Mechanical Properties of Thermally Aged Cast Stainless Steels from Shippingport Reactor Components (NUREG/CR-6275, ANL-94/37)

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Publication Information

Manuscript Completed: July 1994
Date Published: April 1995

Prepared by:
O.K. Chopra and W.J. Shack

Argonne National Laboratory
9700 South Cass Avenue
Argonne, Illinois 60439

Prepared for:
Division of Engineering
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC FIN A-2256

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Abstract

Thermal embrittlement of static-cast CF-8 stainless steel components from the decommissioned Shippingport reactor has been characterized. Cast stainless steel materials were obtained from four cold-leg check valves, three hot-leg main shutoff valves, and two pump volutes. The actual time-at-temperature for the materials was ≈13 y at ≈281°C (538°F) for the hot-leg components and ≈264°C (507°F) for the cold-leg components. Baseline mechanical properties for as-cast material were determined from tests on either recovery-annealed material, i.e., annealed for 1 h at 550°C and then water quenched, or material from the cooler region of the component. The Shippingport materials show modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength because of relatively low service temperatures and ferrite content of the steel. The procedure and correlations developed at Argonne National Laboratory for estimating mechanical properties of cast stainless steels predict accurate or slightly lower values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J_IC of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predicted the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of ≈15 y and the KRB reactor pump cover plate (CF-8) after ≈8 y of service.