Thermal-aged surveillance material was removed from two commercial reactor pressure vessels. The material from the first reactor vessel received a thermal exposure of approximately 103,000 hours at 282°C (540°F), while the material from the second reactor vessel received a thermal exposure of approximately 93,000 hours at 282°C (540°F). The surveillance material removed consisted of SA-508, Class 2 and SA-533, Grade B, Class 1, base metals; Mn-Mo-Ni/Linde 80 flux weld metals; and SA-533, Grade B, Class 1, correlation monitor material. Charpy V-notch impact specimens were fabricated from these materials and tested. In addition, selected thermal-aged specimens were annealed at 343°C (650°F) and 454°C (850°F) and impact tested. Unaged material was tested to establish baseline data for evaluating the thermal-aged material and thermal-aged/annealed material. Small variations in the impact properties were observed for all materials. Small increases in transition temperature were observed for the forging base metal and weld metal B surveillance materials, while the upper-shelf energies exhibited small decreases. The 454°C (850°F) anneal restored the upper-shelf energy for the forging base metal and weld metal B surveillance materials.
Thermal-aged surveillance materials consisting of SA-533, Grade B, Class 1 plate material; SA-508, Class 2 forging material; and 2 Mn-Mo-Ni/Linde 80 weld metals were removed from two commercial reactor pressure vessels. The material from the first reactor vessel received a thermal exposure of approximately 103,000 hours at 282°C (540°F), while the material from the second reactor vessel received a thermal exposure of approximately 93,000 hours at 282°C (540°F). Tensile and 1/2T compact fracture toughness specimens were fabricated from these materials and tested. In addition, to examine the effects of annealing, selected thermal-aged and unaged specimens were annealed at 454°C (850°F) and tested. Varying responses in the fracture toughness properties were observed for all materials after exposure to the thermal-aging temperature. The base metal plate had an observed decrease in J-values after its respective aging exposure, while no significant difference in the J-values were observed for the Linde 80 weld metals. No significant difference was seen in the J-data for the aged/annealed materials, but because of the small number of test specimens available, no conclusion could be determined for the response to annealing.
The B&W Owners Group (B&WOG) developed the Master Integrated Reactor Vessel Surveillance Program (MIRVP) to encompass all domestic operating pressurized water reactors (PWRs) with reactor vessels that contain Linde 80 submerged-arc welds. These PWRs include the reactor vessels fabricated by Babcock & Wilcox (B&W) that include B&W-designed 177-Fuel Assembly (FA) plants and Westinghouse-designed plants. The B&WOG Reactor Vessel Working Group has sponsored studies over the past several years to develop an understanding of the mechanisms and trends of radiation embrittlement for reactor vessel submerged-arc welds fabricated using Linde 80 weld flux. As part of the MIRVP, a Westinghouse-designed capsule was included to provide a comparison of irradiation data in the Westinghouse neutronic environment with the B&W 177-FA environment. This paper presents the effect of irradiation environment on the embrittlement trends of Linde 80 submerged-arc welds based on mechanical test data available through the B&WOG and the literature, and existing trend curves.
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