Primary Water Stress Corrosion Cracking (PWSCC) of Pressurized Water Reactor (PWR) primary loop piping/nozzle Dissimilar Metal Weld (DMW) joints and Inter Granular Stress Corrosion Cracking (IGSCC) of Boiling Water Reactor (BWR) weld joints is an ongoing issue in the nuclear power industry. Recent field experiences with PWSCC of various DMW joints in US plants led to the development and application of an Advanced Finite Element Analyses (AFEA) methodology that permits crack propagation with a natural flaw shape. Crack growth and fracture evaluations for both PWR and BWR components are generally performed based on a conservative, idealized crack shape model, e.g. semi-ellipse, rectangle, etc., depending on the geometry of the crack and the component. Conventional evaluation methodologies and/or assumptions of this kind, in some cases may provide excessive conservatisms. The use of natural flaw shape development with crack propagation might provide a more realistic assessment of crack growth and structural integrity. The prime purpose of this study is to demonstrate the conservatism/margins in the conventional “idealized crack shape” methodology. A comparison study of crack growth behavior between the applications of the idealized and natural crack shape methodologies has been performed in order to assess the level of conservatism/margins in the conventional crack growth evaluation methodology and the possible impacts on the structural integrity evaluation for both PWR and BWR components. Comparison studies on the impacts of the differences in crack growth law and loading condition used for crack growth evaluations have been performed as well.
Evaluation of metal fatigue is one of the identified Time Limited Aging Analyses (TLAA) described in the License Renewal Applications for nuclear power plants. A related area is the evaluation of Environmentally Assisted Fatigue (EAF) as required by the Nuclear Regulatory Commission (NRC) in the United States. The focus is on fatigue sensitive components in power plants selected based on anticipated loading conditions and service experience. The pressurizer surge line was identified as a representative component for the evaluation of EAF for nuclear power plants of both older and newer vintage. A particular location of interest on the surge line is the surge line hot leg nozzle to pipe weld. Typical resolution of EAF concerns in a License Renewal Application requires the incorporation of large Environmental Fatigue Multipliers (Fen) in determining the cumulative usage factors for austenitic stainless steel components. Consequently, the 60-year projected cumulative usage factor at this weld location of interest has the potential to exceed the ASME Section III Code allowable limit of 1.0. For components that fail to demonstrate that the predicted cumulative usage factor is less than 1.0, ASME Section XI Non-Mandatory Appendix L provides guidance for evaluating the component’s fitness for service. The approach is based on a fatigue flaw tolerance evaluation and the implementation of an inspection strategy to demonstrate that growth of a postulated flaw would remain below the allowable flaw size for the component of interest. NUREG/CR-6934 establishes a technical basis for improvements to the initially published version of Appendix L requirements. These improvements have been incorporated in the 2008 Addenda to ASME Section XI Code. One of the improvements is the use of Equivalent Single Crack (ESC) aspect ratio to account for the effects of multiple fatigue crack initiations and the linking of these cracks as they grow to form a single long crack with large aspect ratio. Since the NRC in the United States has not yet endorsed the use of 2008 Addenda to the ASME Section XI Code which contains the latest version of Appendix L, fatigue flaw tolerance analysis is performed in accordance with the requirements in NUREG/CR-6934 for the current license renewal applications. This paper demonstrates the use of the fatigue flaw tolerance approach in accordance with the requirements in NUREG/CR-6934. Typical thermal stratification loading and thermal transients including the use of ESC aspect ratios are considered in the fatigue flaw tolerance evaluation. The results are used to demonstrate fitness for service for a typical 14-inch Schedule 160 pressurizer surge line when the predicted cumulative usage factor exceeds 1.0 for 60 years of plant operation.
Welding residual stress modeling is currently performed by researchers around the world using a wide variety of modeling methods to predict the final stress state of a completed weld. Among the key modeling assumptions used to perform a residual stress simulation are: • The geometric setup of the model, including boundary condition and assumptions. • The thermal and structural model simulating the welding process including lumping of weld passes. • The strain hardening input properties and hardening law. Researchers from Dominion Engineering, Inc. (DEI) and Westinghouse Electric Co. (WEC) have performed a benchmark comparison studying these key modeling assumptions and their results on the predicted welding residual stress distributions. Researchers from DEI and WEC have completed independent studies to validate their respective methods for calculating residual weld stress. In addition to the comparative evaluation, brief descriptions of the individual validations will be included in this paper. The weldment selected for evaluation is a typical reactor pressure vessel (RPV) outlet nozzle dissimilar metal safe end weld in a pressurized water reactor plant. This weld joins a low alloy steel nozzle to a stainless steel safe end using Alloy 182 weld material; this weld is completed in the manufacturing shop. The safe end is then field welded to the stainless steel reactor coolant loop piping. The residual stress distributions in the dissimilar metal welds, like the one selected, are important in predicting stress corrosion crack growth in Reactor Coolant System (RCS) components. The fabrication drawings for the selected RPV outlet nozzle were provided to both organizations, and independent residual stress simulations were performed using the best effort modeling techniques from each organization. This paper investigates the impact of the key modeling assumptions described above on the differences in the predicted welding residual stress distributions between the two simulation techniques. The results from the modeling comparison are provided in this paper.
The Boiling Water Reactor (BWR) shroud support dissimilar metal (DM) joints, which are made of Alloy182 nickel based alloy weld material, are susceptible to Intergranular Stress Corrosion Cracking (IGSCC). IGSCC of these weld joints is an industry wide issue that is receiving constant attention from both power plant operators and regulatory agencies. Crack growth due to IGSCC in the shroud support DM weld joints is typically evaluated based on conventional idealized crack shape instead of natural crack shape dictated by the crack tip stress intensity factors calculated along the entire crack front. Due to the complexity in the geometry and stress field at these weld joints, the use of natural crack shape as the crack propagates would provide a more realistic assessment of crack growth and structural integrity of the shroud support dissimilar metal weld joints. This paper describes the simulation of natural crack growth due to IGSCC in the shroud support DM weld joints through the use of advanced finite element fracture mechanics techniques. Typical normal operating stresses including welding residual stresses at these weld joints were considered in the IGSCC crack growth simulation. A comparison of the crack growth results between the use of idealized and natural crack shape was made to assess the impact on the structural integrity of the BWR shroud support DM weld joints.
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