A. Csontos

NRC Welding Residual Stress Validation Program International Round Robin Program and Findings

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are working cooperatively under a memorandum of understanding to validate welding residual stress (WRS) predictions in pressurized water reactor (PWR) primary cooling loop components containing dissimilar metal (DM) welds. These stresses are of interest as DM welds in PWRs are susceptible to primary water stress corrosion cracking (PWSCC) and tensile weld residual stresses are the primary driver of this degradation mechanism. The NRC/EPRI weld residual stress (WRS) analysis validation program consists of four phases, with each phase increasing in complexity from laboratory size specimens to component mock-ups and cancelled-plant material. This paper discusses Phase 2 of the WRS characterization program involving an international round robin analysis project in which participants analyzed a prototypic reactor coolant pressure boundary component. Mock-up fabrication, WRS measurements and comparison with predicted stresses through the DM weld region are described. The results of this study show that, on average, analysts can develop WRS predictions that are a reasonable estimate for actual configurations as quantified by measurements. However, the scatter in predicted results from analyst to analyst can be quite large. For example, in this study, the scatter in WRSs through the centerline of the main DM weld (prior to stainless steel weld application) predicted by analysts is approximately +/− 200 to 300 MPa at 3 standard deviations for axial stresses and +/− 300 to 400 MPa at 3 standard deviations for hoop stresses. Sensitivity studies that vary important parameters, such as material hardening behavior, can be used to bound such large variations.Copyright

Welding Residual Stress and Multiple Flaw Evaluation for Reactor Pressure Vessel Head Replacement Welds With Alloy 52

Small indications were found in one replacement reactor pressure vessel head (RPVH) mock-up being fabricated from Alloy 690 material and compatible weld metals, Alloy 52/152. The mockups were non-destructively examined and the lowest number of cracks found was five and the highest number was 22. There are numerous indications with some of them quite long (50 mm) in length. The source of these weld fabrication cracks is unknown. However, from experience with other difficult to weld materials, the source can range from slag inclusions in the weld metal to hot cracking during the weld deposition process. Hot cracking includes solidification cracking (weld), liquation cracking (HAZ), and ductility dip cracking (DDC). The indications were mostly circumferential in orientation (with respect to the nozzle axis) but some were axial. This paper includes two parts. The first part includes the welding residual stress analysis of RPVH using Alloy 52/152 metal and provides comparison with similar Alloy 82/182 welds. Alloy 82/182 was the material used in the original dissimilar metal welds in these heads. Primary Water Stress Corrosion Cracking (PWSCC) can occur in the primary coolant system when the welds are exposed to water, tensile stress, and temperature (usually higher than 250 C). PWSCC rates are higher in Alloy 82/182 material due to its lower chromium content compared with the replacement alloy. The results for both center hole (0-degree) and side hill (53-degree) nozzles will be discussed. The second part deals with assessment of multiple small cracks in the reactor pressure vessel head penetration nozzles. The finite element alternating method (FEAM) was used for calculating stress intensity factors for cases where multiple cracks exist. More than twenty cracks, which were inserted based on field measurements, are considered in the analyses for both center hole and side hill nozzles. It is observed that the overall stress trends are similar to those without adding cracks. However, cracks introduce more local stress fluctuations. The magnitude of the local fluctuation can be around 100MPa. Limit analysis was also conducted. A new finite element model with a voided-out weld region was used to simulate loss of structural capacity due to multiple flaws. The voided out volume effects on the structural integrity and future performance of RPVH were examined. Discussions based on weld residual stress, multiple flaw analysis and limit analysis conclude the paper.© 2009 ASME

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

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.

Welding Residual Stress and Flaw Evaluation for Dissimilar Metal Welds With Alloy 52 Inlays

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.Copyright

Development of a Probabilistic Pipe Rupture Assessment Code

Appendix A, General Design Criteria (GDC) 4, of 10 CFR Part 50 states, in part, that the dynamic effects associated with postulated reactor coolant system pipe ruptures may be excluded from the design basis when analyses reviewed and approved by NRC demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis. Licensees have typically demonstrated compliance with this probabilistic criterion through deterministic analyses that do not accurately simulate typical piping degradation. Given recent advances in probabilistic methodologies, the NRC staff and industry believe that performing a probabilistic analysis of primary system piping that fully addresses and quantifies uncertainties and directly demonstrates compliance with GDC 4 is more appropriate. NRC and industry expect that a robust probabilistic software tool, developed cooperatively, will facilitate meeting this goal; will improve licensing, regulatory decision making, and design; and will be mutually beneficial. Based on the terminology of GDC 4, this project is titled Extremely Low Probability of Rupture (xLPR). Development of the xLPR methodology and the corresponding software tool will involve many challenging technical decisions, modeling judgments, and sensitivity analyses. The purpose of the xLPR project is to develop a probabilistic assessment tool that can be used to demonstrate direct compliance with the requirements of 10 CFR 50 Appendix A GDC 4. The xLPR software tool will model active degradation mechanisms, such as primary water stress corrosion cracking, and the associated mitigation activities. The tool will be comprehensive with respect to known challenges, vetted with respect to scientific adequacy of models and inputs, flexible enough to permit analysis of a variety of in-service situations, and adaptable to accommodate evolving and improving knowledge. Successful execution of the xLPR project will involve a complex and diverse array of technical specialties and will require a well-organized and structured team of experts. This paper summarizes the objectives, organizational structure, and program plan of the collaborative NRC / Industry xLPR Project to develop a robust, flexible, probabilistic piping rupture assessment code. While the listed authors are individually responsible within their respective organizations for this project, the project organization includes a management team representing the xLPR Project task group structure. Individual members of this team will be listed in the paper along with their individual roles and contributions within the project. Companion papers from three of the four technical Task Groups of the xLPR Program (Computational, Modeling, and Inputs) provide substantial additional information detailing the technical approach and direction of this project.Copyright

Crack-Opening Displacement and Leak-Rate Calculations for Full Structural Weld Overlays

Full structural weld overlays have been used in the U.S. nuclear power industry for over twenty years in boiling water reactors (BWRs). Primary water stress corrosion cracking (PWSCC) in nickel-based dissimilar metal welds (DMWs) has been experienced in pressurized water reactors (PWRs) since the early 1990s. As a result, the nuclear industry is implementing full structural weld overlays (FSWOL) as a PWSCC mitigation technique that may be used on primary coolant lines previously approved for Leak-Before-Break (LBB). This work investigates the effect of the FSWOL on the leakage behavior of these lines with postulated defects. In this paper, finite element (FE) based crack-opening displacements (CODs) were developed for pipes with a FSWOL with postulated complex cracks. The COD solutions were then employed in standard leak-rate calculations, where equivalent crack morphology parameters were developed to consider a flow through two different crack morphologies, i.e., PWSCC through the DMW and corrosion fatigue through the weld overlay. The results of the sensitivity study and a discussion on the impact of the weld overlay on the leakage behavior concludes this paper.Copyright

An Analytical Evaluation of the Effect of Weld Sequencing on Residual Stresses Produced by Full Structural Weld Overlays on Pressurized Water Reactor Primary Cooling Piping

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 and describes the results of sensitivity studies which were performed to examine the effect of weld sequencing on the residual stresses produced in common configurations of PWR primary cooling system piping.

An Analytical Evaluation of the Effect of Weld Material Thickness on Residual Stresses Produced by Structural Weld Overlays on Pressurized Water Reactor Primary Cooling Piping

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 and describes the results of sensitivity studies which were performed to examine the effect of weld overlay thickness on the residual stresses for typical dissimilar metal weld configurations.

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

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.© 2008 ASME