This article outlines changes proposed for implementation in the ASME Boiler and Pressure Vessel Code Section III Appendix N pertaining to Flow-Induced Vibrations (FIV). Several portions of Appendix N were originally written multiple decades ago and are not necessarily readily prescriptive for present-day analyses. An ASME Task Group on Flow-Induced Vibrations has been formed to identify areas of improvement to Appendix N so that the Appendix may be more useable in present-day FIV analyses, and a means of establishing consensus for characterization of this complex phenomena. This paper identifies opportunities for improvement to portions of Appendix N pertaining to Turbulence, Acoustic, and associated Structural Dynamic Modeling, and Data Processing. Specific suggestions are made for the following: Characterization of acoustic damping and acoustic harmonic responses, updated considerations for aeroacoustic phenomena, structural damping quantification of nuclear components, systematic approaches to modeling (random) turbulence-induced vibrations (including RMS-to-Peak ratio), considerations for leakage flow-induced instabilities, and opportunities to employ computational methods are discussed. Opportunities for the applicability of data processing algorithms such as the proper orthogonal decomposition and circumferential wavenumber decomposition are also discussed and an updated methodology for combination of random and deterministic loads for FIV analyses is presented.
The reactor vessel internals (RVI) are located within the reactor vessel, which is part of the reactor coolant system (RCS) loop in Westinghouse nuclear plant designs. Historically, the coolant passing through the RCS loop has been highly turbulent and has generated significant turbulence-induced excitation (TE) for the RVI. In an effort to analytically quantify the response of RVI structures due to TE for new designs, a methodology is employed which combines both first principle concepts as well as operational experience. As part of this process, TE-induced forcing functions are developed based upon the numerous flow fields around the components of interest. After the TE-induced forcing functions are developed, they are applied to a system finite element model (SFEM), in a transient dynamic finite element analysis to capture dynamic system-level interactions. To benchmark the response of the numerical model, both narrowband and broadband model responses are compared to empirical data extrapolated from model-scale flow-induced vibration test results. This comparison shows a strong agreement between the empirical data and the SFEM, validating that the dynamic response and system interactions of RVI structures due to TE can be accurately characterized through numerical simulation of the system.
Requirements for plant license renewal require the evaluation and management of time limited aging analyses for the period of extended operation. These include metal fatigue analyses, with additional consideration for the effects of the reactor water environment. Current license renewal requirements for evaluation of time limited aging analyses for fatigue have extended environmentally assisted fatigue (EAF) considerations to additional locations beyond the originally prescribed NUREG/CR-6260 locations. These requirements have resulted in implementation of a screening process to identify locations that are potentially more limiting for further evaluation of EAF. The screening process includes comparisons of fatigue locations within common bases, including consideration of applicable system transients, rigor of analysis methods, and material types. To date, the United States Nuclear Regulatory Commission has only accepted EAF screening evaluation results based on comparisons within similar material types. However, there are conditions under which a comparison of locations with different material types is valid. This paper presents a basis to compare locations having different material types, to potentially reduce the number of leading (“sentinel”) locations in EAF screening evaluations. The methodology is described through a systematic approach and basis to determine when locations are more limiting with respect to EAF when different material types are compared.
Nuclear Regulatory Commission (NRC) guidance for Subsequent License Renewal (SLR) applicants is provided in the NUREG-2191. Specifically, Section X.M1 of the NUREG-2191 includes guidance for aging management programs (AMP) which has two aspects related to fatigue analyses, one that verifies the continued acceptability of existing analyses through cycle counting and the other that provides periodically updated evaluations of the fatigue analyses to demonstrate that they continue to meet the appropriate limits. In addition, the NUREG-2191 provides requirements to account for reactor water environment by deriving cumulative fatigue usage (CUF) including environmental effects (CUFen) in component fatigue analyses for a set of sample critical components for the plant outlined in NUREG/CR-6260. Furthermore each applicant is required to determine plant-specific component locations in the reactor coolant pressure boundary that may be more limiting than those considered in NUREG/CR–6260. This paper presents a methodology to identify plant-specific primary equipment component locations that are potentially more limiting than the locations identified in NUREG/CR-6260, through a detailed review and ranking of the analyses of record (AORs) for a plant. The ranking approach implemented in this methodology assesses the level of technical rigor required to derive the CUF values and provides a means to appropriately compare CUF values within the same transient section. This paper also illustrates how this methodology was applied for a recent license renewal application.
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