Daniel V. Sommerville

Additional Improvements to Appendix G of ASME Section XI Code for Nozzles

This paper is the second in a continuing series of papers to highlight additional bases and recommended improvements to Appendix G. In 2008, the authors prepared a paper that reviewed some of the original basis documents for Appendix G for calculating pressure-temperature (P-T) limits and identified recommended areas for improvement. The 2008 paper discussed the fact that the original Appendix G in Section XI of the ASME Code was primarily based on Welding Research Council (WRC) Bulletin 175, and identified the changes that have been made to Appendix G over the past 20 years. However, the nozzle corner solutions have remained the same as those given in WRC 175. Proposed revisions to Appendix G are included in this paper regarding the stress intensity factor (K) calculation procedures for pressure and thermal gradient loading at a nozzle corner based on the various solutions described in the authors’ previous paper and on other more recent investigations. The current paper is focused on incorporating the results of additional studies that have been completed associated with nozzle corner solutions. This additional work has become more important because plants must address the effects of nozzles in the reactor pressure vessel (RPV) as a part of pressure-temperature (P-T) curve development, especially if the nozzles are located sufficiently close to the active core region such that they accumulate significant fluence. In addition, the treatment of operating stresses exceeding the material yield stress is discussed and the basis for the limit of material yield strength to 90 ksi in G-2110(b) is provided. Finally, this paper identifies other areas for future improvements in Appendix G, including those areas remaining to be addressed from prior work.Copyright

Survey of BWR Recirculation Line Break Acoustic Loads

There have been numerous industry communications over the past six years identifying issues pertaining to incorrectly calculated or erroneously omitted recirculation line break acoustic loads. Acoustic loads caused by a postulated recirculation line break (RLB) loss of coolant accident (LOCA) are one of the required design basis events that must be considered for stress analyses of Boiling Water Reactor (BWR) internal components such as jet pumps, core shrouds, access hole covers, and shroud support assemblies. These acoustic loads must also be considered for fracture mechanics evaluations performed to determine allowable operating periods for flaws detected during in-service inspections. Various organizations have studied the recirculation line break event and have developed methodologies to calculate the associated loads on BWR internal components. In general, loads are calculated in a plant specific and operating condition specific manner. Once these loads are available then the necessary structural evaluations can be performed. Load calculations can be expensive and time consuming. The authors have previously proposed a method for determining plant specific bounding RLB acoustic loads such that a utility can eliminate the need to periodically update the acoustic load calculations for each plant. This paper adds to the body of literature on this topic by providing a survey of BWR shroud acoustic loads, representative of the range of loads expected for the fleet, and provides a bounding load estimate that can be used for any plant in the fleet. The load survey supports previous arguments that the AC loads are expected to be substantially similar from plant to plant, based on the substantial similarity in plant operating conditions and geometry. The bounding load estimate provides significant value in that a utility engineer can rapidly obtain conservative loads to support either plant specific structural evaluations or fleet studies intended to investigate the significance of the load on inspection intervals, flaw tolerance, etc. Cost savings can be realized for utilities and fleet wide insight can be gained from understanding the similarity between RLB acoustic loads across the operating BWR fleet.Copyright

Simplified Dissimilar Metal Weld Through-Wall Weld Residual Stress Models for Single V Groove Welds in Cylindrical Components

Two simplified models are developed and benchmarked for predicting through-wall axial and hoop weld residual stress (WRS) distributions in single V groove dissimilar metal welds (DMWs) joining cylindrical components such as piping or nozzles. The models can be used to predict WRS distributions for different pipe mean radius to wall thickness ratios (Rm/t) without an inside surface repair and WRS distributions at a single Rm/t for various inside surface weld repair depth to pipe thickness ratios (x/t). The models are developed by approximating the through-wall WRS distribution using a finite Fourier series where the coefficient of each term in the Fourier series is determined using a linear equation in which the Rm/t or x/t is the independent parameter. The model for the unrepaired condition has been benchmarked against two plant specific finite element WRS analyses of BWR nozzle to safe end welds as well as experimental and FEA WRS data from the PWR pressurizer safety/relief nozzle to safe end weld documented in MRP-317. The weld repair model has been benchmarked against the pressurizer surge nozzle experimental data presented in MRP-316. The models have been used to perform numerous plant specific DMW residual life calculations and can save significant time and money when performing weld specific fracture mechanics analyses.Copyright

Treatment of Stresses Exceeding Material Yield Strength in ASME Code Section XI Appendix G Fracture Toughness Evaluations

A previous PVP paper [1] identified suggested improvements to be made to ASME Code, Section XI, Nonmandatory Appendix G, “Fracture Toughness Criteria for Protection Against Failure” [2]. That paper also identified that the current version of Appendix G does not have any provisions for when the calculated operating stress (pseudo stress) exceeds the material yield strength. The treatment of stresses exceeding yield was included in earlier versions of Appendix G, but it was removed via Code Action ISI-94-40 in 1995. The specific reasons for removal of these provisions were not documented.In some Appendix G postulated flaw evaluations for pressure-temperature (P-T) limits, the calculated total linear-elastic (or pseudo) stress (i.e., including the primary stress due to pressure loading and thermal stress) may exceed the material yield stress. The ASME Section XI Working Group on Operating Plant Criteria (WGOPC) decided that this provision needed to be more fully considered, with appropriate benchmarking and possible adjustments to Appendix G made consistent with the current state of knowledge in elastic-plastic fracture mechanics (EPFM) methods. This is appropriate since the state of knowledge in EPFM has significantly advanced since the time the technical basis for Appendix G was established, as documented in Welding Research Council (WRC) Bulletin No. WRC-175, which was published in 1972. Furthermore, EPFM provides an improved method for evaluating the effects of high stresses.This paper describes the results of preliminary investigations of stresses exceeding the material yield stress in fracture toughness assessments associated with Appendix G. Also included in the technical evaluations presented are the temperature conditions for which upper shelf conditions are present and where EPFM methods are applicable.Copyright

Comparison of Handbook and 3-D Finite Element Analysis LEFM Solutions for a Threaded Fastener

As part of license renewal activities, many Boiling Water Reactor utilities must assess the integrity of bolted components within the reactor pressure vessel. Since inspection techniques capable of identifying cracking in the threaded regions of the core plate bolting do not currently exist it is important for flaw tolerance evaluations to be performed with as much accuracy as possible in order to obtain realistic inspection intervals and to assess the degree of redundancy in the bolted joint designs. Both 3-D finite element analysis and handbook linear elastic fracture mechanics solutions are used to assess the flaw tolerance of a core plate bolt design. Consideration of the relevant degradation and crack growth mechanisms is given for both initial crack configuration(s) and crack growth calculations. The 3-D finite element analysis is performed to investigate the effects of various simplifications in the available handbook solutions presented in the literature. The results of the flaw tolerance evaluations using both methods are compared and conclusions drawn regarding the applicability of the available handbook solutions for similar work in the future.Copyright

In-Service Inspection Strategy for Alloy X-750 BWR Jet Pump Beams Based Upon Linear Elastic Fracture Mechanics Analysis

Jet pumps in a boiling water reactor (BWR) are located in the annulus region between the core shroud and the reactor vessel wall and provide core flow to control reactor power. Between 16 and 24 jet pumps are included in BWR/3 through BWR/6 plants, depending on the plant rating. The inlet mixer assembly of the jet pump is secured in place with a hold down mechanism called a jet pump beam. This beam is fabricated of alloy X-750 and tensioned to 58–74% of the yield stress of the material, depending on the beam design. In recent years, more attention has been placed upon inter-granular stress corrosion cracking (IGSCC) of alloy X-750 BWR internal components as a result of in-service cracking and failures. BWR plant owners have implemented actions to manage IGSCC of jet pump beams and assemblies through increased inspections and changes to process specifications for X-750. However, a thorough understanding of the flaw tolerance of the jet pump beam was not available to guide the periodicity of inspections as well as to define critical flaw sizes needed to validate the capability of inspection techniques. This paper describes a linear elastic fracture mechanics (LEFM) evaluation in which the flaw tolerance of the existing jet pump beam designs is established and used to recommend inspection frequencies for the jet pump beam. Industry operating experience is used to assess the credibility of the results obtained from this evaluation. This work illustrates an example of the use of LEFM to develop a technically defensible basis for the required inspection regions and the frequency of inspection for an alloy X-750 BWR internal component and helps to establish the necessary sensitivity of non-destructive examination technology to be used to examine the component.Copyright