A Study of the Key Parameters Affecting the Design of Optimized Weld Overlays

2011 ◽  
Author(s):  
Doug Killian ◽  
Samer Mahmoud ◽  
Heqin Xu ◽  
Silvester Noronha ◽  
Ashok Nana
Keyword(s):  
Water Stress ◽  
Stress Corrosion Cracking ◽  
Reactor Vessel ◽  
Inside Surface ◽  
Design Parameters ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Surface Stresses ◽  
Asme Code ◽  
Primary Water

The potential for primary water stress corrosion cracking (PWSCC) of large diameter austenitic nickel alloy components and their associated welds presents a particular problem for the nuclear industry due to a limited number of available options for mitigating or repairing large bore pressure boundary components such as reactor vessel, reactor coolant pump, and steam generator inlet or outlet nozzles. While a full structural weld overlay (FSWOL), as governed by ASME Code Case N-740, is commonly used to mitigate and repair small (4″) to medium (10″) bore piping assemblies employing Alloy 82/182 dissimilar metal welds, the large amount of weld metal that would be have to be deposited on large components (and the associated impact on outage schedule) makes this an unattractive strategy for managing the degradation of Alloy 600 type materials. An alternative design option, specifically developed for the mitigation and repair of large bore (30″) components, utilizes a thinner weld overlay whose thickness has been optimized to achieve a specific level of stress on the inside surface of the PWSCC susceptible material. According to ASME Code Case N-754, inside surface stresses should be limited to 10 ksi during the design phase of an optimized weld overlay (OWOL) in order to minimize the initiation or consequences of primary water stress corrosion cracking. With the increased inspection requirements of Code Case N-754 and the corresponding smaller crack growth design flaw size, and along with the reduced weld volume of an OWOL, as compared to a FSWOL, an optimized weld overlay is often the preferred technique for mitigating or repairing large bore piping components. This paper investigates the influence of various parameters on the effectiveness of an optimized weld overlay in satisfying its principle design objective, to reduce the inside surface stresses in PWSCC susceptible materials to no more than 10 ksi. Inherent design parameters are the thickness of the underlying pipe or weld, and the depth of any recorded or postulated weld repairs in the pre-overlay configuration of the welded joint. Explicit design parameters include the thickness of the overlay, the number of weld layers used to form the overlay, and the length of the overlay. Finite element analysis is used to calculate residual and operating stresses in a representative large bore reactor vessel coolant nozzle dissimilar metal weld for various combinations of design parameters. The overall objective of this study is to identify the key parameters influencing inside surface stresses, and thereby provide screening criteria for use in determining the applicability of the optimized weld overlay as a viable PWSCC mitigation or repair option for large bore primary pressure boundary components.

Incorporating Peening Into ASME Section XI Code Cases N-729 and N-770 for PWSCC Mitigation in Alloy 82/182/600 Locations

2016 ◽  
Author(s):  
Dennis P. Weakland ◽  
Glenn White ◽  
Paul Crooker
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Surface Stress ◽  
Corrosion Cracking ◽  
Piping System ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Asme Code ◽  
Primary Water

This paper will discuss the ASME Code Committee activities involved in the incorporation of surface stress improvement (SSI) into ASME Code Cases N-770-4 and N-729-5. ASME Code Cases N-770 [1] and N-770-1 introduced several mitigation approaches for dissimilar metal weld (DMW) locations in PWR primary system piping and provided inspection relief for locations that were mitigated. The initial approaches contained in N-770 and N-770-1 included mechanical stress improvement and weld overlay methods that have a global stress relief effect to achieve a very low tensile surface stress state or a compressive stress state at the weld inside surface to halt crack initiation, as well as growth of acceptably sized cracks. The weld overlay mitigation methods are also effective because they introduce PWSCC-resistant material, i.e., Alloys 52, 152, or their variants. (The initial approaches also included Alloy 52/152 weld inlay and weld onlay, methods that do not require stress improvement but do require access to the weld inside surface.) While the mechanical stress improvement and weld overlay methods address the majority of the DMW locations in the primary piping system, there are locations that cannot be treated by these approaches due to the weld geometry or access limitations for the needed equipment. Additionally the dissimilar metal J-groove welds in the reactor pressure vessel head penetration nozzles (RPVHPN) could not be addressed at all by the approaches developed for DMW locations. To address the industry need to mitigate the unfavorable DMW geometries and locations along with the RPVHPN locations, the use of surface stress improvement (SSI) was studied and documented in EPRI reports Materials Reliability Program (MRP)-267 [2], “Technical Basis for Primary Water Stress Corrosion Cracking by Surface Stress Improvement,” and MRP-335 [3], “Topical Report for Primary Water Stress Corrosion Cracking by Surface Stress Improvement.” These reports formed the technical basis for the SSI-related changes made in Code Cases N-770-4 and N-729-5. Along with the technical bases noted, support from the international community in terms of operational experience with SSI in their power plants was invaluable in providing the necessary understanding, context, and confidence to committee members. The ASME “Task Group High Strength Nickel Alloy Issues” (TGHSNAI) was assigned the task of revising the existing Code Cases, N-770 [1], “Alternate Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated With UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities” and N-729 [4], “Alternate Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds.” To incorporate the SSI approach into these Code Cases, the first action was to determine whether the SSI process was considered to be a peening process as defined by ASME Section III NB-4422 criteria. This required the submittal of an Interpretation of NB-4422 to determine if SSI techniques were considered a peening process under ASME Section III. The interpretation (Interpretation III-1-13-03), documented in ASME File 12-1192 [5], specified that SSI was not considered peening by Section III. This interpretation provided the framework by which SSI could be directly applied to ASME Section XI inspection criteria without the need to first revise ASME Section III NB-4422. SSI (peening) was first incorporated into Code Case N-770 [1] to provide a mitigation alternative for locations unable to be addressed by the methods addressed thus far. The revision to Code Case N-770 [1] does not provide guidance for the application of SSI activities but rather, it provides the process performance criteria and the inspection guidance following the application of SSI and establishes the pre-application inspection acceptance criteria. Following the approval of SSI in Code Case N-770 [1] addressing the DMW in the primary coolant piping system, the SSI approach was applied to the partial penetration dissimilar metal J-groove welds in RPVHPNs in Code Case N-729 [4]. The application to RPVHPNs provides the industry with a valuable asset preservation tool while significantly lowering the safety risks associated with primary water stress corrosion cracking (PWSCC) and degradation from borated water leakage for the RPVHPNs.

Design of a Weld Overlay for a Large Bore Pipe Nozzle to Optimize Residual Stress

2010 ◽  
Author(s):  
Douglas E. Killian
Keyword(s):  
Residual Stress ◽  
Compressive Residual Stress ◽  
Reactor Vessel ◽  
Inside Surface ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Surface Flaws ◽  
Asme Code ◽  
Dissimilar Metal Weld ◽  
Metal Weld

Full Structural Weld Overlays (FSWOL) have been used successfully in the nuclear power industry for a number of years to mitigate and repair small (4″) to medium (10″) bore welded piping components susceptible to primary water stress corrosion cracking (PWSCC). Mitigation is provided by the creation of compressive residual stress on the inside surface of the pipe as layers of weld overlay are deposited over the outside surface of the pipe. ASME Code Case N-740-2 requires that these overlay designs provide adequate structural integrity considering the growth of postulated 75% through-wall inside surface flaws by PWSCC and cyclic fatigue. Application of this repair procedure to larger diameters components such as 30 inch reactor vessel nozzles is not practical due to the large amount of weld metal (overlay thickness) which would be required to satisfy the design requirements of a FSWOL and the associated demands on implementation schedule and exposure to radiation. An alternate procedure is currently being considered for these larger components which utilizes an Optimized Weld Overlay (OWOL) design based on a reduced thickness and smaller postulated flaw. In particular, ASME Code Case 754 specifies, in part, that 50% through-wall inside surface flaws be shown to be acceptable. Furthermore, an OWOL would continue to provide mitigation of materials susceptible to PWSCC by requiring that the thickness of the overlay be sufficient to induce compressive residual stress on the inside surface. This paper presents results of finite element analysis for an optimized weld overlay on a large bore (30″) reactor vessel coolant nozzle dissimilar metal weld, with particular attention to the incremental development of residual stress with each layer of weld metal. Through numerical simulation of the complete fabrication history, including repair of the original dissimilar metal weld, hydrostatic testing, and completion of the nozzle safe end-to-pipe joint prior to implementation of the overlay, the pre-overlay state of stress is defined for use as the basis for evaluating the stress improvement provisions of the weld overlay process. Results are obtained for both kinematic and isotropic hardening rules to study the effect of these two extreme measures of material characterization on the development of residual stress. Additional results are presented to study the sensitivity of the welding simulations to material yield strength and mesh refinement. Predicted stresses are also compared to measured data from a full scale mockup of a large bore reactor vessel nozzle with an optimized weld overlay.

Changing the Frequency of Inspections for PWSCC Susceptible Welds at Cold Leg Temperatures

2011 ◽  
Author(s):  
Nathan A. Palm ◽  
Warren H. Bamford ◽  
Craig Harrington
Keyword(s):  
Water Stress ◽  
Stress Corrosion Cracking ◽  
Large Diameter ◽  
Reactor Vessel ◽  
Detailed Model ◽  
Service Experience ◽  
Dissimilar Metal ◽  
Flaw Tolerance ◽  
Excellent Service ◽  
Primary Water

A project has been completed under the sponsorship of the EPRI Materials Reliability Program to evaluate the acceptability of returning to an inservice inspection (ISI) frequency of ten years for the large diameter cold leg pipes (525 to 580F), with Alloy 82/182 dissimilar metal (DM) welds. This effort addresses alternative inspection requirements with a frequency of 7 years that have recently been imposed in order to address the potential for service induced Primary Water Stress Corrosion Cracking (PWSCC) of these welds. Careful review of the service experience shows that cracking has only been observed in the hot leg piping locations with DM welds, and the cold leg locations continue to exhibit very reliable service. There are a number of technical and practical arguments in favor of making this change, even beyond the excellent service experience, and these arguments are summarized in this paper. • Pulling the reactor vessel (RV) core barrel is a serious activity which can entail many risks, so additional pulls should be avoided. Inspection at a frequency of less than 10 years involves additional core barrel pulls. • The flaw tolerance of these large diameter cold leg pipes is very good, and example calculations show that reasonably large flaws are acceptable for ten years. • The probability of cracks initiating in cold leg piping is significantly lower than that for piping at hotter temperatures, and a detailed model has been developed to demonstrate this. Actions are underway to revise the relevant inspection requirements, back to a more typical Section XI ten-year interval, using this technical work as a basis.

Mechanical Stress Improvement Process (MSIP) Used to Prevent and Mitigate Primary Water Stress Corrosion Cracking (PWSCC) in Reactor Vessel Piping at V.C. Summer

2003 ◽  
Author(s):  
E. A. Ray ◽  
K. Weir ◽  
C. Rice ◽  
T. Damico
Keyword(s):  
Water Stress ◽  
Stress Corrosion ◽  
Mechanical Stress ◽  
Stress Corrosion Cracking ◽  
Reactor Vessel ◽  
Corrosion Cracking ◽  
The Other ◽  
Water Reactor ◽  
Improvement Process ◽  
Primary Water

During the October 2000 refueling outage at the V.C. Summer Nuclear Station, a leak was discovered in one of the three reactor vessel hot leg nozzle to pipe weld connections. The root cause of this leak was determined to be extensive weld repairs causing high tensile stresses throughout the pipe weld; leading to primary water stress corrosion cracking (PWSCC) of the Alloy 82/182 (Inconel). This nozzle was repaired and V.C. Summer began investigating other mitigative or repair techniques on the other nozzles. During the next refueling outage V.C. Summer took mitigative actions by applying the patented Mechanical Stress Improvement Process (MSIP) to the other hot legs. MSIP contracts the pipe on one side of the weldment, placing the inner region of the weld into compression. This is an effective means to prevent and mitigate PWSCC. Analyses were performed to determine the redistribution of residual stresses, amount of strain in the region of application, reactor coolant piping loads and stresses, and effect on equipment supports. In May 2002, using a newly designed 34-inch clamp, MSIP was successfully applied to the two hot-leg nozzle weldments. The pre- and post-MSIP NDE results were highly favorable. MSIP has been used extensively on piping in boiling water reactor (BWR) plants to successfully prevent and mitigate SCC. This includes Reactor Vessel nozzle piping over 30-inch diameter with 2.3-inch wall thickness similar in both size and materials to piping in pressurized water reactor (PWR) plants such as V.C. Summer. The application of MSIP at V.C. Summer was successfully completed and showed the process to be predictable with no significant changes in the overall operation of the plant. The pre- and post-nondestructive examination of the reactor vessel nozzle weldment showed no detrimental effects on the weldment due to the MSIP.

Structural Integrity Evaluation of a Reactor Vessel Outlet Nozzle Weld Inlay

2012 ◽  
Author(s):  
N. L. Glunt ◽  
A. Udyawar ◽  
C. K. Ng ◽  
S. E. Marlette
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Repair Process ◽  
Reactor Vessel ◽  
Nuclear Industry ◽  
Outlet Nozzle ◽  
Weld Material ◽  
Asme Code ◽  
Primary Water

Nickel-base weldments such as Alloy 82/182 dissimilar metal (DM) butt welds used in Pressurized Water Reactor (PWR) nuclear power plant components have experienced Primary Water Stress Corrosion Cracking (PWSCC), resulting in the need to repair/replace these weldments. The nuclear industry has been actively engaged in inspecting and mitigating these susceptible DM butt welds for the past several years. Full and Optimized Structural Weld Overlay as well as Mechanical Stress Improvement Process (MSIP®) are some of the mitigation/repair processes that have been implemented successfully by the nuclear industry to mitigate PWSCC. Three conditions must exist simultaneously for PWSCC to occur: high tensile stresses, susceptible material and an environment that is conducive to stress corrosion cracking. These mitigation/repair processes are effective in minimizing the potential for future initiation and crack propagation resulting from PWSCC by generating compressive residual stress at the inner surface of the susceptible DM weld. Weld inlay is an alternative mitigation/repair process especially for large bore nozzles such as reactor vessel nozzles. The weld inlay process consists of excavating a small portion of the susceptible weld material at the inside surface of the component and then applying a PWSCC resistant Alloy 52/52M repair weld layer on the inside surface of the component to isolate the susceptible DM weld material from the primary water environment. The design and analysis requirements of the weld inlay are provided in ASME Code Case N-766. This paper provides the structural integrity evaluation results for a typical reactor vessel outlet nozzle weld inlay performed in accordance with the ASME Code Case N-766 design and analysis requirements. The evaluation results demonstrate that weld inlay is also a viable PWSCC mitigation and repair process especially for large bore reactor vessel nozzles.

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.

Manufacturing Methodology, Repair History, and Loadings for Westinghouse and Combustion Engineering Pressurizers, to Support Detailed Structural Integrity Evaluations of the Nozzle Safe End Configurations

2008 ◽  
Author(s):  
Cameron Martin ◽  
Warren Bamford ◽  
Nathan Palm
Keyword(s):  
Residual Stress ◽  
Water Stress ◽  
Stress Corrosion Cracking ◽  
Manufacturing Process ◽  
Structural Integrity ◽  
Welding Process ◽  
Dissimilar Metal ◽  
Primary Water ◽  
Compare And Contrast ◽  
Service Inspection

As a result of the finding of several large in-service inspection (ISI) indications at the Wolf Creek Nuclear Plant in the Fall of 2006, a major calculational effort was undertaken to accurately define the structural integrity margins which exist for the dissimilar metal welds, specifically the safety, relief, spray, and surge nozzle-to-safe-end regions. The Alloy 82/182 weld region of each of these nozzles is susceptible to primary water stress corrosion cracking (PWSCC), and predicting the extent of cracking accurately requires a detailed treatment of the residual stress present due to both the welding process used for the manufacture, and the extent of repairs present. This paper will discuss the manufacturing process used and compare and contrast them. A review will be provided of the types of repairs typically made to these components, as well as the types of loadings which typically occur during service. Also discussed will be the process used to ensure the accuracy of these findings in application to the operating plants of interest.

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.

scholarly journals Susceptibility to Severe PWSCC (primary water stress corrosion cracking) of LTMA (low temperature mill anneal) Alloy 600

2020 ◽  
Vol 58 (12) ◽  
pp. 815-821
Author(s):  
Sung Soo Kim ◽  
Jung Jong Yeob ◽  
Young Suk Kim
Keyword(s):  
Water Stress ◽  
Low Temperature ◽  
Stress Corrosion Cracking ◽  
Short Range ◽  
Alloy 600 ◽  
Lattice Contraction ◽  
Short Range Ordering ◽  
Mill Anneal ◽  
Primary Water ◽  
Kinetics Of

It has been proposed that a primary water stress corrosion cracking (PWSCC) in pressurized water reactor (PWR) is governed by a lattice contraction due to a short range ordering reaction in Alloy 600. This leads researcher to think that the kinetics of lattice contraction may control a susceptibility of PWSCC in Alloy 600. A lattice variation with ordering treatment at 400 <sup>o</sup>C was systematically investigated using high resolution neutron diffraction(HRPD) in high temperature mill anneal (HTMA), low temperature mill anneal (LTMA), and sensitized (SEN) Alloy 600. The results showed that ordering treatment caused an isotropic lattice contraction due to short range ordering (SRO) reaction. The lattice contractions of (111) plane are saturated to be 0.04% in 4 to 1500 hours at 400 <sup>o</sup>C according to prior treatment condition. The lattice contraction in the magnitude of 0.03% of (111) plane in LTMA Alloy 600 is faster by 8 times and 66 times than that of SEN and HTMA, respectively. This fact may explain why the LTMA is most susceptible to PWSCC through of kinetics of lattice contraction in Alloy 600. Thus, it is possible to conclude that the susceptibility of Alloy 600 to PWSCC is governed by the kinetics of (111) lattice contraction.

Sign in / Sign up

Export Citation Format

Share Document