IAEA-TECDOC-981 Assessmentand management of ageing of major nuclear power plant components important to safety: Steam generators INTERNATIONAL ATOMIC ENERGY AGENCY The IAEA does not normally maintain stocks of reports in this series. However, microfiche copie f thesso e reportobtainee b n sca d from ClearinghousS I N I e International Atomic Energy Agency Wagramerstrasse 5 0 10 P.Ox Bo . A-1400 Vienna, Austria Orders shoul accompaniee db prepaymeny db f Austriao t n Schillings 100, in the form of a cheque or in the form of IAEA microfiche service coupons which may be ordered separately from the IN IS Clearinghouse. The originating Section of this publication in the IAEA was: Engineering Safety Section International Atomic Energy Agency Wagramerstrasse5 0 10 P.Ox Bo . A-1400 Vienna, Austria ASSESSMENT AND MANAGEMENT OF AGEING OF MAJOR NUCLEAR POWER PLANT COMPONENTS IMPORTANT TO SAFETY: STEAM GENERATORS IAEA, VIENNA, 1997 IAEA-TECDOC-981 ISSN 1011-4289 ©IAEA, 1997 Printed by the IAEA in Austria November 1997 FOREWORD At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experienc s showha e n that ineffective controe ageinth f o gl degradatio majo e componentP th f nNP ro s (e.g. cause unanticipatey db d phenomeny b d aan operating, maintenance, desig manufacturinr no g errors jeopardizn )ca e plant safet alsd yoan plant life. Ageing in NPPs must be therefore effectively managed to ensure the availability desigf o n functions throughou plane tth t service life. Fro safete mth y perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This serieTECDOa n i reportf assessmense o e th on Cn s si o managemend tan ageinf to g majo e componentP th f NP ro s importan safetyo t reporte Th .base e s ar experienc n do d ean practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Importan Safeto t t y whic IAEe issues th h wa A 1992y n i db . They have been compiled using contributions from technical experts in typically 10 to 12 countries for each report feedbaca , k fro mSeptembea r 1994 Technical Committee Meeting attende3 5 y db technical experts from 21 Member States (who reviewed first drafts in specialized working groups) revied an , w comments from invited specialists. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canada deuterium-uranium (CANDU) reactor, boiling water reactor (BWR), pressurized water reactor (PWR) wated an , r moderated, water cooled energy reactor (WWER) plant documentee sar d in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs; and also to provide a common technical basis for dialogue between plant operator regulatord an s s when dealing with age-related licensing issues. Since the reports are written from a safety perspective, they do not address life or life- cycle managemen e planth f t o tcomponents , which involve e integratioth s f ageino n g management and economic planning. The target audience of the reports consists of technical experts from NPP frod san m regulatory, plant design, manufacturin technicad gan l support organizations dealing with specific plant components addressed in the reports. The component addresse presene th n di t publicatio e steath ms ni generatoe th f o r CANDU, PWR and WWER nuclear power plants. The contributors to the drafting and review of this TECDOC are identified at the end of this publication. Their work is greatly appreciated i particularh , contributione th , P.Ef so . MacDonald . MaruskC , V.Nd aan . Shah are acknowledged. The officer who directed the preparation of the report was J. Pachner of Divisioe th Nucleaf no r Installation Safety. EDITORIAI. NOTE preparingIn this publication press,for IAEAthe staffof have pages madethe up fromthe original manuscript(s). The views expressed do not necessarily reflect those of the IAEA, the governments nominating ofthe Member Statesnominatingthe or organizations. Throughout textthe names Memberof States retainedare theyas were when textthe was compiled. Theof use particular designations countriesof territoriesor does imply judgementnot any by the publisher, legalthe IAEA,to the status as suchof countries territories,or theirof authorities institutionsand delimitation ofthe or of their boundaries. The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA. CONTENTS 1. INTRODUCTION ..................................................... 1 1.1. Background ...................................................... 1 1.2. Objective ........................................................ 3 1.3. Scope ........................................................... 3 1.4. Structure ......................................................... 3 2. STEAM GENERATOR DESCRIPTIONS .................................5 . recirculatinR 2.1PW . g steam generators ..................................5 . 2.2. CANDU recirculating steam generators ...............................7 . once-througR 2.3PW . h steam generators ..................................4 1 . 2.4. WWER steam generators ............................................ 16 3. STEAM GENERATOR DESIGN BASIS, FABRICATION AND MATERIALS .... 18 3.1. Code specificationd san s ...........................................8 1 . 3.2. Fabrication and materials ............................................ 18 3.2.1. Heat exchanger tubes ........................................ 18 3.2.2. Tube installation in the tubesheet ............................... 21 3.2.3. Tube supports .............................................. 22 3.2.4. Feedwater nozzl sheld ean l ...................................7 2 . STEA4 M GENERATOR DEGRADATION MECHANISMS ...................8 2 . 4.1. PWR and CANDU recirculating steam generator tubes .................... 28 4.1.1. Primary water stress corrosion cracking (PWSCC) ................. 29 4.1.2. Outside diameter stress corrosion cracking (ODSCC) ............... 36 4.1.3. Fretting, wear and thinning .................................... 38 4.1.4. Pitting .................................................... 41 4.1.5. Denting ................................................... 41 4.1.6. High-cycle fatigue .........................................2 4 . 4.1.7. Wastage .................................................. 43 4.2. PWR once-through steam generator tubes ............................... 43 4.2.1. Erosion-corrosion ..........................................4 4 . 4.2.2. High-cycle fatigue .......................................... 44 4.2.3. Low-temperature primary-side stress corrosion cracking ............ 45 4.2.4. Outside diameter intergranular stress corrosion cracking (IGSCC) and intergranular attack (IGA) ................................5 4 . 4.3. WWER steam generator tubes ........................................ 46 4.4. Tube rupture events ...............................................6 4 . 4.4.1. Tube ruptures .............................................6 4 . 4.4.2. Incipient tube rupture events .................................. 52 CANDd an R 4.5U PW stea. m generator shell, feedwater nozzl tubesheed ean t ....2 5 . 4.5.1. Corrosion-fatigue ..........................................3 5 . 4.5.2. Transgranular stress corrosion cracking .......................... 55 4.5.3. High-cycle fatigue .......................................... 55 4.5.4. Erosion-corrosion ........................................... 55 4.6. WWER collector, shell, and feedwater distribution system .................. 57 4.6.1. Stress corrosion crackin WWER-100e th f go 0 collectors ...........7 5 . 4.6.2. Erosion-corrosio feedwatee th f no r distribution system .............9 5 . 4.6.3. Failur collectof eo r cover bolts ................................0 6 . 4.7. Summar currenf yo t world experience .................................1 6 . 5. STEAM GENERATOR AGEING MANAGEMENT: OPERATIONAL GUIDELINES .......................................... 66 5.1. Primary coolant system water chemistry control parameters ................6 6 . 5.2. Secondary coolant system water chemistry control parameters ..............9 6 . 5.3. Measures to control secondary-side chemical impurity incursions ............ 77 5.4. Measures to remove secondary-side impurities ........................... 79 5.5. Measure controo st l steam generator deposits ...........................0 8 . 6. STEAM GENERATOR INSPECTIO MONITORIND NAN G REQUIREMENTS AND TECHNOLOGIES ...............................................2 8 . 6.1. Tubing inspection requirements ......................................2 8 . 6.1.1. Tubing inspection requirements in the USA ...................... 82 6.1.2. Tubing inspection requirement Canadn si a ......................7 8 . 6.1.3. Tubing inspection requirement Czece th n shi Republic .............8 8 . 6.1.4. Tubing inspection requirements in France ........................ 88 6.1.5. Tubing inspection requirement Germann si y .....................8 8 . 6.1.6.