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Server W.L.,ATI Consulting | Hardin T.,EPRI
Effects of Radiation on Nuclear Materials: 26th Volume | Year: 2014

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. Copyright © 2014 by ASTM international. Source


Server W.L.,ATI Consulting | Nanstad R.K.,Oak Ridge National Laboratory
ASTM Special Technical Publication | Year: 2010

The International Atomic Energy Agency (IAEA) has conducted a series of coordinated research projects (CRPs) that have focused on irradiated reactor pressure vessel (RPV) steel fracture toughness properties and approaches for assuring structural integrity of RPVs throughout operating life. A series of nine CRPs has been sponsored by the IAEA, starting in the early 1970s, focused on neutron radiation effects on RPV steels. The purpose of the CRPs was to develop comparisons and correlations to test the uniformity of irradiated results through coordinated international research studies and data sharing. Consideration of dose rate effects, effects of alloying (nickel, manganese, silicon, etc.) and residual elements (e.g., copper and phosphorus), and drop in upper shelf toughness is also important for assessing neutron embrittlement effects. The ultimate use of embrittlement understanding is assuring structural integrity of the RPV under current and future operation and accident conditions. Material fracture toughness is the key ingredient needed for this assessment, and many of the CRPs have focused on measurement and application of irradiated fracture toughness. This paper presents an overview of the progress made since the inception of the CRPs in the early 1970s. The chronology and importance of each CRP have been reviewed and put into context for continued and long-term safe operation of RPVs. Copyright © 2009 by ASTM International. Source


Sokolov M.A.,Oak Ridge National Laboratory | Server W.L.,ATI Consulting | Nanstad R.K.,Oak Ridge National Laboratory
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2015

Some of the current fleet of nuclear power plants is poised to reach their end of life and will require an operating life time extension. Therefore, the main structural components, including the reactor pressure vessel (RPV), will be subject to higher neutron exposures than originally planned. These longer operating times raise serious concerns regarding our ability to manage the reliability of RPV steels at such high doses. Thermal annealing is the only option that can, to some degree, recover irradiated beltline region transition temperature shift and recover upper shelf energy properties lost during radiation exposure and extend RPV service life. This paper reviews the experience accumulated internationally with development and implementation of thermal annealing to RPV and potential perspectives for carrying out thermal annealing on US nuclear power plant RPVs. © Copyright 2015 by ASME. Source


Kirk M.,U.S. Nuclear Regulatory Commission | Stevens G.,U.S. Nuclear Regulatory Commission | Erickson M.,Phoenix Engineering Associates Inc. | Server W.,ATI Consulting | Gustin H.,Structural Integrity Associates
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2015

This paper evaluates current guidance concerning conditions under which the analyst is advised to transition from a linear-elastic fracture mechanics (LEFM) based analysis to an elastic-plastic fracture mechanics (EPFM) based analysis of pressure vessel steels. Current guidance concerning the upper-temperature (Tc) for LEFM-based analysis can be found in ASME Section XI Code Case N-749. Also, while not explicitly stated, an upper-limit on the KIc value that may be used in LEFM-based evaluations is sometimes taken to be 220 MPaVm (a value herein referred to as KLIM). Evaluations of Tc and KLIM were performed using a recently compiled collection of toughness models that are being considered for incorporation into a revision to ASME Section XI Code Case N-830; those models provide a complete definition of all toughness metrics needed to characterize ferritic steel behavior from lower shelf to upper shelf. Based on these evaluations, new definitions of Tc and KLIM are proposed that are fully consistent with the proposed revisions to Code Case N-830 and, thereby, with the underlying fracture toughness data. Formulas that quantify the following values over the ranges of RTTo and RTNDT characteristic of ferritic RPV steels are proposed: • For Tc, two values, Tc(LOWER) and Tc(UPPER), are defined that bound the temperature range over which the fracture behavior of ferritic RPV steels transitions from brittle to ductile. Below Tc(LOWER), LEFM analysis is acceptable while above Tc(UPPER) EPFM analysis is recommended. Between Tc(LOWER) and Tc(UPPER), the analyst is encouraged to consider EPFM analysis because within this temperature range the competition of the fracture mode combined with the details of a particular analysis suggest that the decision concerning the type of analysis is best made on a case-by-case basis. • For KLIM, two values, KLIM(LOWER) and KLIM(UPPER), are defined that bound the range of applied-K over which ductile tearing will begin to occur. At applied-K values below KLIM(LOWER), ductile tearing is highly unlikely, so the use of the KIc curve is appropriate. At applied-K values above KLIM(UPPER), considerable ductile tearing is expected, so the use of the KIc curve is not appropriate. At applied-K values in between KLIM(LOWER) and KLIM(UPPER), some ductile tearing can be expected, so it is recommended to give consideration to the possible effects of ductile tearing as they may impact the situation being analyzed. These definitions of Tc and KLIM better communicate important information concerning the underlying material and structural behavior to the analyst than do current definitions.. Source


Hall B.,Westinghouse | Server W.,ATI Consulting | Rosier B.,Westinghouse | Hardin T.,EPRI
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2013

Uncertainty regarding radiation embrittlement at high fluence, indicative of extended operating life beyond 60 years for current operating pressurized water reactor (PWR) vessels, has been identified as a potential limiting degradation mechanism. There are limited U.S. power reactor surveillance data at fluences greater than about 4E19 n/cm2 (E > 1 MeV) currently available for comparison with existing embrittlement prediction models. Extended operating life to 80 years is projected to have vessel peak fluence approaching 1E20 n/cm2, for a small number of plants. The two current U.S. embrittlement models are contained in Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2, and the Code of Federal Regulations, 10 CFR 50.61a. This paper compares the latest available high fluence power reactor surveillance data to the predictions of these two models, and to another model that has been proposed as better for high fluence data based on combined test reactor and power reactor data from sources extending beyond the U.S. These comparisons indicate the fluence ranges and material groups where the different models deviate from the measured data. The results from these comparisons have been used to select materials for a proposed new PWR supplemental surveillance program (PSSP) that utilizes previously tested irradiated surveillance specimens reconstituted and subsequently re-irradiated to higher fluences. Copyright © 2013 by ASME. Source

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