Flaws detected in nuclear power plant components during in-service inspections are typically evaluated based on stress intensity factor influence coefficient databases and solutions from industry standards and public literature (e.g. API-579, ASME Section XI Code, WRC-175 Bulletin, Raju-Newman, etc). For certain components in the Pressurized Water Reactor (PWR) nuclear power plants, such as Bottom Mounted Instrumentation (BMI) nozzles, the cylindrical component geometry may fall outside the applicability limits of stress intensity factor influence coefficient databases. This situation occurs where the thickness to inner radius ratios of the cylindrical geometry is greater than 1.0. Accurate stress intensity factor (SIF) solutions are essential to flaw evaluation since the SIFs are used in the determination of both the allowable flaw size and crack growth in order to determine acceptability of the detected flaw. In this paper, stress intensity factor influence coefficients are generated based on a three-dimensional finite element analysis for axial flaws located on the inside surface and outside surface of a cylindrical component with thickness to inner radius ratios (t/Ri) of 1, 2, 4, & 6. Non-dimensional influence coefficients are determined at the deepest point of the crack front and the surface point of the flaw based on a 4th order polynomial fit for a through-wall stress profile. The influence coefficients are generated for semi-elliptical flaws with a/c ratios = 0.125 through 2; where a is depth of the elliptical flaw, and c is the half-length of the elliptical flaw. The influence coefficients developed are suitable for calculating stress intensity factors for cylindrical components with high thickness to inner radius ratios.
Reference fatigue crack growth (da/dN) curves for pressurized water reactor (PWR) environments have been proposed for ASME Section XI flaw evaluation applications in Code Case N-809. The reference curves are dependent on temperature, loading rate, mean stress, and cyclic stress range which are all contained in the da/dN model. This paper presents the application of N-809 in a fatigue crack growth analysis for a large diameter austenitic pipe in a PWR Reactor Coolant System main loop using the current analytical evaluation procedures in Appendix C of ASME Section XI. The example problem was used to evaluate the reference fatigue crack growth curves during the development of the code case and the results have been compared with other industry codes.
Primary Water Stress Corrosion Cracking (PWSCC) has been observed in pressurized water reactor (PWR) coolant system pressure boundary components. This type of cracking has been observed in Alloy 82/182 butt welds. Various repair and mitigation methods have been proposed and employed to address PWSCC. Case N-766 was developed as an alternative method for PWSCC mitigation. It is especially useful for applications in PWSCC susceptible regions where accessibility to the outer surfaces required for the other PWSCC mitigation processes is difficult or impractical. The method in Case N-766 involves isolating the PWSCC susceptible material from the primary water environment, thereby eliminating one of the three conditions that must exist for PWSCC to occur. Inlays provide a means for PWSCC mitigation or repair of existing PWSCC flaws while maintaining the inside surface contour essentially in its original configuration, without flow path restriction. Onlays, which do not require an excavation into the pipe ID, permit maintaining the inside surface contour essentially in its original configuration, or allow for weld buildup on the inside surface with a negligible change to the inner pipe diameter. The technical basis for the design, fabrication, and inspection requirements for inlays and onlays was documented in ASME Pressure Vessel and Piping Division Conference paper PVP2010-26164. Subsequently, the Nuclear Regulatory Commission (NRC) has submitted comments on Code Case N-766. Changes were proposed to the Code Case to address the NRC’s comments and it was suggested that a bounding PWSCC crack growth evaluation be added to the technical basis to demonstrate that a flaw which might have been missed by surface examination would not grow through the inlay or onlay before the next inspection. The purpose of this paper is to augment the existing technical basis for Code Case N-766 to support the changes proposed.
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.
Thermal stratification is a common phenomenon in the surge lines of Pressurized Water Reactors (PWR). The stratification temperature difference (ΔT) and cyclic action severities are most prevalent during the heatup and cooldown operations of a PWR, when the system ΔT between the pressurizer and the Reactor Coolant System (RCS) hot leg is the greatest and system inventory fluctuations are highest. This paper describes the computer simulation of thermal stratification loading in a surge line nozzle connected to the RCS hot leg to correlate to unusual behavior of plant sensor data in the hot leg and the subsequent development of a monitoring model to account for thermal stratification effects in the transient and fatigue evaluation performed in the online monitoring system. What makes this particular investigation unique is the geometry of the nozzle of interest. In many PWRs, the surge line and the surge line hot leg nozzle are horizontal at the hot leg connection. This particular nozzle is oriented at an upward angle before the attached surge line piping bends into a horizontal configuration. This orientation required a more detailed treatment of the stratification effects than has been typically developed for horizontal nozzles, with respect to both the orientation and the potentially detrimental effects of increased cyclic behavior indicated by nearby temperature sensors. This investigation combined Computational Fluid Dynamics (CFD) modeling of the system to correlate the plant data with a detailed stress model that will enable the fatigue usage factor calculation in the plant’s online transient and fatigue monitoring system.
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