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.

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.

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.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

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.

NRC/EPRI Welding Residual Stress Validation Program: Phase III Details and Findings

2011 ◽  
Author(s):  
Lee F. Fredette ◽  
Matthew Kerr ◽  
Howard J. Rathbun ◽  
John E. Broussard
Keyword(s):  
Residual Stress ◽  
Stainless Steel ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Plant Material ◽  
Welding Residual Stress ◽  
Phase Iii ◽  
Pressurized Water ◽  
Dissimilar Metal

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 (DM) 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 welding 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 discusses Phase III of the WRS characterization program, comparing measured and predicted weld residual stresses profiles through the dissimilar metal weld region of pressurizer safety and relief nozzles removed from a cancelled plant in the United States. The DM weld had already been completed on all of the plant nozzles before use in the mock-up program. One of the nozzles was completed with the application of the stainless steel safe-end weld to a section of stainless steel pipe. Measurements were taken on the nozzles with and without the welded pipe section. Several independent finite element analysis predictions were made of the stress state in the DM weld. This paper compares the predicted stresses to those found by through-thickness measurement techniques (Deep Hole Drilling and Contour Method). Comparisons of analysis results with experimental data will allow the NRC staff to develop unbiased measures of uncertainties in weld residual stress predictions with the goal of developing assurances that the analysis predictions are defensible through the blind validation provided using well controlled mock-ups and ex-plant material in this program.

FEA Welding Residual Stress and Fracture Mechanics Model for Nozzles With J-Groove Attachment Welds: Methodology and Application

2004 ◽  
Author(s):  
John E. Broussard ◽  
E. Stephen Hunt ◽  
Glenn A. White
Keyword(s):  
Residual Stress ◽  
Finite Element ◽  
Fracture Mechanics ◽  
Residual Stresses ◽  
Model Development ◽  
Welding Residual Stress ◽  
Element Analysis ◽  
Mechanics Model ◽  
Modeling Techniques ◽  
Reactor Pressure

Residual stresses due to welding in reactor pressure vessel (RPV) top head nozzle penetrations have been predicted using finite element analysis since the early 1990s. While the analyses were originally targeted at calculating nozzle stresses, the finite element methods have been extended to model a number of different aspects of RPV head penetrations. Both top and bottom head penetrations have been modeled, and the effects of J-groove butter weld deposition and subsequent thermal stress relief of the top head are now included in the analytical model. Development work has recently been completed to integrate a fracture mechanics model into the welding residual stress model. This has allowed for the prediction of crack tip stress intensity factors in the presence of welding residual stresses that include the effects of stress redistribution due to the presence of the crack. This paper presents some of the modeling techniques used in these recent analyses, and some key results obtained.

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.

Finite Element Analyses of Residual Stresses in Typical Welded Joints Used in Nuclear Power Plants and Comparisons With Experiments

2010 ◽  
Author(s):  
Dean Deng ◽  
Kazuo Ogawa ◽  
Nobuyoshi Yanagida ◽  
Koichi Saito
Keyword(s):  
Numerical Simulation ◽  
Residual Stress ◽  
Residual Stresses ◽  
Welded Joints ◽  
Nuclear Power ◽  
Power Plants ◽  
Butt Joint ◽  
Welding Residual Stress ◽  
Dissimilar Metal ◽  
Water Reactors

Recent discoveries of stress corrosion cracking (SCC) at nickel-based metals in pressurized water reactors (PWRs) and boiling water reactors (BWRs) have raised concerns about safety and integrity of plant components. It has been recognized that welding residual stress is an important factor causing the issue of SCC in a weldment. In this study, both numerical simulation technology and experimental method were employed to investigate the characteristics of welding residual stress distribution in several typical welded joints, which are used in nuclear power plants. These joints include a thick plate butt-welded Alloy 600 joint, a dissimilar metal J-groove set-in joint and a dissimilar metal girth-butt joint. First of all, numerical simulation technology was used to predict welding residual stresses in these three joints, and the influence of heat source model on welding residual stress was examined. Meanwhile, the influence of other thermal processes such as cladding, buttering and heat treatment on the final residual stresses in the dissimilar metal girth-butt joint was also clarified. Secondly, we also measured the residual stresses in three corresponding mock-ups. Finally, the comparisons of the simulation results and the measured data have shed light on how to effectively simulate welding residual stress in these typical joints.

Influence of Weld Residual Stresses on Ductile Crack Behavior in AISI Type 316LN Stainless Steel Weld Joint

2018 ◽  
Author(s):  
Sai Deepak Namburu ◽  
Lakshmana Rao Chebolu ◽  
A. Krishnan Subramanian ◽  
Raghu Prakash ◽  
Sasikala Gomathy
Keyword(s):  
Residual Stress ◽  
Residual Stresses ◽  
Crack Growth ◽  
Weld Joint ◽  
Volume Fraction ◽  
Growth Behavior ◽  
Welding Residual Stress ◽  
Crack Growth Behavior ◽  
Ductile Crack ◽  
Ductile Crack Growth

Welding residual stress is one of the main concerns in the process of fabrication and operation because of failures in welded steel joints due to its potential effect on structural integrity. This work focuses on the effect of welding residual stress on the ductile crack growth behavior in AISI 316LN welded CT specimens. Two-dimensional plane strain model has been used to simulate the CT specimen. X-ray diffraction technique is used to obtain residual stress value at the SS 316LN weld joint. The GTN model has been employed to estimate the ductile crack growth behavior in the CT-specimen. Results show that residual stresses influence the ductile crack growth behavior. The effect of residual stress has also been investigated for cases with different initial void volume fraction, crack lengths.

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