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.

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.

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.

Large Break LOCA Safety Injection Sensitivity for a CE/ABB System 80+ PWR

2002 ◽  
Author(s):  
J. Pottorf ◽  
S. M. Bajorek
Keyword(s):  
Technical Specification ◽  
Nuclear Regulatory Commission ◽  
Injection System ◽  
Engineering System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Recovery ◽  
Regulatory Commission ◽  
Saturated Boiling

A WCOBRA/TRAC model of an evolutionary pressurized water reactor with direct vessel injection was constructed using publicly available information and a hypothetical double-ended guillotine break of a cold leg pipe was simulated. The model is an approximation of a ABB/Combustion Engineering System 80+ pressurized water reactor (PWR). WCOBRA/TRAC is the thermal-hydraulics code approved by the U.S. Nuclear Regulatory Commission for use in realistic large break LOCA analyses of Westinghouse 3- and 4-loop PWRs and the AP600 passive design. The AP600 design uses direct vessel injection, and the applicability of WCOBRA/TRAC to such designs is supported by comparisons to appropriate test data. This study extends the application of WCOBRA/TRAC to the investigation of the predicted behavior of direct vessel injection in an evolutionary design. A series of large break LOCA simulations were performed assuming a core power of 3914 MWt, and Technical Specification limits of 2.5 on total peaking factor and 1.7 on hot channel enthalpy rise factor. Two cladding temperature peaks were predicted during reflood, one following bottom of core recovery and a second caused by saturated boiling of water in the downcomer. This behavior is consistent with prior WCOBRA/TRAC calculations for some Westinghouse PWRs. The simulation results are described, and the sensitivity to failure assumptions for the safety injection system is presented.

Biaxial Residual Stress Mapping for a Dissimilar Metal Welded Nozzle

2014 ◽  
Author(s):  
Michael R. Hill ◽  
Mitchell D. Olson ◽  
Adrian T. DeWald
Keyword(s):  
Residual Stress ◽  
Hoop Stress ◽  
Axial Stress ◽  
Welding Residual Stress ◽  
Cylinder Wall ◽  
Two Dimensional ◽  
Pressurized Water Reactor ◽  
Steel Cylinder ◽  
Pressurized Water ◽  
Dissimilar Metal

This paper describes a sequence of residual stress measurements made to determine a two-dimensional map of biaxial residual stress in a nozzle mockup having two welds, one a dissimilar metal (DM) weld and the other a stainless steel (SS) weld. The mockup is cylindrical, designed to represent a pressurizer surge nozzle of a nuclear pressurized water reactor (PWR), and was fabricated for Phase 2a of the NRC/EPRI welding residual stress round robin. The mockup has a nickel alloy DM weld joining a SS safe end to a low-alloy steel cylinder and stiffening ring, as well as a SS weld joining the safe end to a section of pipe. The biaxial mapping experiments follow the approach described earlier, in PVP2012-78885 and PVP2013-97246, and comprise a series of experimental steps and a computation to determine a two-dimensional map of biaxial (axial and hoop) residual stress near the SS and DM welds. Specifically, the biaxial stresses are a combination of a contour measurement of hoop stress in the cylinder, slitting measurements of axial stress in thin slices removed from the cylinder wall, and a computation that determines the axial stress induced by measured hoop stress. At the DM weld, hoop stress is tensile near the OD (240 MPa) and compressive at the ID (−320 MPa), and axial stress is tensile near the OD (370 MPa) and compressive near the mid-thickness (−230 MPa) and ID (−250 MPa). At the SS weld, hoop stress is tensile near the OD (330 MPa) and compressive near the ID (−210 MPa), and axial stress is tensile at the OD (220 MPa) and compressive near mid-thickness (−225 MPa) and ID (−30 MPa). The measured stresses are found to be consistent with earlier work in similar configurations.

Fracture Mechanics Analyses of Embedded Cracks Under PTS and Effects of Residual Stresses

2017 ◽  
Author(s):  
Guian Qian ◽  
V. F. González-Albuixech ◽  
Markus Niffenegger
Keyword(s):  
Residual Stress ◽  
Fracture Mechanics ◽  
Residual Stresses ◽  
Welding Residual Stress ◽  
Pressurized Water Reactor ◽  
Extended Finite Element ◽  
Critical Crack ◽  
Pressurized Water ◽  
Pressurized Thermal Shock ◽  
Embedded Crack

One potential challenge to the integrity of a reactor pressure vessel (RPV) of a pressurized water reactor is posed by a pressurized thermal shock (PTS), which is associated with severe cooling of the RPV followed by its repressurization. PTS transients lead to high tensile circumferential and axial stresses in the RPV wall. If the stress intensity factor (SIF) is large enough, a critical crack may grow. Thus, the RPV has to be assessed against cleavage fracture. In this paper, two kinds of embedded cracks, i.e. semielliptical and elliptical crack with depth of 17 mm and length of 102 mm are considered. The extended finite element method (XFEM) is used to model such postulated cracks. The embedded crack with tip in the cladding/base interface causes a high KI. This is due to the stress discontinuities at the interface between the materials. In the FAVOR (probabilistic fracture mechanics code) calculation, for such cracks the closest point to the inner surface is calculated in order to be conservative. However, due to the highly ductile cladding material, it is unlikely for the embedded crack to propagate through the cladding. Thus, it is more appropriate to consider the outer surface point of the crack front. The effect of welding residual stress and cladding/base interface residual stress on the crack driving force is studied. Surface cracks are assumed in the study of residual stresses. Results show that considering realistic welding residual stresses may increase KI by about 5 MPa·m0.5, while the cladding/base interface residual stress has a negligible effect on KI. The reason is that the cladding residual stress is only localized to the interface and it decreases significantly through the vessel wall.

Risk-Informed Pressurized Water Reactor Vessel Inspection Interval Extension

2004 ◽  
Author(s):  
Cheryl L. Boggess ◽  
Bruce A. Bishop ◽  
Nathan A. Palm ◽  
Owen F. Hedden
Keyword(s):  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pilot Studies ◽  
Water Reactor ◽  
Insignificant Change ◽  
Inspection Interval ◽  
Interval Extension

The methodology discussed in this paper provides a risk informed basis for decreasing the frequency of inspection for the Pressurized Water Reactor (PWR) reactor pressure vessel (RPV). The decrease in frequency is based on extending the interval between inspections from the current interval of 10 years to 20 years. Results of pilot studies on typical designs of PWR vessels show that the change in risk associated with extending the inspection interval by more than 10 years is within the guidelines specified in U.S. Regulatory Guide 1.174 for insignificant change in risk. The current requirements for inspection of reactor vessel pressure-containing welds have been in effect since the 1989 Edition of American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, supplemented by U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.150, June 1981. The manner in which these examinations are conducted has recently been augmented by Appendix VIII of Section XI, 1996 Addenda, as implemented by the NRC in amendment to 10CFR50.55a effective November 22, 1999. This paper summarizes the insignificant change in risk results for the PWR pilot-plant studies, including the effects of fatigue crack growth and in-service inspection of postulated surface-breaking flaws. These results demonstrate that the proposed RPV inspection interval extension is a viable option for the industry.

Westinghouse AP1000 Solution to Long-Term Cooling Debris Concerns

2009 ◽  
Author(s):  
Terry L. Schulz ◽  
Timothy S. Andreychek ◽  
Yong J. Song ◽  
Kevin F. McNamee
Keyword(s):  
Nuclear Power ◽  
Safety Evaluation ◽  
Nuclear Regulatory Commission ◽  
Head Loss ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Geographical Regions ◽  
Core Cooling

The AP1000 is a pressurized water reactor with passive safety features and extensive plant simplifications that provides significant and measurable improvements in safety, construction, reliability, operation, maintenance and costs. The design of the AP1000 incorporates a standard approach, which results in a plant design that can be constructed in multiple geographical regions with varying regulatory standards and expectations. The AP1000 uses proven technology, which builds on more than 2,500 reactor years of highly successful Westinghouse PWR operation. The AP1000 received Final Design Approval by the Nuclear Regulatory Commission in September 2004. The AP1000 Nuclear Power Plant uses natural recirculation of coolant to cool the core following a postulated Loss Of Coolant Accident (LOCA). Recirculation screens are provided in strategic areas of the plant to remove debris that might migrate with the water in containment and adversely affect core cooling. The approach used to avoid the potential for debris to plug the AP1000 recirculation screens is consistent with the guidance identified in Regulatory Guide 1.82 Revision 3, the Pressurized Water Reactor (PWR) Industry Guidance of NEI 04–07, and the Nuclear Regulatory Commission’s Safety Evaluation on NEI 04–07. Various contributors to screen plugging were considered, including debris that could be produced by a LOCA, resident containment debris and post accident chemical products that might be generated in the coolant pool that forms on the containment floor post-accident. The solution developed for AP1000 includes three major aspects, including the elimination of debris sources by design, features that prevent transportation of debris to the screens and the use of large advanced screen designs. Measures were taken to design out debris sources including fibers, particles and chemicals. Available industry data from walkdowns in existing plants is used to determine the characteristics and amounts of the fibrous and particulate debris that could exist in containment prior to the LOCA. Materials used in the AP1000 containment are selected to eliminate post accident chemical debris generation. Large, advanced screen designs that can tolerate significant quantities of debris have been incorporated. Testing has been performed which demonstrates that the AP1000 screens will have essentially no head loss considering the debris that could be transported to them. Testing has also been performed on an AP1000 fuel assembly that demonstrates that it will also have essentially no head loss considering the debris that could be transported to it.

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