The precracked Charpy single-edge notched bend, SE(B), specimen (PCC) is the most likely specimen type to be used for determination of the reference temperature, T0, with reactor pressure vessel (RPV) surveillance specimens. Unfortunately, for many RPV steels, significant differences have been observed between the T0 temperature for the PCC specimen and that obtained from the 25-mm thick compact specimen [1TC(T)], generally considered the standard reference specimen for T0. This difference in T0 has often been designated a specimen bias effect, and the primary focus for explaining this effect is loss of constraint in the PCC specimen. The International Atomic Energy Agency (IAEA) has developed a coordinated research project (CRP) to evaluate various issues associated with the fracture toughness Master Curve for application to light-water RPVs. Topic Area 1 of the CRP is focused on the issue of test specimen geometry effects, with emphasis on determination of T0 with the PCC specimen and the bias effect. Topic Area 1 has an experimental part and an analytical part. Participating organizations for the experimental part of the CRP performed fracture toughness testing of various steels, including the reference steel JRQ (A533-B-1) often used for IAEA studies, with various types of specimens under various conditions. Additionally, many of the participants took part in a round robin exercise on finite element modeling of the PCVN specimen, discussed in a separate paper. Results from fracture toughness tests are compared with regard to effects of specimen size and type on the reference temperature T0. It is apparent from the results presented that the bias observed between the PCC specimen and larger specimens for Plate JRQ is not nearly as large as that obtained for Plate 13B (−11°C vs −37°C) and for some of the results in the literature (bias values as much as −45°C). This observation is consistent with observations in the literature that show significant variations in the bias that are dependent on the specific materials being tested. There are various methods for constraint adjustments and two methods were used that reduced the bias for Plate 13B from −37°C to −13°C in one case and to − 11°C in the second case. Unfortunately, there is not a consensus methodology available that accounts for the differences observed with different materials. Increasing the Mlim value in the ASTM E-1921 to ensure no loss of constraint for the PCC specimen is not a practicable solution because the PCC specimen is derived from CVN specimens in RPV surveillance capsules and larger specimens are normally not available. Resolution of these differences are needed for application of the master curve procedure to operating RPVs, but the research needed for such resolution is beyond the scope of this CRP.

There is a lack of pressurized water reactor (PWR) surveillance program transition temperature shift and upper shelf toughness decrease data due to neutron irradiation exposure especially at high fluences indicative of 60 to 80 years of plant operation. The Electric Power Research Institute (EPRI) has funded the development of a supplemental reactor pressure vessel (RPV) surveillance program to allow testing of additionally irradiated specimens in two new capsules being installed in two different commercial reactor surveillance capsule positions. The previously irradiated materials were strategically selected and will be further irradiated to give final fluence levels equal to or above those for PWRs operating up to 80 years. This paper describes the final design of the capsules and selection of the key previously irradiated RPV materials reconstituted into new Charpy-size specimens being irradiated in the two PWR Supplemental Surveillance Program (PSSP) capsules.

In NUREG/CR-6923, "Expert Panel Report on Proactive Materials Degradation Assessment," referred to as the PMDA report, NRC conducted a comprehensive evaluation of potential agingrelated degradation modes for core internal components, as well as primary, secondary, and some tertiary piping systems, considering operation up to 40 years. This document has been a very valuable resource, supporting NRC staff evaluations of licensees' aging management programs and allowing for prioritization of research needs. v FOREWORDAccording to the provisions of Title 10 of the Code of Federal Regulations (CFR), Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants," licensees may apply for twenty-year renewals of their operating license following the initial forty-year operating period. The majority of plants in the United States have received the first license renewal to operate from forty to sixty years and a number of plants have already entered the period of extended operation. Therefore, licensees are now assessing the economic and technical viability of a second license renewal to operate safely from sixty to eighty years. The requirements of 10 CFR, Part 54 include the identification of passive, long-lived structures, systems, and components which may be subject to aging-related degradation, and the development of aging management programs (AMPs) to ensure that their safety function is maintained consistent with the licensing basis during the extended operating period. NRC guidance on the scope of AMPs is found in NUREG-1800 "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants" (SRP-LR) and NUREG-1801, "Generic Aging Lessons Learned (GALL) Report."In anticipation to review applications for reactor operation from sixty to eighty years, the Office of Nuclear Reactor Regulation (NRR) requested the Office of Nuclear Regulatory Research (RES) to conduct research and identify aging-related degradation scenarios that could be important in this timeframe, and to identify issues for which enhanced aging management guidance may be warranted and allowing for prioritization of research needs. As part of this effort, RES agreed to a Memorandum of Understanding with the U.S. Department of Energy (DOE) to jointly develop an Expanded Materials Degradation Assessment (EMDA) at Oak Ridge National Laboratory (ORNL). The EMDA builds upon work previously done by RES in NUREG/CR-6923, "Expert Panel Report on Proactive Materials Degradation Assessment." Potential degradation scenarios for operation up to forty years were identified using an expert panel to develop a phenomena identification and ranking table (PIRT). NUREG/CR-6923 mainly addressed primary system and some secondary system components. The EMDA covers a broader range of components, including piping systems and core internals, reactor pressure vessel, electrical cables, and concrete structures. To conduct the PIRT and to prepare the EMDA report, an expert panel for each of the four component groups was assembled. The pa...

A pressurized water reactor (PWR) supplemental surveillance program (PSSP) is being designed to provide high-fluence reactor vessel material embrittlement data for the operating U.S. PWRs. Peak reactor pressure vessel (RPV) fluence levels as high as 7 × 1019n/cm2 (E > 1.0 MeV) will be attained as PWRs operate to 60 years and potentially beyond. Therefore, a need exists to obtain high-fluence PWR surveillance data to validate or revise embrittlement trend correlations (ETC) applicable for the high-fluence regime. Without the availability of high-fluence PWR surveillance data, it may be necessary to use an overly conservative ETC, or an ETC with a high margin at high fluence, which could constrain plant pressure–temperature operating curves, increasing startup and shutdown times and costs or increasing the potential of exceeding the pressurized thermal shock screening limit. The PSSP is designed to supplement data produced by the existing 10 CFR 50 Appendix H surveillance programs. The two proposed PSSP capsules will contain Charpy specimens reconstituted from tested PWR surveillance capsule materials, carefully selected for material type, chemistry, and fluence to optimize future ETC development. These capsules will be inserted into one or two U.S.-based Westinghouse-designed operating nuclear power plants for continued irradiation. The selected host plant(s) have relatively high capsule irradiation flux locations, enabling production of high-fluence data prior to the U.S. plants reaching 60 years of operation. The PSSP capsule irradiation will increase the fluence levels up to 1 × 1020n/cm2 on select groups of reactor vessel materials. This paper describes the basis for the PSSP, plant selection for irradiation, and material selection.