Advisory engineer, Babcock & Wilcox Co., Nuclear Power Division, Lynchburg, VA
Senior research engineer, Babcock & Wilcox Co., Lynchburg Research Center, Lynchburg, VA
Pages: 11 Published: Jan 1987
The Babcock & Wilcox (B&W) Owners Group Integrated Reactor Vessel Surveillance Program contained two irradiation phases in addition to the plant specific surveillance programs. One phase provided materials for a short-term (test reactor) study of the effects of neutron irradiation on high-copper low-upper-shelf weld metals as a part of the Heavy Section Steel Technology (HSST) Program. This phase was followed with an irradiation study using the long-term (power reactor) radiation source.
These studies were distinguished from other irradiation studies in that production weld metals were used. The chemical composition of the weld metals were similar except for the copper content which ranged from 0.21 to 0.42%. The specimens used for mechanical properties evaluation consisted of tension, Charpy V-notch, and compact fracture specimens.
The power reactor irradiations were conducted in B&W Owner's Group host reactors. The fluence attained in both the test and power reactors was approximately 7 × 1018 neutrons (n)/cm2. These two neutron sources provide a means for evaluating the influence of neutron environment on material damage.
The Charpy impact toughness is evaluated by the 41 J (30 ft · lb) shift and decrease in upper-shelf energy. The tension data is evaluated using the standard mechanical properties. The fracture toughness data obtained from compact fracture specimens, using J-R elastic-plastic fracture test techniques, was evaluated.
These comparisons indicate that test reactor irradiated yield strength data and fracture toughness data is equivalent to that of a power reactor. A comparison of Charpy properties indicates that the data may be sensitive to parameters other than neutron environments.
power reactor, test reactor, submerged-arc welds, radiation damage, Charpy impact data, tension test data, compact fracture data, fracture toughness, reactor vessel
Paper ID: STP25661S