STP1405

    Investigation of Temper Embrittlement in Reactor Pressure Vessel Steels Following Thermal Aging, Irradiation, and Thermal Annealing

    Published: Jan 2001


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    Abstract

    The Heavy-Section Steel Irradiation Program at Oak Ridge National Laboratory includes a task to investigate the propensity for temper embrittlement in coarse grain regions of heat-affected zones in prototypic reactor pressure vessel (RPV) steel weldments as a consequence of irradiation and thermal annealing. For the present studies, five prototypic RPV steels with specifications of A302 grade B, A302 grade B (modified), A533 grade B class 1, and A508 class 2 were given two different austenitization treatments and various thermal aging treatments. Thermal aging treatments were conducted at 399, 425, 454 and 490°C for times of 168 and 2000 h. Charpy V-notch impact toughness vs temperature curves were developed for each condition with ductile-brittle transition temperatures used as the basis for comparing the effects of the various heat treatments. Very high austenitization heat treatment produced extremely large grains which exhibited a very high propensity for temper embrittlement following thermal aging. Intergranular fracture was the predominant mode of failure in many of the materials and Auger analysis confirmed significant segregation of phosphorus at the grain boundaries. Lower temperature austenitization treatment performed in a super Gleeble to simulate prototypic coarse grain microstructures in submerged-arc weldments produced the expected grain size with varying propensity for temper embrittlement dependent on the material as well as on the thermal aging temperature and time. Although the lower temperature treatment resulted in decreased propensity for temper embrittlement, the results did provide motivation for the investigation of the potential for phosphorus segregation as a consequence of neutron irradiation and post-irradiation thermal annealing at 454°C. One of the A 302 grade B (modified) steels was given the Gleeble treatment, irradiated at 288°C to about 0.8 × 1019n/cm (>1 MeV) and given a thermal annealing treatment at 454°C for 168 h. Charpy impact testing was conducted on the material in both the irradiated and irradiated/annealed conditions, as well as in the as-received condition. The results show that, although the material exhibited a relatively small Charpy impact 41-J temperature shift, the heat-affected zone-simulated material did exhibit significant intergranular fracture in the post-irradiation annealed condition.

    Keywords:

    reactor pressure vessel, thermal aging, intergranular fracture base metal, temper embrittlement, thermal annealing


    Author Information:

    Nanstad, RK
    Leader, Oak Ridge National Laboratory, Oak Ridge, TN

    McCabe, DE
    Engineer, Oak Ridge National Laboratory, Oak Ridge, TN

    Sokolov, MA
    Metallurgist, Oak Ridge National Laboratory, Oak Ridge, TN

    English, CA
    Principal Technical Consultant and Technical Consultant, AEA-Technology, Harwell, Didcot, Oxfordshire,

    Ortner, SR
    Principal Technical Consultant and Technical Consultant, AEA-Technology, Harwell, Didcot, Oxfordshire,


    Paper ID: STP10544S

    Committee/Subcommittee: E10.07

    DOI: 10.1520/STP10544S


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