PWSCC Crack Growth Modeling Approaches

2011 ◽  
Author(s):  
F. W. Brust ◽  
D.-J. Shim ◽  
G. Wilkowski ◽  
D. Rudland
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Intensity ◽  
Residual Stresses ◽  
Crack Growth ◽  
Corrosion Cracking ◽  
Growth Time ◽  
Pressurized Water ◽  
Primary Water ◽  
Water Reactors

Flaw indications have been found in some dissimilar metal (DM) nozzle to stainless steel piping welds and reactor pressure vessel heads (RPVH) in pressurized water reactors (PWR) throughout the world. The nozzle welds usually involve welding ferritic (often A508) nozzles to 304/316 stainless steel pipe) using Alloy 182/82 weld metal. The welds may become susceptible to a form of corrosion cracking referred to as primary water stress corrosion cracking (PWSCC). It can occur if the temperature is high enough (usually >300C) and the water chemistry in the PWR is typical of operating plants. The weld residual stresses (WRS) induced by the welds are a main driver of PWSCC. Several mechanical mitigation methods to control PWSCC have been developed for use on a nozzle welds in nuclear PWR plants. These methods consist of applying a weld overlay repair (WOR), using a method called mechanical stress improvement process (MSIP), and applying an inlay to the nozzle ID. The purpose of a mitigation method is to reduce the probability that PWSCC will occur in the nozzle joint. The key to assessing the effectiveness of mitigation is to determine the crack growth time to leak with and without the mitigation. Indeed, for WOR and MSIP, the weld residual stresses are often reduced after application while for inlay they are actually increased. However, all approaches reduce crack growth rates if applied properly. Procedures for modeling PWSCC growth tend to vary between organizations performing the analyses. Currently, the prediction of PWSCC crack growth is based on the stress intensity factors at the crack tips. Several methods for evaluating the stress intensity factor for modeling the crack growth through these WRS fields are possible, including using analytical, natural crack growth using finite element methods, and using the finite element alternating method. This paper will summarize the methods used, critique the procedures, and provide some examples for crack growth with and without mitigation. Suggestions for modeling such growth will be provided.

Weld Residual Stress Analysis and Axial PWSCC Predictions in a Double Vee Groove Large Diameter Nozzle

2013 ◽  
Author(s):  
Frederick W. Brust ◽  
E. Punch ◽  
D. J. Shim ◽  
David Rudland ◽  
Howard Rathbun
Keyword(s):  
Residual Stress ◽  
Stainless Steel ◽  
Residual Stresses ◽  
Crack Growth ◽  
Large Diameter ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Weld Residual Stress ◽  
Primary Water ◽  
Water Reactors

Flaw indications have been found in some dissimilar metal (DM) nozzle to stainless steel piping welds and reactor pressure vessel heads (RPVH) in pressurized water reactors (PWR) throughout the world. The nozzle welds usually involve welding ferritic (often A508) nozzles to 304/316 stainless steel pipe) using Alloy 182/82 weld metal. The welds may become susceptible to a form of corrosion cracking referred to as primary water stress corrosion cracking (PWSCC). It can occur if the temperature is high enough (usually greater than 300°C) and the water chemistry in the PWR is typical of operating plants. The weld residual stresses (WRS) induced by the welds are a main driver of PWSCC. The purpose of this paper is to determine the weld residual stresses in a double-vee groove welded nozzle and then to model the natural crack growth in the weld. The double vee groove geometry has not been modeled much to date especially in such a large nozzle. This leads to a rather unique weld residual stress pattern which changes as the throat of the double vee is approached. Axial crack growth is modeled using a natural crack growth procedure. This was challenging since the crack shape necked down in the region where the tips of the vee grooves meet making the mesh development during this process challenging. This analysis provides information regarding crack growth evolution versus time. In addition, some comments regarding idealized growth are presented.

Modeling Crack Growth in Weld Residual Stress Fields Using the Finite Element Alternating Method

2011 ◽  
Author(s):  
F. W. Brust ◽  
T. Zhang ◽  
D.-J. Shim ◽  
G. Wilkowski ◽  
D. Rudland
Keyword(s):  
Finite Element ◽  
Stainless Steel ◽  
Stress Intensity ◽  
Crack Growth ◽  
Stress Intensity Factors ◽  
Corrosion Cracking ◽  
Multiple Cracks ◽  
Intensity Factors ◽  
Pressurized Water ◽  
Alternating Method

Flaw indications have been found in some nozzle to stainless steel piping dissimilar metal (DM) welds and reactor pressure vessel heads (RPVH) in pressurized water reactors (PWR) throughout the world. The nozzle welds usually involve welding ferritic (often A508) nozzles to 304/316 stainless steel pipe) using Alloy 182/82 weld metal. The welds may become susceptible to a form of corrosion cracking referred to as primary water stress corrosion cracking (PWSCC). It can occur if the temperature is high enough (usually >300C) and the water chemistry in the PWR is typical of operating plants. The weld residual stresses (WRS) induced by the welds are a main driver of PWSCC. Modeling the growth of these cracks in these WRS fields until leakage occurs is important for safety assessments. Currently, the prediction of PWSCC crack growth is based on the stress intensity factors at the crack tips. Several methods for modeling the crack growth through these WRS fields are possible, including using analytical, natural crack growth using finite element methods, and using the finite element alternating method. In this paper, finite element alternating method (FEAM) is used for calculating stress intensity factors and modeling the growth. First the FEAM method for growing cracks is presented. Next, several examples of modeling growth through control rod drive mechanism (CRDM) heads are presented. Finally, a short example examining multiple cracks in CRDM heads is presented. For many problems the FEAM approach for rapidly modeling crack growth is quite convenient, especially for difficult to mesh crack geometries.

Stress Corrosion Cracking of Nickel Weld Metals in PWR Primary Water

2013 ◽  
Author(s):  
Ru Xiong ◽  
Yuxiang Zhao ◽  
Guiliang Liu
Keyword(s):  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Cold Work ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Weld Metals ◽  
Surface Residual Stresses ◽  
Primary Water ◽  
Water Reactors

This discussion reviews the occurrence of stress corrosion cracking (SCC) of Alloys 182 and 82 weld metals in primary water of pressurized water reactors (PWR) from both operating plants and laboratory experiments. Results from in-service experience show that more than 340 Alloy 182/82 welds have sustained SCC, and Alloy 182 with lower Cr have more failures than Alloy 82. Most of these cases have been attributed to the presence of high residual stresses produced during the manufacture aside from the inherent tendency for Alloy 182/82 to sustain SCC. The affected welds were not subjected to a stress relief heat treatment with adjacent low alloy steel components. Results from laboratory studies indicate that time-to-cracking of Alloy 82 (with Cr up to 18%–22%) was a factor of 4 to 10 longer than that for Alloy 182. SCC depends strongly on surface conditions, surface residual stresses and surface cold work, which are consistent with the results of in-service failures. Improvements in the resistance of advanced weld metals, Alloys 152 and 52, to SCC are discussed.

Analysis of a Weld Overlay to Address Fatigue Cracking in a Stainless Steel Nozzle

2018 ◽  
Author(s):  
S. E. Marlette ◽  
A. Udyawar ◽  
J. Broussard
Keyword(s):  
Stainless Steel ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Fatigue Cracking ◽  
Corrosion Cracking ◽  
Nuclear Industry ◽  
Pressurized Water ◽  
Design Life

For several decades the nuclear industry has used structural weld overlays (SWOL) to repair and mitigate cracking within pressurized water reactor (PWR) components such as nozzles, pipes and elbows. There are two known primary mechanisms that have led to cracking within PWR components. One source of cracking has been primary water stress corrosion cracking (PWSCC). Numerous SWOL repairs and mitigations were installed in the early 2000s to address PWSCC in components such as pressurizer nozzles. However, nearly all of the likely candidate components for SWOL repairs have now been addressed in the industry. The other cause for cracking has been by fatigue, which usually results from thermal cycling events such as leakage caused by a faulty valve close to the component. The PWR components of most concern for fatigue cracking are mainly stainless steel. Thus, ASME Section XI Code Case N-504-4 would be a likely basis for SWOL repairs of these components, although this Code Case was originally drafted to address stress corrosion cracking (SCC) in boiling water reactors (BWR). N-504-4 includes the requirements for the SWOL design and subsequent analyses to establish the design life for the overlay based on predicted crack growth after the repair. This paper presents analysis work performed using Code Case N-504-4 to establish the design life of a SWOL repair applied to a boron injection tank (BIT) line nozzle attached to the cold leg of an operating PWR. The overlay was applied to the nozzle to address flaws found within the stainless steel base metal during inservice examination. Analyses were performed to calculate the residual stresses resulting from the original fabrication and the subsequent SWOL repair. In addition, post-SWOL operating stresses were calculated to demonstrate that the overlay does not invalidate the ASME Section III design basis for the nozzle and attached pipe. The operating and residual stresses were also used for input to a fatigue crack growth (FCG) analysis in order to establish the design life of the overlay. Lastly, the weld shrinkage from the application of overlay was evaluated for potential impact on the attached piping, restraints and valves within the BIT line. The combined analyses of the installed SWOL provide a basis for continued operation for the remaining life of the plant.

Stress Corrosion Cracking (SCC) of Nickel-Base Weld Metals in PWR Primary Water

2013 ◽  
Vol 747-748 ◽  
pp. 723-732 ◽  
Author(s):  
Ru Xiong ◽  
Ying Jie Qiao ◽  
Gui Liang Liu
Keyword(s):  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Surface Condition ◽  
Cold Work ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Weld Metals ◽  
Surface Residual Stresses ◽  
Primary Water

This discussion reviewed the occurrence of stress corrosion cracking (SCC) of alloys 182 and 82 weld metals in primary water (PWSCC) of pressurized water reactors (PWR) from both operating plants and laboratory experiments. Results from in-service experience showed that more than 340 Alloy 182/82 welds have sustained PWSCC. Most of these cases have been attributed to the presence of high residual stresses produced during the manufacture aside from the inherent tendency for Alloy 182/82 to sustain SCC. The affected welds were not subjected to a stress relief heat treatment with adjacent low alloy steel components. Results from laboratory studies indicated that time-to-cracking of Alloy 82 was a factor of 4 to 10 longer than that for Alloy 182. PWSCC depended strongly on the surface condition, surface residual stresses and surface cold work, which were consistent with the results of in-service failures. Improvements in the resistance of advanced weld metals, Alloys 152 and 52, to PWSCC were discussed.

Weld Residual Stresses and Primary Water Stress Corrosion Cracking in Bimetal Nuclear Pipe Welds

2007 ◽  
Author(s):  
Frederick W. Brust ◽  
Paul M. Scott
Keyword(s):  
Water Stress ◽  
Residual Stresses ◽  
Weld Metal ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Pressure Vessel ◽  
Large Diameter ◽  
Corrosion Cracking ◽  
Primary Water

There have been incidents recently where cracking has been observed in the bi-metallic welds that join the hot leg to the reactor pressure vessel nozzle. The hot leg pipes are typically large diameter, thick wall pipes. Typically, an inconel weld metal is used to join the ferritic pressure vessel steel to the stainless steel pipe. The cracking, mainly confined to the inconel weld metal, is caused by corrosion mechanisms. Tensile weld residual stresses, in addition to service loads, contribute to PWSCC (Primary Water Stress Corrosion Cracking) crack growth. In addition to the large diameter hot leg pipe, cracking in other piping components of different sizes has been observed. For instance, surge lines and spray line cracking has been observed that has been attributed to this degradation mechanism. Here we present some models which are used to predict the PWSCC behavior in nuclear piping. This includes weld model solutions of bimetal pipe welds along with an example calculation of PWSCC crack growth in a hot leg. Risk based considerations are also discussed.

Continuity of Environmentally Assisted Fatigue and Stress Corrosion Cracking Based on Short Crack Growth Behavior of 316 Stainless Steel in Simulated PWR Primary Water

2017 ◽  
Author(s):  
Choongmoo Shim ◽  
Yoichi Takeda ◽  
Tetsuo Shoji
Keyword(s):  
Stainless Steel ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Water Environment ◽  
Growth Behavior ◽  
Corrosion Cracking ◽  
Short Crack ◽  
Crack Growth Behavior ◽  
Primary Water

Environmental correction factor (Fen) is one of the parameters to evaluate the effect of a pressurized high temperature water environment. It has been reported that Fen for stainless steel saturates at a very low strain rate. However, the relationship between environmentally assisted fatigue (EAF) and stress corrosion cracking (SCC) is still unclear. The aim of this study is to investigate the short crack growth behavior and possible continuity of EAF and SCC at very low strain rates. Short crack initiation and propagation have similar behaviors, which retard the crack growth between 100–200 μm in depth. We find that the striation spacing correlates well with the maximum crack growth rate (CGR) data. Based on the correlation, it is clarified that the local CGR on an intergranular facet was faster than that on a transgranular facet. Furthermore, the overall maximum and average CGR from the EAF data is well interpreted and compared with the SCC data.

Regulatory Perspective on Advanced Finite Element Flaw Growth Analysis of Alloy 82/182 Butt Welds

2008 ◽  
Author(s):  
Edmund J. Sullivan ◽  
Aladar A. Csontos ◽  
Timothy R. Lupold ◽  
Chia-Fu Sheng
Keyword(s):  
Finite Element ◽  
Nuclear Power ◽  
Regulatory Review ◽  
Pressurized Water ◽  
Butt Weld ◽  
Butt Welds ◽  
Primary Water ◽  
Water Reactors ◽  
The Impact ◽  
Growth Analyses

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed preweld overlay inspections using manual ultrasonic testing (UT) techniques on the surge, spray, relief, and safety nozzle-to-safe end dissimilar metal (DM) and safe end-to-pipe stainless steel butt welds. The inspection identified five circumferential indications in the surge, relief, and safety nozzle-to-safe end DM butt welds that the licensee attributed to primary water stress corrosion cracking (PWSCC). These indications were significantly larger and more extensive than previously seen for the case of circumferential indications in commercial pressurized water reactors. As a result of the NRC staff’s initial flaw growth analyses, the NRC staff obtained commitments from the nuclear power industry licensees to complete pressurizer nozzle DM butt weld inspections on an accelerated basis. In addition, the industry informed the NRC staff that it would undertake a task to refine the crack growth analyses using more realistic assumptions to address the NRC staff’s concerns regarding the potential for rupture without prior evidence of leakage from circumferentially oriented PWSCC in pressurizer nozzle welds. These new analyses are referred to as advanced finite element (AFE) analyses. This paper will discuss the regulatory review of the industry’s AFE analyses. This discussion will include the NRC staff’s approach to the review, the differences between the industry’s AFE analyses and the NRC staff’s confirmatory analyses, and the NRC staff’s acceptance criteria. The paper will discuss the impact of the AFE analyses on the regulatory process. Finally, the paper will discuss possible future regulatory and research applications for AFE analyses as well as additional NRC research projects intended to address some of the uncertainties in this type of analysis.

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