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

Randy K. Nanstad, Donald E. McCabe, Mikhail A. Sokolov, Colin A. English, Susan R. Ortner

Research output: Chapter in Book/Report/Conference proceedingConference contributionpeer-review

6 Scopus citations

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 (RP V) 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 x 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.

Original languageEnglish
Title of host publicationEffects of Radiation on Materials
Subtitle of host publication20th International Symposium
EditorsStan T. Rosinski, Martin L. Grossbeck, Todd R. Allen, Arvind S. Kumar
PublisherASTM International
Pages356-382
Number of pages27
ISBN (Electronic)9780803128781
DOIs
StatePublished - 2001
Event20th International Symposium on Effects of Radiation on Materials 2000 - Williamsburg, United States
Duration: Jun 6 2000Jun 8 2000

Publication series

NameASTM Special Technical Publication
VolumeSTP 1405
ISSN (Print)0066-0558

Conference

Conference20th International Symposium on Effects of Radiation on Materials 2000
Country/TerritoryUnited States
CityWilliamsburg
Period06/6/0006/8/00

Funding

This research sponsored by the Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, under Interagency Agreement DOE 1886-N695-3W with the U.S. Department of Energy under Contract No. DE-AC05-00OR22725 with UT-Battelle, LLC. The authors are indebted to the late Michael Vassilaros of the NRC who assisted with the evaluations of this work during the aging studies. Assistance at ORNL of the following individuals is also greatly appreciated: Alan Frederick for Gleeble austenitization, Edward Hatfield for heat treatment, Janie Gardner and Jackie Mayotte for metallography, Ronald Swain for mechanical tests, James King for welding technology guidance, and Julia Bishop for preparation of the manuscript.

FundersFunder number
Office of Nuclear Regulatory Research
U.S. Department of EnergyDE-AC05-00OR22725
U.S. Nuclear Regulatory Commission
National Research Council Canada

    Keywords

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

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