NRC Welding Residual Stress Validation Program International Round Robin Program and Findings

2011 ◽  
Author(s):  
Howard J. Rathbun ◽  
Lee F. Fredette ◽  
Paul M. Scott ◽  
Aladar A. Csontos ◽  
David L. Rudland
Keyword(s):  
Residual Stress ◽  
Degradation Mechanism ◽  
Nuclear Regulatory Commission ◽  
Round Robin ◽  
Welding Residual Stress ◽  
Pressurized Water ◽  
Primary Driver ◽  
International Round ◽  
Standard Deviations ◽  
Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress (WRS) predictions in pressurized water reactor (PWR) primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in PWRs are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are the primary driver of this degradation mechanism. The NRC/EPRI weld residual stress (WRS) analysis validation program consists of four phases, with each phase increasing in complexity from laboratory size specimens to component mock-ups and cancelled-plant material. This paper discusses Phase 2 of the WRS characterization program involving an international round robin analysis project in which participants analyzed a prototypic reactor coolant pressure boundary component. Mock-up fabrication, WRS measurements and comparison with predicted stresses through the DM weld region are described. The results of this study show that, on average, analysts can develop WRS predictions that are a reasonable estimate for actual configurations as quantified by measurements. However, the scatter in predicted results from analyst to analyst can be quite large. For example, in this study, the scatter in WRSs through the centerline of the main DM weld (prior to stainless steel weld application) predicted by analysts is approximately +/− 200 to 300 MPa at 3 standard deviations for axial stresses and +/− 300 to 400 MPa at 3 standard deviations for hoop stresses. Sensitivity studies that vary important parameters, such as material hardening behavior, can be used to bound such large variations.

Summary of Finite Element (FE) Sensitivity Studies Conducted in Support of the NRC/EPRI Welding Residual Stress (WRS) Program

2012 ◽  
Author(s):  
Matthew Kerr ◽  
Howard J. Rathbun
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Degradation Mechanism ◽  
Nuclear Regulatory Commission ◽  
Round Robin ◽  
Welding Residual Stress ◽  
Pressurized Water ◽  
Data Set ◽  
Strain Behavior ◽  
Sensitivity Studies

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under an addendum to the ongoing memorandum of understanding to validate welding residual stress (WRS) predictions in pressurized water reactor (PWR) primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in PWRs are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are the primary driver of this degradation mechanism. The NRC/EPRI weld residual stress (WRS) analysis validation program consists of four phases, with each phase increasing in complexity from laboratory size specimens to component mock-ups and ex-plant material. This paper focuses on Phase 2 of the WRS program that included an international Finite Element (FE) WRS round robin and experimental residuals stress measurements using the Deep Hole Drill (DHD) method on pressurizer surge nozzle mock-up. Characterizing variability in the round robin data set is difficult, as there is significant scatter in the data set and the WRS profile is dependent on the form of the material hardening law assumed. The results of this study show that, on average, analysts can develop WRS predictions that are a reasonable estimate for actual configurations as quantified by measurements. Sensitivity studies assist in determining which input parameters provide significant impact on WRSs, with thermal energy input, post-yield stress-strain behavior, and treatment of strain hardening have the greatest impact on DM WRS distributions.

Prediction of Weld Residual Stress in a Pressurized Water Reactor Pressurizer Surge Nozzle

2015 ◽  
Vol 138 (2) ◽  
Author(s):  
Akira Maekawa ◽  
Atsushi Kawahara ◽  
Hisashi Serizawa ◽  
Hidekazu Murakawa
Keyword(s):  
Residual Stress ◽  
Heat Source ◽  
High Speed ◽  
Nuclear Regulatory Commission ◽  
Round Robin ◽  
Welding Residual Stress ◽  
3D Analysis ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

Primary water stress corrosion cracking (PWSCC) phenomenon in dissimilar metal welds is one of the safety issues in ageing pressurized water reactor (PWR) piping systems. It is well known that analysis accuracy of cracking propagation due to PWSCC depends on welding residual stress conditions. The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) carried out an international round robin validation program to evaluate and quantify welding residual stress analysis accuracy and uncertainty. In this paper, participation results of the authors in the round robin program were reported. The three-dimensional (3D) analysis based on a fast weld simulation using an iterative substructure method (ISM), was shown to provide accurate results in a high-speed computation. Furthermore, the influence of different heat source models on analysis results was investigated. It was demonstrated that the residual stress and distortion calculated using the moving heat source model were more accurate.

Prediction of Weld Residual Stress in a PWR Pressurizer Surge Nozzle: A Proposed Fast Computational 3D Analysis Method and Influence of Its Heat Source Model

2012 ◽  
Author(s):  
Akira Maekawa ◽  
Atsushi Kawahara ◽  
Hisashi Serizawa ◽  
Hidekazu Murakawa
Keyword(s):  
Residual Stress ◽  
Heat Source ◽  
High Speed ◽  
Nuclear Regulatory Commission ◽  
Source Model ◽  
Round Robin ◽  
Welding Residual Stress ◽  
3D Analysis ◽  
Heat Source Model ◽  
Pressurized Water

Primary water stress corrosion cracking (PWSCC) generated in dissimilar metal welds is one of the safety issues in ageing pressurized water reactor piping systems. It is well known that analysis accuracy of cracking propagation due to PWSCC depends on welding residual stress conditions. The U.S. Nuclear Regulatory Commission carried out an international round robin program for welding residual stress analysis validation to evaluate the accuracy and uncertainty quantitatively. In this study, participation results in the round robin program were reported. The three-dimensional analysis based on a fast weld simulation using the Iterative Substructure Method was clarified to provide accurate results in a high-speed computation. Furthermore, the influence of different heat source models on analysis results was investigated. It was demonstrated that the residual stress and distortion calculated using the moving heat source model were more accurate.

Comparison of Welding Residual Stress Solutions for Control Rod Drive Mechanism Nozzles

2007 ◽  
Author(s):  
D. Rudland ◽  
Y. Chen ◽  
T. Zhang ◽  
G. Wilkowski ◽  
J. Broussard ◽  
...  
Keyword(s):  
Residual Stress ◽  
Yield Strength ◽  
Residual Stresses ◽  
Nuclear Regulatory Commission ◽  
Welding Residual Stress ◽  
Pressurized Water ◽  
Control Rod ◽  
Drive Mechanism ◽  
Time Period ◽  
Regulatory Commission

In the last 7 years, the incidences of cracking in Alloy 600 control rod drive mechanism (CRDM) tubes and their associated welds have increased significantly. The cracking mechanism has been attributed to pressurized water stress corrosion cracking (PWSCC) and has been shown to be driven by welding residual stresses and operational stresses in the weld region. During this time period, both the industry and the US Nuclear Regulatory Commission have been conducting detailed welding simulation analyses to predict the magnitude of these stresses in both the weld and tube material. To this point, a direct comparison of these analysis methodologies and results has not been made. In this paper, weld residual stress results from U.S. industry (conducted by Dominion Engineering) and the U.S. NRC (conducted by Engineering Mechanics Corporation of Columbus) for a steep angle (53 degrees) CRDM nozzle are compared. This comparison was performed for different yield strength tube materials, however only the low yield strength results are presented in this paper. The comparison illustrates the effect of weld analyses assumptions and suggests that simplifications in the analyses, i.e., lumping weld passes or material property assumptions, may lead to high predicted weld residual stresses.

Phase 2b Weld Residual Stress Round Robin: Mockup Design and Comparisons of Measurement and Simulation Results

2015 ◽  
Author(s):  
Michael L. Benson ◽  
Minh N. Tran ◽  
Michael R. Hill
Keyword(s):  
Residual Stress ◽  
Nuclear Regulatory Commission ◽  
Round Robin ◽  
Contour Method ◽  
Hole Drilling ◽  
Problem Statement ◽  
Double Blind ◽  
Weld Residual Stress ◽  
Surge Line ◽  
Regulatory Commission

The U. S. Nuclear Regulatory Commission and the Electric Power Research Institute, cooperating under the auspices of a memorandum of understanding, conducted a double-blind round robin study for prediction of weld residual stress in a full-scale pressurizer surge line mockup. This work is the latest in a series of studies aimed at understanding and reducing uncertainty in the numerical prediction of weld residual stress. The round robin study involved both measurements and modeling. The measurements included deep hole drilling and contour method. Ten international participants submitted finite element modeling results to the study. This paper summarizes the mockup design, the modeling problem statement, and the measurement and modeling results.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

NRC/EPRI Residual Stress Validation Program Phase I: Experimental Specimen Modeling and Measurement

2011 ◽  
Author(s):  
J. Broussard ◽  
P. Crooker
Keyword(s):  
Residual Stress ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Phase 1 ◽  
Measurement Techniques ◽  
Corrosion Cracking ◽  
Welding Residual Stress ◽  
Hole Drilling ◽  
Pressurized Water

The US Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress predictions in pressurized water reactor primary cooling loop components containing dissimilar metal welds. These stresses are of interest as DM welds in pressurized water reactors are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are one of the primary drivers of this stress corrosion cracking mechanism. The NRC/EPRI weld residual stress (WRS) program currently consists of four phases, with each phase increasing in complexity from lab size specimens to component mock-ups and ex-plant material. This paper describes the Phase 1 program, which comprised an initial period of learning and research for both FEA methods and measurement techniques using simple welded specimens. The Phase 1 specimens include a number of plate and cylinder geometries, each designed to provide a controlled configuration for maximum repeatability of measurements and modeling. A spectrum of surface and through-wall residual stress measurement techniques have been explored using the Phase 1 specimens, including incremental hole drilling, ring-core, and x-ray diffraction for surface stresses and neutron diffraction, deep-hole drilling, and contour method for through-wall stresses. The measured residual stresses are compared to the predicted stress results from a number of researchers employing a variety of modeling techniques. Comparisons between the various measurement techniques and among the modeling results have allowed for greater insight into the impact of various parameters on predicted versus measured residual stress. This paper will also discuss the technical challenges and lessons learned as part of the DM weld materials residual stress measurements.

An Experimental Study on Head Loss of Prototypical Fibrous Debris Beds During Loss-of-Coolant Accident Conditions

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Amir Ali ◽  
Edward D. Blandford
Keyword(s):  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Head Loss ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Debris Accumulation ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (NRC) initiated a generic safety issue (GSI-191) assessing debris accumulation and resultant chemical effects on pressurized water reactor (PWR) sump performance. GSI-191 has been investigated using reduced-scale separate-effects testing and integral-effects testing facilities. These experiments focused on developing a procedure to generate prototypical debris beds that provide stable and reproducible conventional head loss (CHL). These beds also have the ability to filter out chemical precipitates resulting in chemical head loss. The newly developed procedure presented in this paper is used to generate debris beds with different particulate to fiber ratios (η). Results from this experimental investigation show that the prepared beds can provide reproducible CHL for different η in a single and multivertical loops facility within ±7% under the same flow conditions. The measured CHL values are consistent with the predicted values using the NUREG-6224 correlation. Also, the results showed that the prepared debris beds following the proposed procedure are capable of detecting standard aluminum and calcium precipitates, and the head loss increase (chemical head loss) was measured and reported in this paper.

Status of the United States Nuclear Regulatory Commission Pressurized Thermal Shock Rule Re-Evaluation Project

2002 ◽  
Author(s):  
Terry L. Dickson ◽  
Shah N. Malik ◽  
Mark T. Kirk ◽  
Deborah A. Jackson
Keyword(s):  
United States ◽  
Thermal Shock ◽  
Computational Models ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Computational Methodology ◽  
Regulatory Commission

The current federal regulations to ensure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models that were developed in the early to mid 1980s. Since that time, there have been advancements in relevant technologies associated with the physics of PTS events that impact RPV integrity assessment. Preliminary studies performed in 1999 suggested that application of the improved technology could reduce the conservatism in the current regulations while continuing to provide reasonable assurance of adequate protection to public health and safety. A relaxation of PTS regulations could have profound implications for plant license extension considerations. Based on the above, in 1999, the United States Nuclear Regulatory Commission (USNRC) initiated a comprehensive project, with the nuclear power industry as a participant, to re-evaluate the current PTS regulations within the framework established by modern probabilistic risk assessment (PRA) techniques. During the last three years, improved computational models have evolved through interactions between experts in the relevant disciplines of thermal hydraulics, PRA, human reliability analysis (HRA), materials embrittlement effects on fracture toughness (crack initiation and arrest), fracture mechanics methodology, and fabrication-induced flaw characterization. These experts were from the NRC staff, their contractors, and representatives from the nuclear industry. These improved models have now been implemented into the FAVOR (Fracture Analysis of Vessels: Oak Ridge) computer code, which is an applications tool for performing risk-informed structural integrity evaluations of embrittled RPVs subjected to transient thermal-hydraulic loading conditions. The baseline version of FAVOR (version 1.0) was released in October 2001. The updated risk-informed computational methodology in the FAVOR code is currently being applied to selected domestic commercial pressurized water reactors to evaluate the adequacy of the current regulations and to determine whether a technical basis can be established to support a relaxation of the current regulations. This paper provides a status report on the application of the updated computational methodology to a commercial pressurized water reactor (PWR) and discusses the results and interpretation of those results. It is anticipated that this re-evaluation effort will be completed in 2002.

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