Through-wall welding residual stress profiles for dissimilar metal nozzle butt welds in pressurized water reactors

2011 ◽  
Vol 34 (8) ◽  
pp. 624-641 ◽  
Author(s):  
TAE-KWANG SONG ◽  
JI-SOO KIM ◽  
CHANG-YOUNG OH ◽  
YUN-JAE KIM ◽  
CHI-YONG PARK ◽  
...  
Author(s):  
Tae-Kwang Song ◽  
Ji-Soo Kim ◽  
Chang-Young Oh ◽  
Hong-Yeol Bae ◽  
Jun-Young Jeon ◽  
...  

This paper provides the through-thickness welding residual stress profile in dissimilar metal nozzle butt welds of pressurized water reactors. For systematic investigations of the effects of geometric variables, i.e. the thickness and the radius of the nozzle and the length of the safe end, on welding residual stresses, idealized shape of nozzle is proposed and elastic-plastic thermo-mechanical finite element analyses are conducted. Through-wall welding residual stress profiles for dissimilar metal nozzle butt welds are proposed, which take a modified form of existing welding residual stress profiles developed for austenitic pipe butt weld in R6 code.


Author(s):  
D. Rudland ◽  
A. Csontos ◽  
F. Brust ◽  
T. Zhang

With the recent occurrences of primary water stress corrosion cracking (PWSCC) at nickel-based dissimilar metal welds (specifically Alloy 82/182 welds) in the nation’s pressurized water reactors (PWRs), the commercial nuclear power industry has been proposing a number of mitigation strategies for dealing with the problem. Some of these methods include Mechanical Stress Improvement Process (MSIP), Full and Optimized Structural Weld Overlay (FSWOL, OWOL) and Inlay and Onlay welds. All of these methods provide either a reduction in the ID residual stress field, (MSIP and WOL) and/or apply a corrosion resistant layer to stop or retard a leak path from forming (WOL, Inlay, Onlay). For the larger bore pipe, i.e. hot leg outlet nozzle, methods such as FSWOL become cost prohibitive due to the amount of weld metal that must be deposited. Therefore, inlay welds are being proposed since only a small layer (3 weld beads) needs to be deposited on the inside surface of the pipe. Currently the ASME code is developing Code Case N-766 ‘Nickel Alloy Reactor Coolant Inlay and Cladding for Repair or Mitigation of PWR Full Penetration Circumferential Nickel Alloy Welds in Class 1 Items.’ This code case is documenting the procedures for applying these inlay welds. As part of a confirmatory analysis, the US NRC staff and its contractor, Engineering Mechanics Corporation of Columbus, (Emc2) have conducted both welding residual stress and flaw evaluation analyses to determine the effectiveness of inlay welds as a mitigative technique. This paper presents the ongoing results from this effort. Using several large bore geometries, detailed welding simulation analyses were conducted on the procedures set forth in draft Code Case N-766. Effects of weld repairs and temper bead welding are included. Using these residual stress results, PWSCC growth analyses were conducted using simulated crack growth rates as a function of chromium content to estimate the time to leakage and rupture for small initial flaws in the inlay. The paper concludes with discussions on the effectiveness of inlays based on these analyses.


Author(s):  
Dean Deng ◽  
Kazuo Ogawa ◽  
Nobuyoshi Yanagida ◽  
Koichi Saito

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.


Author(s):  
G. White ◽  
J. Broussard ◽  
J. Collin ◽  
M. Klug ◽  
C. Harrington ◽  
...  

In late summer 2005, the U.S. pressurized water reactor (PWR) fleet imposed mandatory inspection requirements upon itself to address the challenge posed by primary water stress corrosion cracking (PWSCC) in PWR reactor coolant system (RCS) dissimilar metal (DM) piping butt welds. Under this program, the highest temperature, and thus most susceptible, locations have been addressed first. The set of highest temperature locations comprises the DM piping butt welds on the pressurizer. Within three years of promulgating the requirements, all pressurizer locations will have been inspected and nearly 90% of these locations will have been mitigated. In October 2006, several indications of circumferential flaws were reported in the pressurizer nozzles at Wolf Creek. These indications raised questions about the need to accelerate refueling outages or take mid-cycle outages at other plants. In order to address these concerns, an industry effort was undertaken to evaluate the viability of detection of leakage from a through-wall flaw in an operating plant to preclude the potential for rupture of pressurizer nozzle DM welds given the potential concern about growing circumferential stress corrosion cracks. Previous calculations of growth of PWSCC in Alloy 600 wrought materials and Alloy 82/182 weld metal materials have assumed an idealized crack shape, typically a semi-ellipse characterized by a length-to-depth aspect ratio. A key aspect of the industry effort involved developing an advanced finite-element analysis (FEA) methodology for predicting crack growth when loading conditions do not lead to a semi-elliptical flaw shape. The work also investigated an extensive crack growth sensitivity matrix to cover geometry, load, and fabrication factors, as well as the uncertainty in key modeling parameters including the effect of multiple flaw initiation sites in a single weld. Other key activities included detailed welding residual stress simulations covering the subject welds, development of a conservative crack stability calculation methodology, development of a leak rate calculation procedure using existing software tools (EPRI PICEP and NRC SQUIRT), and verification and validation studies. This paper will describe the study undertaken to model growth of circumferential weld cracks and its application to a group of nine PWRs with regard to implementation of the industry inspection and mitigation program [1]. The paper will also explore implementation progress of the industry program as the three-year mark approaches, as well as industry actions to support completion of baseline DM weld examinations.


Author(s):  
Michael R. Hill ◽  
Mitchell D. Olson ◽  
Adrian T. DeWald

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.


Author(s):  
Anne Teughels ◽  
Rodolfo L. M. Suanno ◽  
Christian Malekian ◽  
Lucio D. B. Ferrari

The penetrations in the early Pressurized Water Reactors Vessels are characterized by Alloy 600 tubes, welded by Alloy 182/82. The Alloy 600 tubes have been shown to be susceptible to PWSCC (Primary Water Stress Corrosion Cracking) which may lead to crack forming. The cracking mechanism is driven mainly by the welding residual stress and, in a second place, by the operational stress in the weld region. It is therefore of big interest to quantify the weld residual stress field correctly. In this paper the weld residual stress field is calculated by finite elements, using a common approach well known in nuclear domain. It includes a transient thermal analysis simulating the heating during the multipass welding, followed by a transient thermo-mechanical analysis for the determination of the stresses involved with it. The welding consists of a sequence of weld beads, each of which is deposited in its entirety, at once, instead of gradually. Central as well as eccentric sidehill nozzles on the vessel head are analyzed in the paper. For the former a 2-dimensional axisymmetrical finite element model is used, whereas for the latter a 3-dimensional model is set up. Different positions on the vessel head are compared and the influence of the sidehill effect is illustrated. In the framework of a common project for Angra 1, Tractebel Engineering (Belgium) and Eletronuclear (Nuclear Utility, Brazil) had the opportunity to compare their analysis method, which they applied to the Belgian and the Brazilian nuclear reactors, respectively. The global approach in both cases is very similar but is applied to different configurations, specific for each NPP. In the article the results of both cases are compared.


Author(s):  
Maan-Won Kim ◽  
Young-Jong Kim ◽  
Byoung Chul Kim

Primary water stress corrosion cracking (PWSCC) of Alloy 82/182 butt welds has been a concern for pressurized water reactor (PWR) plants worldwide for the past decade. A lot of works have been performed to calculate exact welding residual stresses and PWSCC growth rate for Alloy 82/182 butt welds. The PWSCC growth analysis of Alloy 82/182 butt welds has been performed by using the Raju-Newman type solutions for the crack tip stress intensity factors (SIFs). In this study, a finite element alternating method (FEAM) was used to calculate the SIFs and crack propagation in Alloy 82/182 butt weld. The FEAM is consisted of two solution parts: an exact theoretical SIF solution for a crack embedded in infinite body and a simple finite element analysis model with coarse mesh for finite body. In this study, the theoretical SIFs were derived by using a dislocation density function. Finite element (FE) analysis was performed to obtain the welding residual stress distribution for a nozzle with Alloy 82/182 butt weld and PWSCC growth analysis was performed by using the FEAM with PWSCC growth model described in MRP reports under welding residual stress along the nozzle thickness.


Author(s):  
Lee F. Fredette ◽  
Matthew Kerr ◽  
Howard J. Rathbun ◽  
John E. Broussard

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.


Author(s):  
D. Rudland ◽  
F. Brust ◽  
D. J. Shim ◽  
T. Zhang

Primary water stress corrosion cracking (PWSCC) in nickel-based dissimilar metal (DM) welds (specifically Alloy 82/182 welds) in pressurized water reactors (PWRs) can cause a safety concern due to the high crack growth rate and irregular shaped flaws. Since many of these welds reside in primary piping systems that have been approved for Leak-Before-Break (LBB), the domestic commercial nuclear power industry has proposed a number of mitigation strategies for dealing with the issue and assuring LBB is still applicable. Some of these methods include Mechanical Stress Improvement Process (MSIP), Full and Optimized Structural Weld Overlay (FSWOL, OWOL), and Inlay and Onlay cladding. The industry claims that these methods provide either a reduction in the inner diameter residual stress field (MSIP and WOL), and/or apply a non-susceptible corrosion resistant barrier to stop or retard PWSCC crack growth to form a through-wall leak path (WOL, Inlay, Onlay). At last years PVP conference, a companion paper was published that described the initial welding residual stress and flaw evaluation analyses to investigate the effectiveness of inlay welds as a mitigative technique. The results from that effort suggested that the time to leakage with an inlayed weld is highly affected by the depth of the inlay and the crack growth rate within the inlay. In this ongoing effort, further welding residual stress analyses are presented that investigate the effects of the inlay depth and a variety of weld repair options before the standard 3mm deep inlay. In addition, further crack growth analyses, assuming idealized crack shapes, were conducted to investigate the effects of weld residual stress, crack growth rate, global bending stress, and flaw size and orientation. The results of these analyses aid in determining appropriate inspection intervals for dissimilar metal welds with this mitigation technique.


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