D. Rudland

Effects of crack morphology parameters on leak-rate calculations in LBB evaluations

Abstract LBB analyses typically involve a two-step process. The first step is the determination of the through-wall crack length that leaks at the minimum detectable rate at normal operating conditions. In the second step, a safety factor is applied to this leakage detectable flaw length (i.e. the flaw length is increased by a factor) and then the stability of the resulting flaw at transient loads (i.e. seismic) is calculated. Probabilistic LBB analyses have shown that the leakage flaw size is more important to the failure probability than the fracture mechanics analysis at the seismic loads. The crack morphology is an important parameter in the calculation of the leakage flaw size. This paper contrasts results from using various crack morphology assumptions in leak-rate calculations. The results show that using a statistical analysis of surface roughness and numbers of turns from cracks removed from service can result in leakage crack sizes that are greatly different for different cracking mechanisms. Using improper morphology parameters can lead to large errors in leakage crack size and can produce non-conservative margins on the LBB critical crack size.

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 Solutions for Dissimilar Metal Surge Line Nozzle Welds

At the end of 2006, defects were identified using ultrasonic testing in three of the pressurizer nozzle dissimilar metal (DM) welds at the Wolf Creek nuclear power plant. Understanding welding residual stress is important in the evaluation of why and how these defects occur, which in turn helps to determine the reliability of nuclear power plants. This paper presents analytical predictions of welding residual stress in the surge nozzle geometry identified at Wolf Creek. The analysis procedure in this paper includes not only the pass-by-pass welding steps, but also other essential fabrication steps of pressurizer surge nozzles. Detailed welding simulation analyses have been conducted to predict the magnitude of these stresses in the weld material. Case studies were carried out to investigate the change in the DM main weld stress fields resulting from different boundary conditions, material strength, weld sequencing, as well as simulation of the remaining piping system stiffness. A direct comparison of these analysis methodologies and results has been made in this paper. Weld residual stress results are compared directly to those calculated by the nuclear industry.

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

Modeling of Subcritical Crack Growth due to Stress Corrosion Cracking: Transition From Surface Crack to Through-Wall Crack

Recent work conducted using the Advanced Finite Element Analysis (AFEA) method to simulate the ‘natural’ crack growth of a circumferential PWSCC demonstrated that a subcritical surface crack can transition to a through-wall crack with significant differences between the inner diameter and outer diameter crack lengths. In the current version of the xLPR (Extremely Low Probability of Rupture) code, once the surface crack penetrates the wall thickness, an idealized through-wall crack (which has an equivalent area as the final surface crack) is formed. This type of crack transition was selected since no general stress intensity factor (K) solutions were available for crack shapes that would form during the transitioning stages, i.e., non-idealized or slanted through-wall cracks. However, during the pilot study of the xLPR code, it has been identified that this crack transition method may provide non-conservative results in terms of leak-rate calculations. In this paper, in order to compare the ‘natural’ versus ‘idealized’ crack transition behavior, limited example cases were considered where both crack transitions were simulated using 3D finite element analyses. In addition, leak-rate calculations were performed to study how the two different crack transition methods can affect the leak-rates. The results of the present study demonstrate that the ‘idealized’ transition from surface to through-wall crack can significantly affect the leak-rate calculations.Copyright

Stress Corrosion Crack Shape Development Using AFEA

Typical ASME Section XI subcritical cracking analyses assume an idealized flaw shape driven by stress intensity factors developed for semi-elliptical shaped flaws. Recent advanced finite element analyses (AFEA) conducted by both the United States Nuclear Regulatory Commission (U.S.NRC) and the nuclear industry for long circumferential indications found in the pressurizer nozzle dissimilar metal welds at the Wolf Creek power plant suggest that the semi-elliptical flaw assumption may be overly conservative in some cases. The AFEA methodology that was developed allowed the progression of a planar flaw subjected to typical stress corrosion cracking (SCC)-type growth laws by calculating stress intensity factors at every nodal point along the crack front, and incrementally advancing the crack front in a more natural manner. Typically, crack growth analyses increment the semi-elliptical flaw by considering only the stress intensity factor at the deepest and surface locations along the crack front, while keeping the flaw shape semi-elliptical. In this paper, a brief background to the AFEA methodology and the analyses conducted in the Wolf Creek effort will be discussed. In addition, the predicted behavior of surface cracks under normal operating conditions (plus welding residual stress) using AFEA will be investigated and compared with the semi-elliptical assumption. Conclusions on the observation of when semi-elliptical flaw assumptions are appropriate will be made. These observations will add insight into the conservatism of using an idealized flaw shape assumption.

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

Fracture Toughness Behavior of Complex Cracks in Dissimilar Metal Welds

Currently, J-estimation scheme procedures to predict the load-carrying capacity of idealized circumferential through-wall cracks in nuclear grade piping materials employ analytical or numerical procedures coupled with the fracture toughness of the material to predict the pipe response. However, with the advent of primary water stress corrosion cracking (PWSCC), complex-shaped cracks occur in dissimilar metal (DM) welds. These welds consist of a nickel-based weld joining stainless steel and carbon steel base metals.The NRC Office of Nuclear Regulatory Research (RES) is conducting a program to investigate the behavior of circumferential through-wall and complex cracks in DM welds. In a prior paper, a series of full-scale pipe bend and laboratory-sized fracture experiments were documented. Initial analyses of those test results suggest that reasonable prediction of through-wall crack response is obtained from typical J-estimation scheme procedures using the weld toughness from a compact tension (CT) specimen and the appropriate material strength. In addition, the J-R curves from the through-wall cracked pipe tests, calculated using published η-factor solutions and numerical techniques, were very similar to the CT J-R curves. In this paper, the fracture toughness for the circumferential complex cracked experiments, which was developed from a modified η-factor solution, is presented. These results are compared to the CT and through-wall crack pipe J-R curve results. In addition, predictions of load carrying capacity using the complex crack J-R curve and through-wall crack J-estimation schemes are presented and illustrate the need for the development of a complex crack J-estimation scheme. To support this development, a net-section collapse solution and a modified K-solution is presented. Finally, the need for additional work to generalize the elastic solution and its incorporation into a closed-form J-estimation scheme is discussed.Copyright

Comparison of Through-Wall and Complex Crack Behaviors in Dissimilar Metal Weld Pipe Using Cohesive Zone Modeling

Cohesive zone modeling has been shown to be a convenient and effective method to simulate and analyze the ductile crack growth behavior in fracture specimens and structures. Recently, authors have applied the cohesive zone model to simulate the ductile fracture behavior of a through-wall cracked pipe test consisting of a single material. In this paper, cohesive zone modeling has been applied to simulate the ductile crack growth in dissimilar metal weld pipe tests that was recently conducted by the U.S. NRC. Two crack types, i.e. through-wall and complex cracks, were simulated in the work. This paper describes how the cohesive parameters were determined and discusses in detail about the finite element modeling of the cohesive zone model. Various fracture parameters were compared between the finite element analyses and the experiments to validate the model. The results of the cohesive zone models showed good agreement with the pipe test results. Furthermore, the results of the cohesive zone model demonstrate that the fracture toughness (J at crack initiation, Jinit.) of the complex cracked pipe can be significantly lower (factor of 0.41) than that of the through-wall cracked pipe.Copyright

Dissimilar Metal Weld Pipe Fracture Testing: Analysis of Results and Their Implications

Typically in flaw evaluation procedures, idealized crack shapes are assumed for both subcritical and critical crack analyses. Past NRC-sponsored research have developed estimation schemes for predicting the load-carrying capacity of idealized cracks in nuclear grade piping and similar metal welds at the operating conditions of nuclear power reactors. However, recent analyses have shown that growth of primary water stress corrosion cracks (PWSCC) in dissimilar metal (DM) welds is not ideal; in fact, very unusual complex crack shapes may form, i.e., a very long surface crack that has a finite length through-wall crack in the same plane. Even though some experimental data on base metals exists to demonstrate that complex shaped cracks in high toughness materials fail under limit load conditions, other experiments demonstrate that the tearing resistance is significantly reduced. At this point, no experimental data exists for complex cracks in DM welds. In addition, it is unclear whether the idealized estimation schemes developed can be used to predict the load-carrying capacity of these complex-shaped cracks, even though they have been used in past analyses by the nuclear industry. Finally, it is unclear what material strength data should be used to assess the stability of a crack in a DM weld.The NRC Office of Nuclear Regulatory Research, with their contractor Battelle Memorial Institute, has concluded an experimental program to confirm the stability behavior of complex shaped circumferential cracks in DM welds. A combination of full-scale pipe experiments and a variety of laboratory experiments were conducted. A description of the pipe test experimental results is given in a companion paper. This paper describes the ongoing analyses of those results, and the prediction of the load-carrying capacity of the circumferential cracked pipe using a variety of J-estimation scheme procedures. Discussions include the effects of constraint, appropriate base metal material properties, effects of crack location relative to the dissimilar base metals, and the limitations of the currently available J-estimation scheme procedures. This paper concludes with plans for further development of J-estimation scheme procedures for circumferential complex cracks in DM welds.Copyright