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.

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.

An Analytical Evaluation of the Full Structural Weld Overlay as a Stress Improving Mitigation Strategy to Prevent Primary Water Stress Corrosion Cracking in Pressurized Water Reactor Piping

2009 ◽  
Author(s):  
L. F. Fredette ◽  
Paul M. Scott ◽  
F. W. Brust ◽  
A. Csontos
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Before And After ◽  
Primary Water ◽  
Water Reactors ◽  
Metal Weld

Full Structural Weld Overlay (FSWOL) has been used successfully to mitigate intergranular stress corrosion cracking in boiling water reactor (BWR) welded stainless steel piping for many years. The FSWOL technique adds structural reinforcement, can add crack resistant material, and can create compressive residual stresses at the inside surface of the welded joint which reduces the possibility of further stress corrosion cracking. Recently, the FSWOL has been applied as a preemptive measure to prevent primary water stress corrosion cracking (PWSCC) in pressurized water reactors (PWR) on susceptible welded pipes with dissimilar metal welds common to PWR primary cooling piping. This study uses finite element models to evaluate the likely residual and operating stress profiles remaining after FSWOL for typical dissimilar metal weld configurations, some of which are approved for leak-before-break (LBB) applications in pressurized water reactors. Circumferential cracks were modeled in the dissimilar metal weld area and forced to grow in order to evaluate their crack opening displacements and stress intensity factors vs. depth before and after weld overlay and before and after application of operating pressure and temperature.

The Stress Corrosion Cracking Behavior of Alloys 690 and 152 Weld in a PWR Environment

2008 ◽  
Author(s):  
B. Alexandreanu ◽  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Water Environment ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Alloy 690 ◽  
Primary Water ◽  
Cold Rolled ◽  
Water Reactors

Alloys 690 and 152 are the replacement materials of choice for Alloys 600 and 182, respectively. The latter two alloys are used as structural materials in pressurized water reactors (PWRs) and have been found to undergo stress corrosion cracking (SCC). The objective of this work is to determine the crack growth rates (CGRs) in a simulated PWR water environment for the replacement alloys. The study involved Alloy 690 cold-rolled by 26% and a laboratory-prepared Alloy 152 double-J weld in the as-welded condition. The experimental approach involved pre-cracking in a primary water environment and monitoring the cyclic CGRs to determine the optimum conditions for transitioning from the fatigue transgranular to intergranular SCC fracture mode. The cyclic CGRs of cold-rolled Alloy 690 showed significant environmental enhancement, while those for Alloy 152 were minimal. Both materials exhibited SCC of 10−11 m/s under constant loading at moderate stress intensity factors. The paper also presents tensile property data for Alloy 690TT and Alloy 152 weld in the temperature range 25–870°C.

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.

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.

Effects of Weld Overlays on Leak-Before-Break Margins

2011 ◽  
Author(s):  
G. Angah Miessi ◽  
Peter C. Riccardella ◽  
Peihua Jing
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Operating Conditions ◽  
Intergranular Stress Corrosion ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Water Reactors ◽  
Weld Overlays ◽  
Leak Before Break

Weld overlays have been used to remedy intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWRs) since the 1980s. Overlays have also been applied in the last few years in pressurized water reactors (PWRs) where primary water stress corrosion cracking (PWSCC) has developed. The weld overlay provides a structural reinforcement with SCC resistant material and favorable residual stresses at the ID of the overlaid component. Leak-before-break (LBB) had been applied to several piping systems in PWRs prior to recognizing the PWSCC susceptibility of Alloy 82/182 welds. The application of the weld overlay changes the geometric configuration of the component and as such, the original LBB evaluation is updated to reflect the new configuration at the susceptible weld. This paper describes a generic leak-before-break (LBB) analysis program which demonstrates that the application of weld overlays always improves LBB margins, relative to un-overlaid, PWSCC susceptible welds when all the other parameters or variables of the analyses (loads, geometry, operating conditions, analysis method, etc…) are kept equal. Analyses are performed using LBB methodology previously approved by the US NRC for weld overlaid components. The analyses are performed for a range of nozzle sizes (from 6″ to 34″) spanning the nominal pipe sizes to which LBB has been commonly applied, using associated representative loads and operating conditions. The analyses are performed for both overlaid and un-overlaid configurations of the same nozzles, and using both fatigue and PWSCC crack morphologies in the leakage rate calculations and the LBB margins are compared to show the benefit of the weld overlays.

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