Weld residual stress (WRS) distributions are an important input into fracture mechanics evaluations necessary to determine the residual lives of dissimilar metal welds (DMWs). Since the DMW geometry and the presence or absence, size, and location of weld repairs is nozzle specific, finite element WRS analysis is often used to predict through-wall weld residual stress distributions. It is important to note that despite small differences in plant specific geometry or weld location specific weld repair geometry there are substantial similarities between the configurations that have been evaluated in the numerous weld specific finite element WRS analyses documented in the literature. Important insight can be gained from parametric studies of simplified geometries in order to understand the significance of different parameters on the resulting WRS distributions. The results of such studies can allow engineers to focus resources on refining accuracy of critical inputs and to support simplified model development suitable for incorporation into design and fitness for service codes. This paper documents the results of various studies performed to validate the ability to use a simplified pipe-to-pipe model for simulating relative effects on through-wall WRS distributions of pipe and weld repair geometry, investigate the effect of pipe mean radius to wall thickness ratio, weld repair depth (ID and OD), and weld repair sequence. Fifteen cases are analyzed. The dimensions selected for each case span a range of wall thickness, Rm/t and depth of repair values representative of typical Boiling Water Reactor (BWR) nozzle DMWs. The results are used as input into a simplified WRS model presented in a separate paper [17].
In April of 2008, the U.S. Nuclear Regulatory Commission’s Office of Nuclear Regulation released a draft Regulatory Issue Summary (RIS) regarding fatigue analysis of nuclear power plant components. The final RIS was published in December of 2008 with little modification. At issue in the RIS was the use of a single stress term for use with Green’s Functions, as opposed to the ASME Code, Section III method of combining all six stress components, which could potentially lead to non-conservative results in a fatigue usage evaluation. Single stress term Green’s Functions have been commonly deployed in on-line fatigue monitoring systems, and they have been used to quickly generate thermal transient stress histories in place of more rigorous finite element analysis of transients. The conclusions of the RIS were based on a comparison of results from a single-stress term Green’s Function analysis and a follow-on six-stress component confirmatory analysis performed on a BWR-4 feedwater nozzle. The confirmatory analysis considered all six stress components in accordance with ASME Code, Section III, Subsection NB, Subarticle NB-3200 methodology, and yielded higher fatigue usage in the nozzle corner region, thus calling into question the adequacy of the Green’s Function analysis and its associated judgments. Further concerns were raised because of the use of the single stress term Green’s Function approach in some on-line fatigue monitoring systems. However, the RIS did not address the many differences in inputs that were used in the confirmatory analysis that also contributed to the increase in fatigue usage. This paper re-investigates the BWR-4 feedwater nozzle analysis that originally motivated the promulgation of the subject RIS. The paper parametrically investigates the different inputs that were used in each analysis to further understand the differences between the single stress term Green’s Function and follow-on confirmatory analyses, and to clarify some of the misunderstandings about single stress term Green’s functions that may have been created by the publication of RIS 2008-30 [1].
As plants prepare for license renewal, extending their original operating licenses from 40 years to 60 years of plant operation, various tasks are required to be performed. Some of the required tasks include validation of the current fatigue evaluations for ASME Class 1 components, evaluation of environmental affects for a subset of those components, and calculation of revised component life to demonstrate acceptability for the entire license renewal period. Validation of the current fatigue evaluations is a significant challenge for older-vintage plants because the documentation of transient parameters and event severity was often not rigorous, especially during the early years of plant operation. This is primarily because historical data gathered in the early years was difficult to obtain before the implementation of modern computer systems and was therefore only collected at limiting times and locations. As a result, transients were typically considered to be full design basis severity events. Also, the design basis transient definitions, as defined in the original Final Safety Analysis Report (FSAR), were limited to a simpler, bounding set to help simplify the process. Whereas these simplifications are typically conservative and make it easier for the staff implementing plant procedures, they can challenge the ability of utilities to be able to demonstrate long-term acceptability for license renewal where additional margins are required. The purpose of this paper is to provide an example of the efforts expended on this issue for the Pilgrim Nuclear Power Station (PNPS). The intent is to educate the reader on the efforts involved, the vast array of input that is needed, and the possible hurdles that may be encountered for the validation and re-evaluation of Class 1 component fatigue for license renewal. Prior to starting work on the fatigue analysis, PNPS requested that SI develop a design specification for the fatigue reevaluation project. The importance of this document is that it maintained consistency and documented methods to be used along with design inputs.
The Pilgrim Nuclear Power Station (PNPS) License Renewal Application (LRA) was submitted to the NRC in January 2006. As a part of the LRA submittal, Entergy committed to manage the aging effects of the reactor coolant environment on fatigue usage during the extended period of operation. As a part of PNPS’s aging management strategy, fatigue analysis was performed for 60-years of operation including EAF. This document provides some background information on the PNPS fatigue analysis, a summary of related information contained in the LRA submitted for NRC review and the specific actions taken to address the LRA commitments. A previous paper (PVP2010-25329) was prepared to provide an example of efforts expended to demonstrate long-term acceptability for license renewal where additional margins are required for PNPS Nuclear Power Station. The intent was to educate the reader on the efforts involved, the vast array of input that is needed and the possible hurdles that may be encountered for the validation and re-evaluation of Class 1 component fatigue for license renewal. The previous PVP paper discussed the foundation aspects necessary to prepare a refined fatigue evaluation with environmental affects for the plant. These include plant history, B31.1 piping analyses, environmental considerations, transient history, design specification and refined nozzle evaluations. The present paper discusses the importance of laying the appropriate foundation in the context of the final results that were obtained. It provides additional background information and actions taken to address the license renewal commitments and summarizes detailed results.
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