Methods will enable fabrication (e.g., welding and repair welding) procedures and material aging effects to be evaluated and ensure adequate cracking resistance during the service lifetime of reactor components. This work is being conducted by the Oregon Graduate Institute with interactive input from personnelmore » at Pacific Northwest Laboratory.« less
Authors:
 [1] ;  [2]
  1. Pacific Northwest Lab., Richland, WA (United States)
  2. Oregon Graduate Inst. of Science and Technology, Beaverton, OR (United States). Dept. of Materials Science and Engineering
Publication Date:
OSTI Identifier:
10186134
Report Number(s):
NUREG/GR--0001
ON: TI93001983
Resource Type:
Technical Report
Resource Relation:
Other Information: PBD: Sep 1992
Research Org:
Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering; Oregon Graduate Inst. of Science and Technology, Beaverton, OR (United States). Dept. of Materials Science and Engineering
Sponsoring Org:
Nuclear Regulatory Commission, Washington, DC (United States)
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; 22 GENERAL STUDIES OF NUCLEAR REACTORS; STAINLESS STEEL-304; MATERIALS TESTING; STAINLESS STEEL-304L; STAINLESS STEEL-316; HEAT TREATMENTS; REACTOR COMPONENTS; FABRICATION; WELDING; WELDED JOINTS; AGING; MECHANICAL PROPERTIES; STRESS CORROSION; INTERGRANULAR CORROSION; COMPUTER CODES 360101; 360103; 220200; PREPARATION AND FABRICATION; MECHANICAL PROPERTIES; COMPONENTS AND ACCESSORIES