Steam Generator Tube Failures

NUREG/CR-6365 INEL-95/0383 Steam Generator Tube Failures RECEIVED HAY 2 2 t99S Prepared by P. E. MacDonald, V. N. Shah, L. W. Ward, P. G. Ellison Idaho National Engineering Laboratory Lockheed Idaho Technologies Company Prepared for U.S. Nuclear Regulatory Commission DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsi• bility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. 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DISCLAIMER Portions of this document may lie illegible in electronic image products, images ane produced foam the best available original document* NUREG/CR-6365 INEL-95/0383 Steam Generator Tube Failures Manuscript Completed: April 1996 Date Published: April 1996 Prepared by P. E. MacDonald, V. N. Shah, L. W. Ward, P. G. Ellison Idaho National Engineering Laboratory Lockheed Idaho Technologies Company Idaho Falls, ID 84315 J. R. Boardman, NRC Project Manager Prepared for Safety Programs Division Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC Job Code E8238 NUREG/CR-6365 has been reproduced from the best available copy. ABSTRACT A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Also, nuclear power plant design basis accidents, such as a main steam line break, may cause multiple failures of badly degraded steam generator tubes. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service. However, a continuing issue has been exactly what constitutes an appropriate and timely inspection and which degraded tubes are still fit for service. There have been many different approaches to this problem throughout the world. Also, the most widely used inspection equipment is not able to detect and size all the degradation of concern. Job Code E8238-Specialized Technical Assistance iii NUREG/CR-6365 CONTENTS ABSTRACT iii LIST OF FIGURES xi LIST OF TABLES xvi EXECUTIVE SUMMARY xix ACRONYMS xxix ACKNOWLEDGMENTS xxxi 1. INTRODUCTION 1 2. STEAM GENERATOR DESIGN 4 2.1 Pressurized Water Reactor Recirculating Steam Generators 4 2.2 CANDU Reactor Recirculating Steam Generators 6 2.3 Pressurized Water Reactor Once-Through Steam Generators 14 2.4 Russian VVER Steam Generators 15 2.5 Codes and Specifications 15 2.6 Fabrication and Materials 20 2.6.1 Heat Exchanger Tubes 20 2.6.2 Tube Installation in the Tubesheet 24 2.6.3 Tube Supports 25 2.6.4 Steam Generator Shells and Feedwater Nozzles 28 3. STEAM GENERATOR DEGRADATION MECHANISMS, SITES, AND FAILURE MODES 34 3.1 Summary of the PWR and CANDU Tube Degradation Problems 34 3.2 Pressurized Water Reactor and CANDU Reactor Recirculating Steam Generator Tubes 39 3.2.1 Primary Water Stress Corrosion Cracking 39 3.2.2 Outside Diameter Stress Corrosion Cracking 48 3.2.3 Fretting, Wear and Thinning 52 3.2.4 Pitting 57 3.2.5 Denting 59 3.2.6 High-Cycle Fatigue 61 3.2.7 Wastage 61 v NUREG/CR-6365 3.3 Pressurized Water Reactor Once-Through Steam Generator Tubes 62 3.3.1 Erosion-Corrosion 62 3.3.2 High-Cycle Fatigue 63 3.3.3 Low-Temperature Primary-Side Stress Corrosion Cracking 64 3.3.4 Outside Diameter Intergranular Stress Corrosion Cracking and Intergranular Attack 64 3.4 Russian VVER Steam Generator Tubes 64 3.5 Steam Generator Tubing Residual Life Estimates 65 3.6 Pressurized Water Reactor and CANDU Reactor Steam Generator Shell, Feedwater Nozzle, and Tubesheet Degradation 67 3.6.1 PWR Steam Generator Shells 70 3.6.2 Feedwater Nozzles 71 3.6.3 Erosion-Corrosion of the Thermal Sleeves, J-tubes, Feedrings, and Divider Plates 71 3.7 Russian VVER Collector, Shell, and Feedwater Distribution System Degradation .... 73 3.7.1 Stress Corrosion Cracking of the WER-1000 Collectors 73 3.7.2 Erosion-Corrosion of the WER Feedwater Distribution Systems

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