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ATI Consulting

Pinehurst, NC, United States

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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.


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.


Server W.L.,ATI Consulting | Hardin T.C.,EPRI | Hall J.B.,Westinghouse Electrical Co. | Nanstad R.K.,Oak Ridge National Laboratory
Journal of Pressure Vessel Technology, Transactions of the ASME | Year: 2014

Enhanced 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. Currently, there are limited U.S. power reactor surveillance data available at fluences greater than 4 × 1019 n/cm2 (E > 1 MeV) for comparison with existing embrittlement prediction models. Additional data will be required to support extended operations to 80+ years, where some plants are projected to have peak vessel fluences approaching 1 × 1020 n/cm2. A number of programs are designed to contribute to the high fluence surveillance data to support extended operating life. The U.S programs include the Coordinated PWR Reactor Vessel Surveillance Program (CRVSP), the PWR Supplemental Surveillance Program (PSSP), and the Light Water Reactor Sustainability (LWRS) Program. The LWRS Program involves generation of high fluence test reactor data on many different reactor pressure vessel steels and model alloys, including some of the same PWR vessel materials irradiated to higher fluences in conventional power reactor surveillance programs. This paper surveys the existing high fluence data and the data projected to come from the above listed programs to show when such data will become available. The data will be used to validate or revise embrittlement trend correlations applicable for the high fluence regime. Mechanical property data are being developed, and fine-scale microstructure data are being produced using state-of-the-art methods. © 2014 by ASME.


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..


PubMed | Research Engineer at EnTech Engineering, Shiraz University of Medical Sciences, Shiraz University, ATI Consulting and University of Tabriz
Type: | Journal: Journal of environmental health science & engineering | Year: 2016

Extensive human activities and unplanned land uses have put groundwater resources of Shiraz plain at a high risk of nitrate pollution, causing several environmental and human health issues. To address these issues, water resources managers utilize groundwater vulnerability assessment and determination of protection. This study aimed to prepare the vulnerability maps of Shiraz aquifer by using Composite DRASTIC index, Nitrate Vulnerability index, and artificial neural network and also to compare their efficiency.The parameters of the indexes that were employed in this study are: depth to water table, net recharge, aquifer media, soil media, topography, impact of the vadose zone, hydraulic conductivity, and land use. These parameters were rated, weighted, and integrated using GIS, and then, used to develop the risk maps of Shiraz aquifer.The results indicated that the southeastern part of the aquifer was at the highest potential risk. Given the distribution of groundwater nitrate concentrations from the wells in the underlying aquifer, the artificial neural network model offered greater accuracy compared to the other two indexes. The study concluded that the artificial neural network model is an effective model to improve the DRASTIC index and provides a confident estimate of the pollution risk.As intensive agricultural activities are the dominant land use and water table is shallow in the vulnerable zones, optimized irrigation techniques and a lower rate of fertilizers are suggested. The findings of our study could be used as a scientific basis in future for sustainable groundwater management in Shiraz plain.


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.


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.


Server W.,ATI Consulting | Cipolla R.,Intertek
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2013

The ASME Code, Section XI, has adopted the indirect use of the fracture toughness Master Curve to define an alternative index (RTT0) rather than RTNDT for using the Code KIC and KIa curves in Appendices A and G. RTT0 is defined as T0 + 19.7°C (T0+ 35°F), where T0is the Master Curve reference temperature as defined in ASTM Standard Test Method E 1921. This alternative approach was first approved in ASME Code Case N-629 for Section XI and Code Case N-631 for Section III. Most recently this approach has been integrated directly into the Code, Section XI, and will be published in the 2013 Edition. When this alternative indexing approach was developed, it was recognized that the direct use of the Master Curve itself also could be used as an alternative to the Code KIC curve. A Code Case for the direct use of the fracture toughness Master Curve has been developed and has been presented to Section XI for approval. This paper provides the technical basis for using the fracture toughness Master Curve as an alternative to the Section XI K IC curve. An adjustment to the Master Curve at very low temperatures is included which alleviates a potential problem for low temperature overpressure (LTOP) protection setpoints as would be determined using the existing Code KIC curve. © 2013 by ASME.


Kirk M.,U.S. Nuclear Regulatory Commission | Hein H.,AREVA | Erickson M.,Phoenix Engineering Associates Inc. | Server W.,ATI Consulting | Stevens G.,U.S. Nuclear Regulatory Commission
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2014

In the early 2000s, ASME adopted Code Cases N-629 and N-631 [1- 2], both of which permit the use of the Master Curve reference temperature (To) to define an reference temperature RTTo, as follows (in SI units, as are used throughout the paper): RTTo=To+19.4°C The Code Cases state that "this reference temperature ... may be used as an alternative to [the] indexing reference temperature RTNDT for the KIc and KIa toughness curves, as applicable, in Appendix A and Appendix G [of Section XI of the ASME Code]." KIa is now only used in Appendix A. The functional form of the ASME KIc and KIa curves dictate that the temperature separation between them remains constant irrespective of the degree of neutron radiation embrittlement, as quantified by ΔRTNDT or ΔRTTo. However, data collected from the literature and new data reported by Hein et al. show that radiation embrittlement brings the KIc and KIa curves closer together as embrittlement increases. As a result, current Code guidance will not produce a bounding KIa curve in all situations when RTTo is used as an reference temperature. To reconcile this issue, this paper summarizes available data and, on that basis, concludes that use of the following reference temperature will ensure that the ASME KIa curve bounds currently available KIa data: RTKIa=(RTTo-19.4)+44.97×exp[-0.00613×(RTTo-19.4)] Copyright © 2014 by ASME.


Viehrig H.-W.,Helmholtz Center Dresden | Lucon E.,Belgian Nuclear Research Center | Server W.L.,ATI Consulting
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2010

The Master Curve (MC) approach procedure standardized in ASTM E1921 is defined for quasi-static loading conditions. However, the extension of the MC method to dynamic testing is still under discussion. The effect of loading rate can be broken down into two distinct aspects: 1) the effect of loading rate on Master Curve T0 values for loading rates within the loading rate range specified in ASTM E1921 for quasi-static loading, and 2) the effect of loading rate on Master Curve T0 values for higher loading rates. The IAEA CRP8 includes both aspects, but primarily focuses on the second element of loading rate effects, i.e. loading rate ranges above the upper limit of the E1921 standard and it comprises: - results of a round-robin exercise to validate the application of the Master Curve approach to precracked Charpy (PCC) specimens tested in the ductile-to-brittle transition region using an instrumented pendulum, - Master Curve data obtained at different loading rates on various RPV steels, in order to assess the loading rate dependence of T 0 and compare it with an empirical model proposed by Wallin, and - the comparison of results from unloading compliance and monotonic loading in the quasi-static range. Copyright © 2009 by ASME.

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