Steam Generator Tube Integrity Program/Steam Generator Group Project: Final Project Summary Report (NUREG/CR-5117)

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Publication Information

Manuscript Completed: April 1990
Date Published: May 1990

Prepared by:
R.J. Kurtz, R.A. Clark, E.R. Bradley, W.M. Bowen,
P.G. Doctor, R.H. Ferris, F.A. Simonen

Pacific Northwest Laboratory
Richland, WA 99352

Prepared for:
Division of Engineering
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555

NRC FIN B2097

Availability Notice

Abstract

The Steam Generator Tube Integrity Program/Steam Generator, Group Project was a three-phase program conducted for the U.S. Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory.^(a) The main goal of the program was to provide the NRC with validated information on the reliability of nondestructive examination techniques to detect and size flaws in steam generator tubing and to determine the remaining integrity of service-degraded tubing. The information was used to determine the effectiveness of NRC Regulatory Guides 1.83, Rev. 1, and 1.121 to 1) define the frequency, extent and procedure for conducting nondestructive inservice inspections of steam generator tubing, and 2) define plugging limits of service-degraded tubing under normal operating and accident loading conditions. The program was performed in three phases. The first phase involved burst and collapse tests and single-frequency eddy-current (EC) examinations of typical steam generator tubing with precision machined flaws. The goal of Phase I was to develop empirical models of remaining tube integrity as a function of flaw type and size, and to determine the capability of EC inspection methods to detect and size tube degradation. In Phase II, a smaller number of specimens with the same flaw types were investigated, but tube specimens were degraded by chemical means rather than machining methods. This approach was used to better simulate the irregular geometry of service-induced degradation. In the final phase of the program, the retired-from-service Surry 2A Steam Generator was used as a test bed to investigate the reliability of inservice EC inspection equipment, personnel, and procedures, and as a source of service-degraded tubes for further validating the empirical equations of remaining tube integrity. In addition, the generator was also used to study the effects of primary-side chemical decontamination, the effectiveness of secondary-side visual examinations, characterizationý of tubesheet crevice corrosion products and sludge pile composition, and demonstrations of tube unplugging and repair techniques. This portion of the program also included participation by three foreign countries.

This report summarizes the findings of more than eleven years of research activity on steam generator tube integrity and inspection issues. The results of the Phase I pressure tests on mechanically-flawed steam generator tubing are presented. In addition, the laboratory EC sizing results on those tubes are summarized. A discussion of Phase II pressure test results on chemically-degraded tubes is given, along with a brief overview of leak-rate data for tubes with laboratory-produced axial or circumferential stress-corrosion cracks (SCC). Comparisons with a simple analytical leak rate model are discussed. Laboratory EC estimates of flaw size in Phase II specimens are described. To supplement the laboratory EC data and obtain an estimate of EC reliability to detect and size SCC, results of a mini-round robin involving several firms that routinely perform field inservice inspections are presented.

A major portion of the report is devoted to summarizing and integrating the results from the more than six years of research on the Surry 2A Steam Generator. A brief description of the acquisition, transportation and placement of the steam generator into a specially built examination facility is given. Prior to performing extensive nondestructive examination (NDE) of the generator tubing, dilute chemical decontamination of the channel head and removal of plugs placed during service was performed. To determine a post-service baseline condition of the generator, two nearly 100% multifrequency EC inspections were performed along with visual inspections from the secondary side. From these inspections, a subset of 320 tubes was selected for round robin inspection and development of the data base on EC inspection reliability. A description of the four round robins conducted is provided. A summary of the methods used to remove more than 550 tube segments to validate the round robin inspection data is given. Results from the metallurgical and visual examinations of these specimens are discussed. Burst tests of removed-from-generator specimens with pitting/wastage-type degradation are presented and correlated with the empirical models of remaining tube integrity. Statistical analyses of the combined metallurgical and EC data to determine probability of detection and sizing accuracy are reported along with a discussion of the factors which influenced the results. Analyses and Monte Carlo simulations used to evaluate and compare various sampling plans for inservice inspection are described. Finally, recommendations for improved inservice inspection and maintenance of steam generator tube integrity are given.

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^(a) Pacific Northwest Laboratory is operated for the U.S. Department of Energy by Battelle Memorial Institute under Contract DE-ACO6-76RLO 1830.